ORNL-TM-0732.txt

 

OAK RIDGE NATIONAL LABORATORY

operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION

ORNL- TM- 732
/73

MSRE DESIGN AND OPERATIONS REPORT
Part V
REACTOR SAFETY ANALYSIS REPORT
. E. Bedll
. N. Haubenreich

. B. Lindauer
. R. Tallackson

— A 9w

NOTICE

This document contains information of a preliminary nature and was prepared
primarily for internal use at the Oak Ridge National Laboratory. it is subject
to revision or correction and therafore does not represent a final report. The
information is not to be abstracted, reprinted or otherwise given public dis-
semination without the approval of the ORNL patent branch, Legal and Infor-
mation Control Department.

  
 
  
 
 
  
  
LEGAL NOTICE

 

 

 

This report was prepared as an account of Government sponsored work. Neither the United States,

nor the Commission, nor any persen acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or implied, with respect to the accuracy,
completeness, or usefulness of the information contained in this report, or that the use of
any information, cpparatus, method, or process disclesed in this report may not infringe
privately owned rights; or

B. Assumes any liobilities with respect to the use of, or for damages resulting from the use of
any information, apporatus, method, or process disclosed in this report,

As used in the above, '‘person acting on behalf of the Commission”” includes any employes or

contractor of the Commission, or employee of such contractor, to the extent that such employese

or contractor of the Commission, or smployee of such contracter prepares, disseminates, or
provides access to, any information pursuant to his employment or contract with the Commission,

or his employment with such contractor.

 

 

 

 

 

 

 

 
ORNL-TM-732

Contract No. W=-7405-eng-26

Reactor Division

MSRE DESIGN AND OPERATIQNS REPORT

Part V

REACTOR SAFETY ANALYSIS REPCRT

S. BE. Beall R. B. Lindauer
P. N. Haubenreich J. R. Tallackson

AUGUST 1964

 

 

 

OAK RIDGE NATTIONAL TABORATORY
Oak Ridge, Tennessee
operated by
UNLON CARBIDE CORPORATION
for the
U.5. ATOMIC ENERGY COMMISSION
iid

PREFACE

This report is one of a series that describes the design and opera-

tion of the Molten-Salt Reactor Experiment. All the reports are listed

below.

ORNL-TM-"728

ORNL~TM-729

ORNL~-TM~730%

ORNL-TM-731

ORNL-TM~732%

ORNL-TM-'733

CRNL-TM-907%%*

ORNL-TM-908%*

ORNL-TM-209%*

ORNL-TM-910%*

 

*Issued.

**¥These reports

MSRE Desgsign and Operations Report, Part I,
Degcription of Reactor Design, by
R. C. Robertson

MSRE Design and Operations Report, Part II,
Nuclear and Process Instrumentation, by
J. R. Tallacksocon

MSRE Design and Operations Report, Part III,
Nuclear Analysis, by P. N. Haubenreich and
J. R. Engel, B. E. Prince, and H. C. Claiborne

MSRE Design and Operations Report, Part 1V,
Chemistry and Materials, by F. F. Blankenship
and A, Taboada

MSRE Design and Operations Report, Part V,
Reactor Safety Analysis Report, by S. E. Beall,
P. N, Haubenreich, R. B. Lindauer, and
J. R, Tallackson

MSRE Design and Operations Report, Part VI,
Operating Limits, by 5. E. Beall and
R. H. Guymon

MSRE Desilgn and Operations Report, Part VII,
Fuel Handling and Processing Plant, by
R. B. Lindauer

MSRE Design and Operations Report, Part VIILL,
Operating Procedures, by R. H. Guymon

MSRE Design and Operations Report, Part IX,
Safety Procedures and Emergency Plans, by
R. H. Guymon

MSRE Design and Operations Report, Part X,
Maintenance Equipment and Procedures, by
E. C. Hise and R, Blumberg

will be the last in the series to be published.
iv

ORNL-TM-911%** MSRE Design and Operations Report, Part XTI,
Test Program, by R. H. Guymon and
P. N. Haubenreich

*¥ MSRE Desgsign and Operations Report, Part XIT,
Lists: Drawings, Specifications, Line
Schedules, Instrument Tabulations (Vol. 1
and 2)
1.

2.

PART

REACTOR SYSTEM
1.1 Fuel and Primary System Materials
Fuel and Coolant Balts ..., v,
Structural Material — INOR=8 ... .iiiiiiinnnnnn

1.1.1
1.1.2
1.1.3
1.1.4
1.2 Bystem
L.2.1
1.2.
1.2.
1.2.
1.2.
1.2.
1.2.
1.2.
1.2.
1.2.10
1.2.11

O 8 -3 &0 U M~ WM

CONTROLS AND INSTRUMENTATION
2.1 Control Rods and Rod Drives

2.2 Safety
2.2.1
2.2.2

2.2.3
2.2.4

CONTENTS

--------------------------------------------------------

---------------------------------------------------

1. DESCRIPTION OF PLANT AND OPERATING PLAN

Moderator Material — Graphite ..................
Compatibility of Salt, Graphite, and INCR-8 ....
COMPONENTS vttt ittt et cn st e it antnranennsnnanss
Reactor Vessel ...ttt iiiiiiiinannnns
Fuel and Coolant Pumps ..veeiiniienianennnacasns
Primary Heat Exchanger .......c.iviiiiintineocanns
Salt-to~Air Radiator ....... i,
Drain and Storage Tanks .ot iiitienotssensennns
Piping and Flanges ...u.eitiiieiiniieniiinanasarass
Freeze Valves ... .ottt iiiiiinneanns
Cover-Gas Supply and Disposal ... eeivinvinrnerss
Sampler-FEnricher ... .ttt riteiecesstesssasns
Electric Heaters ..i. vt ittt iassosnnnes

Tigquid Waste System .. vv ettt ieiinnennans

TN trumMe Nt a iy v vttt s s e vt ot oresennasseconossoss

Nuclear Safety System ... en i

Temperature Instrumentation for Safety

System Inpuls ovivti ittt it i ittt e s
Radiator Door Emergency Closure System .........

Reactor Fill and Drain System ........covveee...

---------------------------------------------

ooooooooooooooooooooo

-------------------------------

Page

 

iii

O ~3 ~1 W

10
12
15
16
19
23
25
28
33
33
34
39
41
43
45
47
55
58

68
70
71
vi

2.2.5 Helium Pressure Measurements in the Fuel
= T P I T o
2.2.6 Afterheat Removal System .....ccuiiiitiiineniann
2.2.7 Containment System Instrumentation .............
2.2.8 Health-Physics Radiation Monitoring ............
2.3 Control Instrumentation .......... i,
2.3.1 Nuclear Instrumentation .......civivieeieinnnne,
2.3.2 Plant Conbrol vueirvirtiiiinnrinrrscnonnsnroanss
2.4 Neutron Source Considerationsg .....veveverierereeeeenen.
2.5 Electrical Power System ....vviiiii it rsensenscnons
2.6 Control Room and Plant Instrumentation Layout .........
2.6.1 Main Control Area ......viiiiiiiirrerininnnenens
2.6.2 Auxiliary Control Area .....viieitereonennneeanns
2.6.3 Transmitter Room ......couiiiiiiiiniiieiininnn,
2.06.4 TField Panels tueeeiet oo nsnesnscsensnssssons
2.6,5 Interconnections .......c.iiiiiinieerinnnnncnnns
2.6.6 Data ROOM v.ivviir it inirieeentnsensneessnsnnens
PLAN D LAY QU it ittt it ettt s enonnnessennsnnssesesensnsas
3.1 Fguipment Arrangement . ......c.eeueniernneeeseroneonnasens
3.2 Biological ShieldiIng . .vveiirinrinetnerorneaneenennnnns
ST E FRATURE S ittt ittt ittt iie i tinnsrisnassonaananens
ol LOCATION vttt i i e e et e e e e e
4.2 Population Density . .vviii ittt it i e
4.3 Geophysical FeatUres .uuiieiiiie it iieernenenennaneanns
b.3.] MetEOrOlOEy vttt e tvat et e
3.2 Temperature .. ...ttt ittt ittt
4.3.3 Precipitation ...iieiiiiiiiiiiii it
L O i o P
4.3.5 Atmospheric Diffusion Characteristics ..........
4,3.6 Environmmental Radicactivity ........ceiiiiiien...
4.3.7 Geology and HydroloZY cueeieeiveecorsosnnoassnnas
e 3.8 el oMol ettt it et et et e

77

79

g1

92

96

96
104
117
118
119
119
119
123
124
124
125
128
128
132
138
138
138
146
146
146
147
149
157
158
158
164
vii

CONSTRUCTION, STARTUP, AND OPERATION ... vivnernrvenneonnns 168
5.1 Construction ittt it it it et e 168
5.2 Flush-Salt Operation .v.eeieiiieeien ittt rnnnienannnnss 168

5.2.1 Critical Experiments ..v.ieeiiiiinreninnnennnanns 170

5.2.2 Power Operation .....ieieeintiennrnenesnaenannes 170
5.3 Operatlons Personnel cu.iieiiineersnrneesoersooesnsnsns 171
5.4 Maintenante .u.euireeier it iteneeeeetaoestassansesoesans 173

PART 2. GSAFETY ANALYSES

O AT M T ettt tee et ennsseesasennereonnsosnnsanssnnnsesns 177
6.1 General Design Considerations ....c.oivieernerrnnnrnnnns 177
6.1.1 Reactor Cell Design ..v.vieriiteinreerenonnoneens 178
6.1.2 Drain Tank Cell DESigll .vv vt ietnnienereaneenns 182
6.1.3 Penetrations and Methods of Sealing ............ 184
6.1.4 Leak Testing voveirrrereerioanernoosonronanossnn 187
6.2 Vapor-Condensing System ... eiiiiertoeroennsoennnonas 187
6.3 Containment Ventilation System ....vven it iniennreas 191
DAMAGE TO PRIMARY CONTAINMENT SYSTEM ..vveerniiinriinnnnnnns 196
7.1 Nuclear Tncidents ...ttt innnsioneenns 196
7.1.1 General Considerations in Reactivity
Incidents .. ittt it ittt et e 196
7.1.2 Uncontrolled Rod Withdrawal ......vivivennveann. 199
7.1.3 "Cold-Slug" Accident ....ieiiiiiiiiieiiiieaaenn, 203
7.1.4 Filling Accidents .o iiiiirninninienernsceranns 205
7.1.5 Fuel AdAitions ...c.iiiiiin i initneiennnnennes 213
7.1.6 U0y Precipitation ....coiiiiiiiinienineninnecas 214
7.1.7 Graphite Loss or Permeation ...........ccvu... 219
7.1.8 Loss Of FlOW ittt iiesiinnnnensnaesnonnnnnns 221
7.1.9 Loss OFf Load «uviivinnnrrnnnrentensanonasnnnsenns 225
7.1.10 Afterheat ... ittt iiiinierionnsarassennenns 226
7.1.11 Criticality in the Drain Tanks ....eceveensrens 230
7.2 Nonnuclear ITncidents ....iiieiiiiiriiniinriinionaereano. 231
7.2.1 Freeze-Valve Fallure ........o.uieiitennvennnnans 231

7.2.2 TFreeze-Flange Failure .......ceiueevenenonnne nn 231
7.2.3 Excessive Wall Temperatures and Stresses .......
7.2.4 COrrOSIOn tuiiiriieeriereeeroeettonssneeeanennns
7.2.5 Material Surveillance Testing ...... et aeeaa
7.3 Detection of Salt Spillage ....... et cen
7.4 Most Probable Accident .............. et
8. DAMAGE TO THE SECONDARY CONTATINER . .vivivinninrnernnnnnnennn
8.1 Missile Damage .....veeeertrennrennonnenas et
8.2 EXCeSSIVE PressUrt v vivevrnrnnrnnrnearesneetoeeoneeenns
8.2.1 Salt Spillage .....iviiriinrnnnnnan e
8.2.2 0il Line Rupture ......... ..., C et
8.3 Acte of Hature ..ot et tie i .o
B8.3.1 EBarthnguUaKe vttt et instneonsenanoeanonennas
S T o o T
S TN Loy v ==
8.5 Corrosion from Spilled Salt v.iiiiieiirrnennnens e
8.6 Maximum Credible Accident ......ovvvvinnn... e e
8.7 Release of Radicactivity from Secondary Container .....
g.7.1 Rupture of Secondary Container ................ .
8.7.2 Release of Activity After Maximum Credible
Accident ... e s i s
8.2 Release of Beryllium from Secondary Container .........
Appendix A. Calculations of Activity Levels ......eeiivinnne...
Appendix B. Process Flowsheels . .i. ittt i ieenneennennens
Appendix C. Component Develcpment Program in Support of
The MORE ittt i i i i it et i e
Appendix D. Calculaticn of Activity Concentrations Resulting
from Most Likelyv Accident ...t ie ittt it inersnnn
Appendlx E. Time Required for Pressure in Containment Vessel

viii

To Be Lowered to Atmospheric Pressure .............

232
233
235
236
236
238
238
238
238
239
239
239
239
240
240
240
245
245
Fig. 1.1
Fig. 1.2
Fig. 1.3
Fig. 1.4
Fig. 1.5
Fig. 1.6
Fig. 1.7
Fig., 1.8
Fig. 1.9
Fig. 1.10
FPig. 1.11
Pig. 1.12
Fig. 1.13
Fig. 1.14
Fig. 1.15
Fig. 1.16
Fig. 1.1%7
Pig. 1.18
Fig. 2.1
Fig. 2.2
Fig. 2.3
Fig. 2.4
Fig. 2.5
Fig. 2.6
Fig. 2.7
Fig. 2.8
Fig. 2.9
Fig. 2.10

ix

LIST OF FIGURES

Fuel and Coolant Flow Diagram

MSRE Graphite Showing Cracks Resulting from Impregna-
tion and Baking Operations

MSRE Layout

Reactor Vessel and Access Nozzle

Typical Graphite Stringer Arrangement

Lattice Arrangement at Control Rods
Fuel-Circulation Pump

Primarj Heat Exchanger

Salt-to-Air Radiator

Fuel Salt Drain Tank

Bayonet Cooling Thimble for Fuel Drain Tanks
Freeze Flange

Freeze Valve

Cover-Gas System Flow Diagram

Off'-Gas System Flow Diagram

MSRE Sampler-Enricher

Single-Line Diagram of MSRE Power System
Simplified Flowsheet of MSRE Liquid Waste System
Control Rod Drive Unit Installed in Reactor
Diagram of Control Rod

Electromechanical Diagram of Control Rod Drive Train

Reactivity Worth of Control Rod as a Function of
Depth of Insertion in Core

Control Rod Height Versus Time During a Scram
Control Red Shock Absorber

Control Rod Thimble

Prototype Control Rod Drive Assembly
Functional Diagram of Safety System

Block Disgram of Safety Instrumentation for Control
Rod Scram

Page

 

13

15
17
18
18
20
23
26
29
31
34
35
37
38
39
42

48
49
50
51

52
54
56
57
59
60
Fig.

Fig.
Fig.
Fig.
Fig.

Fig.
Fig.
Fig.
Fig.
. 2.20
Fig.

Fig.
Fig.

Fig.

Fig.

Fig.
Fig,
Fig.
Fig.
Fig.
. 2.31

Fig.

Fig.
Fig.
Fig.
Fig.
Fig.

2.11

2.12
2.13
2.14
2.15

2.16
2.17
2.18
2.19

2.21

2.22
2.23

2 .24

2.25

2.206
2.27
2.28
2.29
2.30

2.32

2.33
2 .34
2.35
2.36
2.37

Typical Temperature-Measuring Channel Used in Safety
System

Radiator Door Emergency Closure System
Pump-Speed Monitoring System
Reactor Fill and Drain System Valving

Typical Instrumentation for Measuring Helium Pressures
in the Primary Loop

Afterheat Removal System

Helium Supply Block Valving

Off-Gas System Instrumentation and Valving

Lube 0il System Off-Gas Monitors

In-Cell Liquid Waste and Instrument Air Block Valving

Pressure Switch Matrix Used with Instrument Air Line
Block Valves

In-Cell Cooling Watér System Block Valving

Reactor Building (7503) at 852-ft Elevation Showing
Locations of Monitors

Reasctor Building (7503) at 840-ft Elevation Showing
Locations of Monitors

Typical Low-Level BFs Counting Channel for Initial
Critical Testus

Locations of BF3 Chanbers

Wide-Range Ccunting Channel

Linear Power Channels and Automatic Rod Controller
Nuclear Instrumentaticn Penetration

Computer Diagran for Servo-Controller Simulation

Results of Analog Simulatiocn of Sysftem Response to
Step Changes in Power Demand with Reactor on Automa-
tic Temperature Servo Control

Results of Analog Simuistion of Reactor Response to
Ramped Changes in Outlet Temperature Set Point with
Reactor Under Automatic Control

Simplified Master Plant Control Block Diagram
Simplified Rod Control Block Diagram

Regulating Rod Limit Switch Assembly

Regulating Rod Contrcl Circuit

Diagram of Safety System Bypassing with Jumper Board

€9

72
73
74
78

g0
83
84
85
88
89

91
93

94

96

07
)
99
190
103
105

105

106
109
110
111
116
Fig.
Pig.
Pig.
Fig.,
Fig.
Fig.
Fig.
Fig.
Fig.
'ig.
Fig.
Fig.
Fig.
Fig.
Fig.

Fig.
Fig.
Fig.

Fig.

Fig.
Fig.
Fig.
Fig.
Fig.

Fig.

Pig.

Fig.

Fig.

N~

MMM DD W W WD NNDNNN NN DD

oo W W

I O O O O
O W oo e

xi

Main Floor Layout of Building 7503 at 852-ft Elevation
Main and Auxiliary Control Areas

Main Control Board

Layout of Building 7503 at 840-ft Elevation
Transmitter Room

Layout of Data Room

Typical Process Computer System (TRW-340)

First Floor Plan of Regctor Building

Elevation Drawing of Reactor Building

Arrangement of Shielding Blocks on Top of Reactor
Area Surrounding MSRE Site

Contour Map of Area Surrounding MSRE Site

Map of Oak Ridge Area

beasonal Temperature Gradient Frequency

Annval Frequency Distribution of Winds in the Vicinity
of X-10 Aresa

X-10 Ares Seasonal Wind Roses
X-10 Area Seasonal Wind Roses

Wind Roses at Knoxville and Nashville for Various
Altitudes

Reactor Division, Operations Department Organization
for the MSRE

Reactor Cell Model

Drain Tank Cell Model

Typical Flectric Lead Penetration of Reactor Cell Wall
Vapor-Condensing System

Power and Temperature Transients Produced by Uncon-
trolled Rod Withdrawal in Reactor Operating with Fuel B

Power and Temperature Transients Produced by Uncon-
trolled Rod Withdrawal in Reactor Operating with Fuel C

Effect of Dropping Two Control Rods at 15 Mw During
Uncontrolled Rod Withdrawal in Reactor Operating with
Fuel C

Power and Temperature Transients Following InJjection
of Fuel B at 900°F into Core at 1200°F; No Corrective
Action Taken

System Used in Filling Fuel Loop

120
120
121
122
123
126
127
122
130
134
139
140
141
148
150

152
153
156

172
179
183
185

190
201

202

204

206

209
Fig.

Fig.

Fig.

Fig.

Fig.

Fig,

Fig.

Fig.

Fig.
Fig. 8

Fig.

Fig.

Pig.,

Fig.

Fig.

Fig.

Fig.

3

.10

A1

.12

13

Xii

Net Reactivity Addition During Most Severe Filling
Accident

Power and Temperature Transients Following Most Severe
Filling Accident

Effects of Deposited Uranium on Afterheat, Graphite
Temperature, and Core Reactivity

Power and Temperatures Following Fuel Pump Failure,
with No Corrective Action

Power and Temperastures Following Fuel Pump Power
Failure. Radiator doors closed and control rods
driven in after failure.

Bffects of Afterheat in Reactor Vessel Filled with
Fuel Salt After Operation for 1000 hr at 10 Mw

Temperature Rise of Fuel in Drain Tank Beginning 15
min After Reactor Operation for 1000 hr at 10 Mw

Heatlng of Reactor Vessel Lower Head by UOp Deposits
with Power Level at 10 Mw

Release of Fuel Through Severed Lines

Relationship Between Cell Pressure and Weights of
Fluids Equilibrated

Radiation Level in Bullding Following Maximum Credible
Accident (r/hr = 935 x upc/cm® x Mev)

Activity Concentrations in Building Air Following Maxi-
mum Credible Accident

Noble Gas Activity After Maximum Credible Accildent
as Function of Distance Downwind

Change in Rave of Activity Release from Building
with Time

Total Integrated Doses Following Maximum Credible
Accident

Peak Todine and Solids Activities After Maximum
Credible Acclident as Function of Distance Downwind

Beryllium Contaminaticn After Maximum Credible Acci-
dent as Functicon of Distance Downwind

211

212

218

222

223

225

227

229

2472
243

2417

248

249

251

252
MSRE DESIGN AND OPERATIONS REPORT

Part V

REACTOR SAFETY ANALYSIS REPORT

S. E. Beall R. B. Lindauer
P. N. Haubenreich J. R. Tallackson
INTRODUCTION

The Oak Ridge Natilonal Laboratory undertoock, in 1951, the develop-
ment of a molten-salt-fueled reactor for the Aircraft Nuclear Propulsion
Program, and the Aircraft Reactor Experiment (ARE) was successfully op-
eragted, in 1954, to demonstrate the nuclear feasibility of a system in
which molten-salt fuel was circulated.? Development work on molten-salt-
fueled systems has been continued, with the ultimate objective of design-
ing a molten-salt reactor capable of breeding.

The Molten-Salt Reactor Experiment (MSRE)?,? is presently being as-
sembled as an engineering demonstration of a single-region, nonbreeding
salt-fueled reactor. The goals in the development of this system have
been to show that this advanced-concept reactor is practical with present-
day knowledge, that the system can be operated safely and reliably, and
that is can be serviced without unusual difficulty.

It is the purpose of this report to describe the MSRE (Part I) and

to evaluate the hazards associated with this reactor (Part II).

 

W. B. Cottrell et al., "Operation of the Aircraft Reactor Experi-
ment," USAEC Report ORNL-1845, Oak Ridge National Laboratory, August
1955,

?0ak Ridge National Laboratory, "Molten Salt Reactor Experiment
Preliminary Hazards Report,'" USAEC Report ORNL CF 61-2-46, Feb. 28, 1961;
Addendum, Aug. 14, 1961; Addendum No. 2, May 8, 1962.

3A. L. Boch, E. S. Bettis, and W. B. McDonald, ."The Molten Salt Re-
actor Experiment," paper prepared for presentation at the International
Atomic Fnergy Agency Symposium, Vienna, Austria, October 23-27, 1961.
PART 1. DESCRIPTION OF PLANT AND OPERATING PLAN
1. REACTOR SYSTEM

The Molten-5alt Reactor Experiment comprises a circulating-fuel re-
actor designéd for a heat generation rate of 10 Mw and the auxiliary
equipment necessary for 1ts operation. The fuel circuit and the heat re-
moval, or coolant, circuit are illustrated schematically in Fig. 1.1,%
which also gives some standard operating conditions for the circuits.

The fuel salt, a mixture of LiF, BeF., ZrF,, and UF,, is pumped
through a cylindrical reactor vessel filled with graphite blocks. At
the 10-Mw power level, the fuel enters the vessel at 1175°F and is heated
to 1225°F. It then flows to a 1200-gpm sump-type centrifugal pump, which
discharges through the shell side of the fuel-to-coolant heat exchanger.
The fuel returns from the heat exchanger to the reactor inlet.

The coolant salt, a mixture of LiF and BeF,;, is heated from 1025 to
1100°F as it passes through the tubes of the fuel-to-coolant heat ex-
changer. Following dissipation of the heat in an air-cocled radiator,
the coolant salt is returned to the heat exchanger at a flow rate of
850 gpm by a second sump-type centrifugal pump.

Electric heaters on the piping and equipment keep the salt mixtures
above their melting points. In order to assure that the salt mixtures
remain molten, the normal electric utility supply is augmented by an
emergency supply from the diesel generators. The molten salts may be
drained from the reactor system within a few minutes.

The fission gases xenon and krypton are stripped from the fuel salt
continuously in the gas space above the liquid surface within the fuel
pump bowl. A continuous flow (~3 liters/min) of helium removes the fis-
sion gases from the pump bowl and carries them to activated-carbon beds
outside the reactor cell. Helium is also used as the cover gas through-
out the fuel and coclant systems.

The components of the fuel and coolant systems are constructed of
INOR-8, a nickel-molybdenum~chromium alloy developed especially to con-
tain molten fluoride salts. In the fuel system the piping to the

 

*Detailed flowsheets for all the reactor systems are presented in
Appendix B.
UNCLASSIFIED
ORNL-LR-DWG. 56870R1

COOLANT
PUMP

  

850 gpm

  
 
    
  
 
   
 

 

 

 

] I
II FREEZE ”
FLANGE H “
H 1200 gpm ” J COOEEEI ]]
I ’
“ REACTOR Il J‘ I
I VESSEL I i
|[ REACTOR CELL “ RADIATOR H
! w H
e = e — e — — — | 300°F n
AIR
LT . - —— 200,000 cfm
I oraIN - 7[ 100°F I
i | , I

I ) |

| )

COOLANT

DRAIN TANK
(44 cu f1)

 

SPARE FILL AND FLUSH
FILL AND DRAIN TANK TANK
DRAIN TANK (73 cu ft) {73 cu ft)

(73 cu ft)

Fig. 1.1. Fuel and Coolant Flow Diagram.

components is connected by "freeze" flanges which utilize frozen salt as
the sealant. These flanges are also provided with gas-buffered ring-
Jjoint gaskets. The use of freeze-flange joints facilitates removal and
replacement of radiocactive equipment after power operation. Joints of
this type are not needed in the coclant system, because the radiocactivity
there is low enough to permit direct maintenance within a few minutes after
shutdown.

No valves of the ordinary type are used in contact with salt. Flow
in lines connecting the reactor vessel and the drain tanks is prevented

by freezing salt in designated sections of pipe. The "freeze valves" thus
formed can be thawed by stopping the flow of cooling alr and heating the
pipe. By this means, salt can be drained from the reactor vessel,.

In addition to the fuel and coolant circuits, there are auxiliary
components, such as controls and instrumentation, salt-sampling equipment,
facilities for handling radioactive liquid wastes, a chemical processing
cell for purification of the salts, and remote-maintenance equipment.
Details of the reactor components, auxiliary equipment, and instrumenta-
tion are discussed in the following sections.

The control problems of a molten-salt reactor differ in many respects
from those of solid fuel reactors‘(see sec. 1.3). The reactor core is
provided with three control rods that will be used principally to main-
tain the fuel salt temperature within the operating limits. The reactor
power level will be controlled by regulating the rate of heat removal at

the radiator.

1.1 Fuel and Primary System Materials

 

1.1.1 Fuel and Coclant Salts

 

The MSRE fuel will be a solution of U??°F, and ZrF, in a molten
Li7F-BeF, solvent. Thorium tetrafluoride may be added as a fertile
material. Both Li’F and BeFy have relatively good neutron cross sections;
the ratio of these materials was chosen to provide the optimum compromise
among freezing point, fluid flow properties, and heat transfer character-
istics of the final mixtures in which UF, and ZrF, concentrations are
firmly fixed. The ZrF, is included in the fuel mixture because oxygenated
species (i.e., Ho0) capable of yielding oxide ions on contact with the
fuel precipitate ZrO, from LiF-BelFj,-~-ZrF,-UF, solutions whose Zrt ato-Utt
ratio exceeds 3; U0y precipitates if oxide ion is admitted to solutions
in which the ratio is appreciably below that value. To prevent precipi-
tation of U0z upon inadvertent contamination of the reactor system, the
Zr**-to-U** ratio is fixed at a conservative value (>5:1). Moreover,
since precipitation of ZrO; is also undesirable, the fuel mixture is pro-
tected at all times from gases and vapors bearing oxygenated species by

using helium as a blanket gas.
The compositions and properties of three fuel salts are listed in
Table 1.1. EFach of these salt mixtures is expected to be employed at
some stage of the experiment. Salt C, the partially enriched salt, will
be loaded for the first series of experiments because its chemical char-
acteristics have been studied more thoroughly than those of salts A and
B. Salt B, the highly enriched salt, is the composition which is pro-
posed for the core of s large two-region breeder reactor (or a U23°
burner). It will be used after a long period of operation with the par-
tially enriched salt. Salt A contains thorium and has been proposed for
single-region molten-salt breeder reactors. It will be tested in a third
series of MSRE experiments.

The coolant salt is a mixture of Li”F and BeF, whose composition and
properties are included in Table 1.1. The coolant mixture can withstand
higher oxide contaminant levels than the fuel before an oxide (BeOQ in
this case) precipitates, but the same practice of blanketing the mixture
with helium will be folilowed.

Another salt mixture of essentially the same composition ag tThe

coolant salt, but called the flush salt, will be used to rinse the fuel

Teble 1.1. Compositions and Physical Properties of the
Fuel, Flusk, and Coolant Salts

 

uel Salt B, Fuel Salt C,

 

F?el Salu-A, with 02% with 35% Flush and
with Thorium . . Coolant
and Uraniunm Enrlghed Enrlghed Salt
Uranium Uranium
Composition, mole %
LiF (99.99+% Li") 70 66. 8 65 66
BeF», R2. 6 29 29.1 34
ZrFy 5 %4 5 0
ThF, 1 0 0 0
UF, 0.4 0.2 0.9 0
Physical properties at 1200°F
Density, 1b/ft? 140 130 13 120
Viscosity, 1b/ft-hr 18 17 20 248
Heat capacity, Btu/lb.°F 0.45 0.4 0.47 0.53
Thermal conductivity, 3.2 3.2 3.2 3.58
Btu/hr- 2 (*F/ft)
Ligquidus temperature, °F 840 840 840 850

 

%At 1060°F.
system of fuel salt and fission-product residues after shutdown and before
maintenance is begun. If the maintenance operations allow the inner sur-
faces of the fuel system to be exposed to air, the flush salt will also

be circulated after maintenance as an 0p-HpO cleanup measure. This should
protect the fuel salt from oxygen contamination. However, if any of the
salt mixtures become excessively contaminated with oxygen, they will be

purified by treatment with an Hy-HF mixture (see sec. 2).

1.1.2 Structural Material — INCR-8

 

The principal material of construction for the reactor system is
INOR-8, a nickel-molybdenum-chromium alloy developed at the Oak Ridge
National Laboratory for use with fluoride salts at high temperature.l

The composition and properties of INOR-8 are listed in Tables 1.2, 1.3,

Table 1.2. Composition of INCR-8

 

 

. Quantity . Quantity
Constituent (vt %) Constituent (vt %)

i 66~71. Mn 1.0 (max)
Mo 15-18 Si 1.0 {(max)
Cr 6-8 Cu 0.35 (max)
Fe 5 (max) B 0.010 (max)
C 0.04-0.08 W 0.50 (max)
Ti + Al 0.50 (max) P 0.015 (max)
S 0.02 (max) Co 0.20 (max)

 

 

and 1l.4. When INOR-8 is corrosively attacked, chromium is leached from
it and tiny subsurface voids are formed. The rate of attack is governed
by the rate of diffusion of chromium in the alloy. Measured rates of
attack in typical fuel and coolant salts for many thousands of hours have
been less than 1 mil/yr. Furthermore, no greater attack has been observed
in several thousand hours of in-pile tests (see sec. 7.2.4). The perfor-

mance of INOR-8 in the MSRE will be followed by removing surveillance

 

IR, W. Swindeman, "The Mechanical Properties of INOR-8," USAEC Re-
port ORNL-2780, Oak Ridge National Laboratory, January 1961.
10

Table 1.3. Physical Properties of INOR-8

 

Density, 1b/in.> 0.320
Melting point, °F 2470-2555
Thermal conductivity, Btu/hr-ft? (°F/ft) 12.7

at 1300°F
Modulus of elasticity at ~1300°F, psi 24.8 x 10°
Specific heat, Btu/lb-°F at 1300°F 0.138
Mean coefficient of thermal expansion in 8.0 x 107°

70-1300°F range, in./in.-°F

 

Table 1.4. Mechanical Properties of INOR-8

 

 

Ultimate Yield Maximum

Temperature Tensile Strength at Allowable

(°F) Strength 0.2% Offset Stress?
(psi) (psi) (psi)
ok 82,000 26,200 17,000
1100 77,000 25,500 13,000
1280 62,500 24,200 6,000
1300 58,000 24,200 3,500

 

aASME Boiler and Pressure Vessel Code Case
131¢%.

specimens, at 3- to 6-rmonth intervals, from the sample assembly located

in the spare control rod position of the reactor vessel (see sec. 1.2.1).

1.1.3 Modergtor Material — Graphite

 

Although the salt has moderating properties, the use of a separate
moderator has the advantage of reducing the reactor fuel inventory. Un-
clad graphite (see Table 1.5 for properties) was chosen as the MSRE moder-
ator to avoid the problems of cladding, such as neutron losses and de-
velopment of cladding technigues, and because no serious problems were
foreseen with the bare graphite. At The time the decision was made,
graphite that could meet the requirements cf high density and low per-

meability to salt had been manufactured on an experimental basis by the
11

Table 1.5. Properties of MSRE Core Graphite, Grade CCB

 

Physical prcperties

Bulk density,® g/cm?
Range
Average
Porosity,? %
Accessible
Inaccessible
Total
Thermal conductivity,® Btu/ft.hr.°F
With grain
At 90°F
At 1200°F
Nermal to grain
At 90°F
At 1200°F
Temperature coefficient of expansion,
With grain, at G8°F
Normal to grain, at 68°F
gpecific heat,d Btu/lb. °F
At O°F
At 200°F
At 600°F
At 1000°F
At 1200°F
Matrix ceoefficient of permeability to helium at
70°F,€ cm? /sec
Salt absorption at 150 peig,? vol %

Mechsnical strength at 68 °FD

C OF—l

Tensile strength, psi
With grain
Range
Average
Wormal to grain
Range
Average
Flexural gstrength, psi
With grain
Range
Average
Normal to grain
Range
Average
Modulus of elasticity, psi
With grain
Normal to grain
Compressive strength,d psi

1.82-1,¢87
1.86

.49 x 10-°
.8 x 10-°

N O

.14
.22
.33
.39
A2
% 104

WOOOOO0o

O

.20

1500=6200
1200

1100-4500
1400

3000-5000
4300

2200~3650
3400

3.2 x 106
1.0 x 108
8600

 

"Messurements made by the Oak Ridge National Laboratory.

b
and the Oak Ridge National Laboratory.

Based on measurements made by the Carbon Products Division

cMéasurements made by the Carbon Products Division.

d
Representative data from the Carbon Products Division.

e
Based on measurements made by the Carbon Products Division

on pilot-production MSRE graphite,
12

National Carbon Company. Since it appeared that bars 2 in. by 2 in. could
be processed without difficulty, this size was selected as the basic ele-
ment for the MBRE core assembly.

During the manufacture of the MSRE graphite, it was learned that scme
cracking resulted from the impregnation and baking operations and that
the density was not quite as high as had been expected. The cracks raised
guestions about the strength and the permeability. The cracks in some
of the poorest material are shown in Fig. 1.2. After extensive examing-
tion and testing2 at the Oak Ridge National Laboratory, it was determined
that the strength was as great or greater than specified, that the re-
duction in density was inconsequential, that the cracks could not be
propagated by mechanical forces or thermal stresses far in excess of con-
ditions which will exist in the reactor, and that the total penetration
of salt in the cracked graphite would be considerably below the specified
penetration (0.5% at 150 psig). Furthermore, two to three months of in-
pile testing of AGOT-grade graphite (a more permeable material with 12%
salt penetration) with fuel salt at power densities more than six times
the MSRE power density produced no observable damage to the graphite.
Therefore, after consultation with representatives of the AEC Division
of Reactor Development and the manufacturer, the graphite was accepted

for use in the MSRE.

1.1.4 Compatibility of Salt, Graphite, and INOR-&

 

Out-of-pile tests of combinations of fluoride salt, graphite, and
INOR-8 over a periocd of several years have convincingly demonstrated the
compatibility of these materials.? However, in-pile testing of fuel and
graphite in INOR-8 capsules under conditions similar to those anticipated

in the MSRE has been accomplished only since 1961. Although these first

 

20ak Ridge National Laboratory, "MSRP Semiann. Prog. Rep. Jan. 31,
1963," USARC Report ORNL-3419, pp. 70-76, and "MSRP Semiann. Prog. Rep.
July 31, 1963," USAEC Report ORNL-3529, Chapter 4.

W. D. Manly et al., "Metallurgical Problems in Molten Fluoride Sys-
tems," Progress in Nuclear Energy, Series IV, Vol. 2, 164~179, Pergamon
Press, London, 1960.

 
13

‘Unclassified
. Photo ¥-49020

 

 

. Unclassified
__Photo ¥-49021

«}

 

 

 

 

| Fig. 1.2. MSRE Gréfihite Shcnnng Cracks Resulting from Impregnation
and Baking Operationms. o - , ,

W'}

 

 
14

irradiation experiments4 showed absolutely no evidence of attack on the
INOR-8 or wetting of the graphite by the salt, appreciable duantities of
Fo, and CF, were observed. Furthermore, in some of the capsules, Xenon
was not found in the quantities expected.

These anomalies were studied in two later series of in-pile capsule
experiments during a total of seven months of exposure in the MIR. The
following conclusions have been reached after a thorough analysis of in-
formation accumulated during and after these experiments. Neither Fp nor
significant quantities of CF, is generated when fission occurs in the
molten salt at power densities as high as 65 w/cm®. However, Fp is re-
leased from the irradiated salt after 36 or more hours of exposure to
high-level beta and gamma fluxes at temperatures below 200°F. It also
appears that CF, can be formed if Fp is present and avallable to the
graphite when radiation is present at low temperatures or when the system
temperature is raised to 1200 to 1400°F. Undoubtedly, the fluorine would
also react with the INOR-& at these high temperatures. Finally, it must
be concluded thas the "disappearance” of xenon in the tests resulted from
a reaction of the fluorine to form one or more of the solid xenon fiuo-
rides.

These conclusions suppor: the important additional judgment that Fa
and CF, formation should not affect the success of the MSRE. First, the
salt in the MSRE should never be in the frozen state, except for small
guantities at isolated locatiocns (e.g., freeze flanges and freeze valves).
Even these isolated deposits would not be at temperatures as low as 200°F.
Second, if Fp were evolved from the frozen seals, the rate of reaction
with the INOR-8 metal at 200°F would be insignificant; at higher tempera-
tures the back reaction would reccmbine the Fp and the reduced salt with-
out an observable effect.

Samples of INOR and graphite are located at a position of maximum

flux within the reactor vessel to allow examination and testing after

 

“0ak Ridge National Laboratory, "MSRP Semiann. Prog. Rep. Jan. 31,
1963," USAEC Report ORNL-3419, pp. 8§0-107, and "MSRP Semiann. Prog. Rep.
July 31, 1963," USAEC Report ORNL-3529, Chapter 4.
15

various periods of exposure to salt. Salt samples for detaliled chemical

gnalysis will be removed and will be analyzed daily.

1.2 System Components

 

The components of the reactor are arranged in the fuel and coolant
salt systems as depicted in Fig. 1.3. The individual pieces of equipment
are described below. All the components are designed to meet Section VIII
of the ASME Boiler and Pressure Vessel Code except that no pressure-relief
devices are provided on the primary system. Instead, supply pressures
(for example, cover gas) are limlted by control and protective devices.

UNCLASSIFIED
ORNL-DWG 63-1209R

N i

"% REMCTE MAINTENANCE f |

-

o .+ CONTROL ROOM : : ‘[

 
  
   

REAGTOR CONTROL [k““Am
ROOM., o

  

\ I, REACTOR VESSEL 7. RADIATOR

S 2. HEAT EXCHANGER 8. GOOLANT DRAIN TANK
U 3. FUEL PUMP 9. FANS
4 FREEZE FLANGE 10. FUEL DRAIN TANKS
Y T 5. THERMAL SHIELD 11. FLUSH TANK
AN T 6 COOLANT PUMP 12. CONTAINMENT VESSEL

13. FREEZE VALVE

Fig. 1.3. MSRE Layout.
16

1.2.1 Reactor Vessel

The reactor consists of a cylindrical vessel approximately 5 ft in
diameter and 7 1/2 £t high, fitted with an inner cylinder that forms the
inner wall of the shell-cooling annulus and serves to support the 55-in.-
diam by 64-in.-high graphite matrix with its positioning and supporting
members. Figure 1.4 is an assembly drawing of the reactor vessel and its
graphite core, Fluid enters the vessel at the top of the cylindrical sec-
tion and flows downward in a spiral path along the wall. With the design
flow of 1200 gpm in the 1l-in. annulus, the Reynolds modulus is about 22,000.
At the estimated heat generation rate of 0.24 w/cm3 in the wall, 28 kw of
heat 1s removed by the incoming salt, while maintaining the wall tempera-
ture at less than 5°F above the bulk fluid temperature. Design data for
the reactor vessel are listed in Table 1,6,

The fuel loses its rotational motion in the lower plenum and then
flows upward through the graphite core matrix, which constitutes about
77.5% of the core volume. The moderator matrix is constructed of 2- by
2- by 63-in, stringers of graphite, which are loosely pinned to restrain-
ing beams at the bottom of the ccre.

Flow passages in the graphite matrix are 0.4- by 1.2-in. rectangular
channels, with rounded corners, machined into the faces of the stringers.
A typical arrangement of graphite stringers is shown in Fig. 1.5. Flow
through the core is laminar, but because of the good thermal properties
of the graphite and the fuel, the graphite temperature at the midpoint
is only 60°F above the fuel mixed-mean temperature at the center of the
core. The average power density in the fuel is 14 kw/liter and the maxi-
mum is 31 kw/liter.

The salt leaves the vessel at the top through & 10-in. pipe with a
5-in.-diam side outlet connecting to the circulating pump. A course screen
is provided here to catch pieces of graphite larger than 1/8 in, in diame-
ter should they be present in the salt stream. The 10-in. pipe accommo-
dates an assembly of three 2-in.-diam control rod thimbles and permits
access to a group of graphite and INOR-8 surveillance specimens arranged
in the center of the core as indicated in Fig. 1l.6. The assembly 1is

flanged for removal, if necessary, and is provided with cooling air to
17

UNCLASSIFIED
ORNL-LR- DWG 61097 /1

FLEXIBLE CONDUIT TO
fq,,. /CONTROL ROD DRIVES
GRAPHITE SAMPLE ACCESS PORT f}p{;

  
 
  

COOLING AIR LINES

     
   
   

CORE
CENTERING GRID

e

/é//rFLOW DISTRIBUTOR

FUEL INLET

REACTOR CORE CAN

REACTOR VESSEL

 
 
  

e RS T

e

ANTI-SWIRL VANES A X< MODERATOR
VESSEL DRAIN LINE A SUPPORT GRID

Fig. 1.4. Reactor Vessel and Access Nozzle.
18

SSSSSSSSSSSS

     
  
    
  
 

PLAN VIEW

TYPICAL MODERATOR
STRINGERS

REMOVABLE
STRINGER

 

NOTE:
NOT TO SCALE

 

 

 

 

 

Fig. 1.5. Typical Graphite Stringer Arrangement.

SSSSSSSSSSSS
RRRRRRRRRRRRRRRRRR

..-CONTROL ROD

@by S\ &
N \\/ 0 § GUIDE TUBE
\\\\Q KJ/\\ GUIDE BAR

 

TYPICAL

X 7
NV AN

N ©
\| NN ...

\

0.200-in. R/
0.400 S R = REMOVABLE
R STRINGER
4 GRAPHITE IRRADIATION 274 R N
SAMPLES (7g-in.DIA)} 7 ‘
-
2 in. TYPICAL

Fig. 1.6. Lattice Arrangement at Control Rods.
19

Table 1.6. Reactor Vessel and Core Design Data and Dimensions

 

Construction material INOR-8
Inlet nozzle 5 in., sched-40, IPS
Outlet nozzle 5 in., sched-40, IPS
Core vessel
Outside diameter 59 1/8 in. (60 in. max)
Inside diameter 58 in.
Wall thickness 9/16 1in.
Overall height (to centerline of 100 3/4 in.
5-in. nozzle)
Head thickness L in.
Design pressure 50 psi
Design temperature 1300°F
Fuel inlet temperature 1175°F
Fuel outlet temperature 1225°F
Cocling annulus
Inside diameter 56 in.
Qutside diameter 58 in,

Graphite core matrix

Diameter 55 1/4 in,
Number of fuel channels 1140
Fuel channel size 1.2 x 0.4 in. with

rounded cormers

Core inner container

Inside diameter 55 1/2 in.
Outside diameter 56 in.
Wall thickness 1/4 in.
Height 68 in.

 

permit freezing the salt in the gap below the flange. The control rods

and their drive mechanisms are described in Section 1.3.

1.2.2 Fuel and Coolant Pumps

The fuel-circulation pump is a sump-type centrifugal pump with a
vertical shaft. It has a 75-hp motor and is capable of circulating 1200
gpm of salt against a head of 48.5 ft. TFigure 1.7 is a drawing of the
pump, and design data are presented in Table 1.7.

The pump assenbly consists of motor and housing; bearing, shaft, and

impeller assembly; and a 36-in.-diam sump tank. The sump tank, which is
 

_ SHAFT SEAL—
" (See Inset)

LEAK
DETECTOR

LUBE OIL IN-

BALL BEARINGS
{Face to Face)

BALL BEARINGS ~,__
{ Back to Back) ‘

 

LUBE OfL OUT._

SEAL OIL LEAKAGE &>
DRAIN — -

LEAK DETECTOR-"

SAMPLER ENRICHER -
{Qut of Section)
. {(See Inset)

BUBBLE TYPE»//;«"
LEVEL INDICATOR #

 

OPERATING - I
LEVEL " |

To Overflow Tank

.f"v P
P

 
  

   
 
 
 

| —
,J] l___4 -
G BN
SHAFT Cheom == = 3 NWATER
COUPLING . Mo T T T e 2D COOLED
T vy 1 : MOTOR

 

 

A M~
1 4
‘ '4' s
7 - o

4 T
i .

UNCLASSIFIED
ORNL- LR -DWG-56043-B

 

 

_»LUBE OIL BREATHER

1{.@.__” ~BEARING HOUSING

7 __GAS PURGE IN
A - "
' ’T" .~ SHAFT SEAL (See Inset ) —

|
~~—SHIELD COOLANT PASSAGES

{in Parallel With Lube Oil )}
SHIELD PLUG

   
  

GAS FILLED EXPANSION
SPACE

XENON STRIPPER
{Spray Ring)

>+ SPRAY

 

 

 

 

 

Fig. 1.7. Fuel-Circulation Pump.

 

 

 

 

 

/GAS PURGE OQUT ( See Inset)—

 

oc¢
21

Table 1.7. Design Data for Fuel and Coolant Pumps

 

 

Fuel Coolant
Design flow, gpm

Circulation 1200 850

Bypass (50-gpm spray, fuel pump only) 65 15
Head at 1200 gpm, £t (fuel pump) 48.5
Head at 850 gpm, £t (coolant pump) 78
Motor, hp 75 75
Speed, rpm 1160 1750
Intake, INOR-8, sched-40 IPS, in. 8 6

Outside diameter, in. 8.625 6.625

Wall thickness, in. 0.322 0.280
Discharge nozzle, INOR-8, sched-40 IPS, in. 5 5

Outside diameter, in. 5.563 5.563

Wall thickness, in. 0.258 0.258
Pump bowl, INOR-8

Diameter, in. 36 36

Height, in. 15 15
Volumes, £t

Minimum starting and normal operating volume 4.l 4.1

(including volute)

Maximum operating volume 5.2 5.2

Maximum emergency volume (includes space 6.1 6.1

above vent)

Normal gas volume 2.0 2.0
Overall height of pump and motor assembly, It 8.6 8.6
Design pressure, psi 50 50
Design temperature, °F 1300 1300

 

welded into the reactor system piping, serves as the expansion tank for
the fuel and as a place for the separation of gaseous fission products.
Separation is accomplished by spraying a 50-gpm stream of salt through

the atmosphere of helium In the sump tank. The bearing houging is flanged
to the sump tank so that the rotating parts can be removed and replaced.

The motor is loosely coupled to the pump shaft, and the motor housing is
22

flanged to the upper end of the bearing housing to permit separate removal
cf the motor.

The pump is equipped with ball bearings, which are lubricated and
cooled with oil circulated by an external pumping system. The oil is
confined to the bearing housing by mechanical shaft seals. Helium is
circulated into a labyrinth between The lower bearing and the sump tank.
Part of the gas passes through the lower seal chamber to remove oil vapors
which might lesk through the seal. The remainder of the helium flows
downward along the shaft to prevent radioactive gas from reaching the
oll chamber.

Bubbler tubes are provided in the pump bowl to measure the liquid
level so that the pump will not be overfilled and to permit a determina-
tion of the salt inventory. The salt-sampling device ig attached to the
bowl at another opening to allow the removal of approximately 10~-g por-
tiong of salt for chemical analysis or to add enriched uranium salt in
quantities of 150 g (20 g of U).

Massive metal sections are incorporated in the pump assembly as
shielding for the lubricant and the motor. The motor is enclosed and
sealed to prevent the escape of radiocactive gas or fluids that might leak
through the pump assembly under unusual conditions. Water cooling coils
are attached to the housing to remove heat generated by the motor.

Tmmediately underncash the purp is a torus-shaped tank (5.5 £t2 )
which serves as an emergency overflow tank for collecting the fuel in
the event of coverfilling of the purp or expansion of the fuel in the pump
as a result of an excessive salt temperature. The overflow tank is vented
to the offgas system and is egquipped with level indicators. ©5Should salt
flow into the tank, 1t can be pressurized back into the pump bowl. The
pump bowl and the overflow tank are enclosed in an electrically heated
furnace.

The same type of pump, without the overflow provision, is used in
the coolant system. This pump, however, dces not require asg much pro-
tection against radiaticn. It is driven by a 75-hp motor and is designed
to circulate 850 gpm of salt against a head of 78 ft of fluid. The com-
plete design data for the coolant pump are included in Table 1.7.
23

1.2.3 Primary Heat Exchanger

The primary heat exchanger (Fig. 1.8) contains 159 tubes (1/2 in.
0D, 0.042-in. wall) and is designed to transfer 10 Mw of heat from the
fuel salt (in the shell) to the coolant salt {in the tubes). The ex-
changer has a conventional, cross-baffled, shell-and-tube configuration.
The tube bundle is laced with metal strips (not shown) to prevent vibra-
tion of the tubes. Design data are listed in Table 1.8&.

Space limitations in the reactor cell require a short unit. The
U-tube configuration makes possible a length of & ft without greatly re-

ducing the efficiency of heat transfer, as compared with a straight

UNCLASSIFIED
ORNL-LR-DWG 52036R2

FUEL INLET

  
     
    
    

i/2-in.-0D TUBES

THERMAL-BARRIER PLATE CROSS BAFFLES

TUBE SHEET
COCLANT INLET

16.4-in. OD x 0.2-in. WALL x 8-ft LONG

COOLANT-STREAM

T T COOLANT OUTLET
SEPARATING BAFFLE fl

FUEL OUTLET

Fig. 1.8. Primary Heat Exchanger.
24

Table 1.8. Design Data for Primary Heat Exchanger

 

Construction material INCR-8
Heat load 10 Mw
Shell-side fluid Fuel salt
Tube-side fluid Coolant salt
Layout 25%-cut cross-baffled shell
with U tubes
Baffle pitch 12 in.
Tube pitch G.775 in., triangular
Active shell length ~6 1t
Overall shell length ~8 ft
Shell diameter 16 in.
Shell thickness 1/2 in.
Average tube length 14 £t
Number of U tubes 159
Tube size 1/2 in. OD; 0.042-in. wall
Effective heat transfer surface 259 1%
Tube sheet thickness 1 1/2 in.
Fuel salt holdup 6.1 ft°
Design temperature
Shell side 1300°F
Tube side 1300°F

Design pressure

Shell side 55 psig

Tube side 90 psig
Allowable working pressure®

Shell side 75 psig

Tube side 125 psig
Hydrostatic test pressure

Shell side 800 psig

Tube side 1335 psig

 

aBased on actual thicknesses of materials and stresses al-
lowed by ASME Code.
25

Table 1.8 (continued)

 

Terminal temperatures

Fuel salt 1225°F, inlet; 1175°F, outlet
Coolant 1025°F, inlet; 1100°F, outlet
Effective log mean temperature 133°F
difference

Pressure drop

Shell side 28 psi

Tube side 27 psi
Nozzles

Shell 5 in., sched-40

Tube 5 in., sched-40
Fuel-salt flow rate 1200 gpm
Coolant-salt flow rate 850 gpm

 

counter-flow unit, and eliminates a thermal expansion problem. The tubes
are welded and back brazed to the tube sheet in order to reduce the prob-
abllity of leakage between the fuel and coolant, The coolant pressure is
kept higher than the fuel pressure to reduce the likelihood of fuel out-

leakage in case of a tube failure.

1.2.4 Salt-to-Air Radiastor

The thermal energy of the reactor 1s transferred to the atmosphere
at a salt-to-air radiator, which is cooled by two 100,000-cfm blowers.
The radiator contains 120 tubes (3/4 in. OD, 0.072-in. wall, 30 ft long)
and is assembled as shown in Fig. 1.9. Design data for the radiator are
listed in Table 1.9.

Several features were incorporated in the design as protection against
freezing of coolant salt in the radiator:

1. The tubes are of large diameter.

2. The heat-removal rate per unit area is kept low by using tubes
without fins so that most of the temperature drop is in the air fiim.

3. The nminimum salt temperature is kept 175°F above the freezing
point (850°F). A thermocouple is attached to each tube for temperature

monitoring.
 

   
 
     
 
   
  
    
  
   
   
   
    
 
 
   
    
  

        

/'/ DRI
7 S

DOOR DRIVE MOTO
GEAR REDUCE

:‘5
‘|‘ ;'1
1t
SUPPORTING STEEL-TT

DOOR CAM GUIDE ——J17 .

BLDG. 7503, FIRST FLOOR
(ELEV. 852 ft-0in.} .

   

/

¢
i
t

  

;

7 \

/ L

AIR INLET DUCT —~.__ |
/ ~

/" AIR DUCT FLANGE -~~~

1
|

 

Fig.

_SHOCK SPRING

7| INLET DOOR --}
S

AIR BAFFLES ————F

,-"'}’- \
MAIN AIR BYPASS DUCT
7

26

v/ ;’:’/’ | P
VE CHAIN ~_ ( @

R AND -
R T
! e

i

 

-

 
 

 

AN

 

AIR FLOW

/
/
/

1.9. Salt-to-Alr Radiator.

S =~ . COOLANT PUMP

UNCLASSIFIED
ORNL-LR-DWG 55841R2

_~PENTHOUSE

 

[—AIR CUTLET DUCT

A

5 RADIATOR ENCLOSURE

 

 

 

RADIATOR TUBES
27

Table 1.9. Design Data for Salt-to-Air Radiator

 

Structural material INOR-8
Heat load 10 Mw
Terminal temperatures
Coolant salt 1100°F, inlet; 1025°F, outlet
Air 100°F, inlet; 300°F, outlet
Air flow 164,000 cfm at 15 in. HpO
Salt flow 850 gpm at avg temperature
Effective mean temperature difference 920°F
Overall coefficient of heat transfer 53 Btu/hr-ft?.°F
Heat transfer surface area 685 £t?
Design temperature 1300°F
Design pressure 75 psi
Tube diameter 0.750 in.
Wall thickness 0.072 in.
Tube matrix 12 tubes per row; 10 rows deep
Tube spacing i 1/2 in., triangular
Subheaders 2 1/2 in., IPS
Main headers g in., IPS
Alr-side pressure drop 11.6 in. Hy0
Salt-side pressure drop 6.5 psi

 

4. The headers are designed to assure even flow distribution be-
tween the tubes.

5. In the event of flow stoppage, doors on the radiator housing
can close within 30 sec.

6. The electric heaters mounted inside the radiator enclosure are
never turned off.

7. The salt can be drained in approximately 10 min.

The layout of the tube matrix will allow movement of the tubes with
minimum restraint during thermal expansion. The tubes are pitched to

promote drainage.
28

The radiator is supported and retained in a structural steel frame
that is completely enclosed and insulated. Reflective shields protect
structural members from excessive temperatures. The frame also provides
guides for the two vertical sliding doors, which can be closed to thermally
isolate the radiator. The doors are installed on the radiator enclosure,
one upstream and one downstream, and they can be raised and lowered at a
speed of 10 ft/min during normal operation by a gear-reduced motor driving
an overhead line shaft. The doors are suspended from wire rope, which
runs over sheaves mounted on the line shaft. The enclosure is capable
of sustaining full blower pressure with the doors in any position. The
doors may be used to regulate the air flow across the radiator as a means
of controlling the reactor load. However, the load is normally controlled
by positioning a damper in a bypass duct and by switching fans on and off.
More detail of the load-control plan may ke found in Section 2.

Emergency closure is effected by de-energizing a magnetic brake on
the line shaft; this permits the doors to fall freely. ©Shock absorbers
are provided. Accidental closure has no scrious effects except for loss

of time; it reduces the reactor vnower to <100 kw.

1.2.5 Drain and Storage Tanks

 

Five tanks are provided for safe storage of salt mixtures when they
are not in use in the reactor and coolant systems. They comprise two
fuel drain tanks, a fiush salt tank, a fuel-and-flush-salt storage tank,
and a coolant drain tank.

Fuel Drain Tanks. The fuel drain tanks serve the important function

 

of subecritical storage of the fuel. They are water cooled for removing
fission-product decay heat, and an electric furnace is installed for
maintaining the salte molten when the internal heat generaticn rate is
low. Two tanks of the design shown in Fig. 1.10 are provided; each has
a volume of 80 ft?. Each tank can hold an entire fuel charge, so one is
for normal use and the other is a spare. The low moderating power of the
salt makes criticality very unlikely, even with nearly double the planned
U235 loading (see sec. 7.1.9).

After long-term operation at 10 Mw, sudden draining of the fuel will

regquire that it be cooled at a rate of 100 kw to prevent excessive fuel
29

UNCLASSIFIED
ORNL-LR-DWG 61719

INSPECTION, SAMPLER, AND
LEVEL PROBE ACCESS

   
  
  
   
 
  
    
  
 
  
  

STEAM OUTLET
STEAM DOME

CONDENSATE RETURN

WATER DOWNCOMER INLETS

CORRUGATED FLEXIBLE HOSE
STEAM RISER

BAYONET SUPPORT PLATE

STRIP WOUND FLEX!BLE
HOSE WATER DOWNGCOMER

BAYONET SUPPORT PLATE
HANGER CABLE

GAS PRESSURIZATION S

AND VENT LINES INSTRUMENT THIMBLE

FUEL SALT SYSTEM
FILL AND DRAIN LINE

SUPPORT RING

FUEL SALT DRAIN TANK

BAYONET HEAT EXCHANGER

THIMBLES (32) TANK FILL LINE

3

 

 

 

 

 

 

 

 

FUEL SALT SYSTEM
FILL AND DRAIN LINE TANK FILL LINE

Fig. 1.10. Fuel Salt Drain Tank.
30

°  Evaporative cooling was chosen over gas or other means on

temperatures.
the basis of simplicity and independence from utilities. Heat is removed
by 32 bayonet cocoling tubes (Fig. 1.11) inserted in thimbles in the tank.
Water is fed through the center tube of the bayonet assembly, and steam

is generated in the surrounding snnuvlus. Heat is transferred from the
thimble to the cooling tube by radiation and conduction. Normally the
steam is condensed in a water-cooled condenser, but it can be exhausted

o the vapor-condensing system in the event of failure of the coolant
supply. A 300-gal feed-vwater reserve can provide cooling for 6 hr.

The drain tanks have dip-tube fill and drain lines and gas connections
for maintaining a helium blanket for ventilating the space over the salt
and for pressurizing to transfer the salt. Desigh data for the drain
tanks gre presented in Table 1.10. |

Flugh Salt Tank. The LiF-Bel's salt mixture with which the fuel sys-

 

tem is flushed before maintenance will not accumulate sufficient fission
products to require cooling during storage. For this reason the flush
salt tank does not have a cooling system. Otherwise its design 1s similar
to that of the fuel drain tanks.

Coolant Drain Tank. A tank (see Table 1.1C) of 50-ft3 capacity that

 

ig similar tc the fuel drain tanks but without cooling tubes 1s alsc pro-
vided for the coolant salt.

Tuel Storage Tank and Chemical Processing System. Batches of fuel

 

or flush salt removed from the reactor circulating system can be processed
in the fuel storage tank and its associated equipment to permit their re-
use or to recover uranium. Salts that have been contaminated with oxygen
conctituents as oxides can be treated with a hydrogen-hydrogen fluoride
gas mixture to remove the oxygen as water vapor.

A salt batch unacceptably contaminated with fission products, or one
in which it is desirable to drastically change the uranium content, can be
treated with fluorine gas to separate the uranium from the carrier salt

by volatilization of UFg. In some instances the carrier salt will be

 

°L. F. Parsly, "MSRE Drain Tank Heat Removal Studies,"” USAEC Report
ORNL CF-60-9-59, 0Ozk Ridge National Laboratory, September 1960.
31

UNCLASSIFIED
ORNL-LR-DWG 60838A1

 

 

 

 

- L
i v
e e
3 u
s o
< >
w L
~ )
wy L

| \
\ 1

AN
A
A
X

 

 

 

 

T
L
—
z
o
i
<[
=

 

STEAM DOME
LOWER HEAD

BAYONET SUPPORT PLATE

 

 

DRAIN TANK HEAD

_?;_
\ __
_ _:

    

Bayonet Cooling Thimble for Fuel Drain Tanks.

Fig. 1.11.
Table 1.10.

32

Design Data for Fuel Drain Tank, Coolant

Drain Tank, and Flush Salt Tank

 

Tyel drain tank

Material
Height

Diametcr, outside
Volume

Total

Fuel (normal)

Gas blanket (normal)
Wall thickness

Vegeel

Dished head
Design temperature
Design pressure
Cocling methed

Coocliling rate

Coolant thimbles
Number
Diareter, IPS

Coolant drzirn tank

Mztcrial
Height
Diameter, outside
Volume
Total
Coolant salt
Gas tlanket
Wall thickness
Vesgel
Dished head
Degign temverature
Desipgn pressure
Cooling methed

Flush salt tank

Material
Height
Diameter, ocutside
Velume

Total

Flush salt

Gas blanket
Well thickness

Vessel

Dished head
Design temperature
Design vressure
Cooling method

INOR-8
86 in. (without coolant
headers )

62 psi

Boiling water in double-
walled thimbles

o0 kw

INOR-&

78 in.
40 in.

50 T3
~l T3
~6 117

2/8 in.
5/8 in.
13C0°F
65 psi
Nene

INOR-8
84 in.,
50 in.

g2.2 £t
73.2 3
o f3

1/2 in.
3/4 in.
1300°F
65 psi
None

 
33

discarded; in others, uranium of a different enrichment, thorium, or other
constituents will be added to give the desired composition.

The processing system consists of a 117-ft° salt storage and process-
ing tank, supply tanks for the Hp, HF, and F, treating gases, a high-
temperature (750°F) sodium fluoride adsorber for decontaminating the UFg,
several low-temperature portable NaF adsorbers for UFg, a caustic scrubber,
and associated piping and instrumentation. All except the UFg adsorbers
(which do not require shielding) are located in the fuel processing cell
below the operating floor of Building 7503. The process is described in
Part VII of the MSRE Design and Operations Report. After the uranium has
been transferred to the UFg adsorbers, they will be transported to the
ORNL Volatility Pilot Plant, where the UFg will be removed and prepared
for reuse. The entire processing system is separated from the reactor

system with two freeze valves in series to ensure complete isclation.

1.2.6 Piping and Flanges

 

The reactor vessel, pumps, and heat exchanger are interconnected by
5-in.-IPS, sched-40, INOR-8 piping. The 0.258-in. wall is several times
the thickness necessary for operation at 1200°F and 75 psig. Piping of
smaller diameter is used for the drain lines (1 1/2 in.) and other aux-
iliaries, but no salt lines are smaller than 1/2 in.,

The components of the fuel-circulating system are equipped with a
special type of flange, called a freeze flange {(shown in Fig. 1.12), to
permit replacement. These flanges utilize a seal made by freezing salt
between the flange faces, in addition to a conventional ring gasket seal

which is helium buffered for leak detection.

1.2.7 TFreeze Valves

The molten salt in both the fuel and cooclant circuits will be sealed
off from the respective drain tanks by means of freeze valves in the
drain lines. These valves (Fig. 1.13) are simply short, flattened sec-
tions of pipe which are cocled tc freeze the salt in that section. Heaters
surround each valve so that the salt can be thawed quickly when necessary

to drain the system. The salt can also be thawed slowly without the
34

UNCLASSIFIED
ORNL-LR-DWG 63248R2

 

CLAMP— — =

  

NS o
FLANGE \ f'n\"f"‘}é}n \\
;GAPmeH

BUFFER |
CONNECTION
(SHOWN ROTATED)

 

MODIFIED R-68 | i<
RING GASKET -~

 

FROZEN

o 4% —in.
SALT SEAL | v in. R

 

1 Y%-in-R (TYP)
|

 

 

 

 

SLOPE 1:4

//// % (TYP)

!1/4 in.——[
5-in. SCHED-40 PIPE
f
— . L I I ) - - ] . (3

Fig. 1.12. Freeze Flange.

 

 

heaters by stopping the cooling. In addition to the freeze valves in
each drain line, freeze valves are also provided in the transfer lines

between the drain tanks, flush tank, and the storage tank.

1.2.8 Cover-Gas Supply and Disposal

 

Because the fuel salt is sensitive to oxygen-containing compounds,
it must be protected by an oxygen- and moisture-free cover gas at all

times. The principal functions of the cover-gas systems are to supply

an inert gas for blanketing the salt and for the pressure transfer of
 

   

: ! UNCLASSIFIED
 PHOTO 70158
x

 

 

 

 

THERMOCOUPLES RN el SAEME ‘ e - e Wi g

 

   

 

 

 

 

     

-
-

 

 

   
36

salt between components, to provide a means for disposing of radiocactive
gas, and to maintain a higher gas pressure in the coolant system than in
the fuel system. A simplified flow diagram of the system is presented
in Fig. 1.1l4.

The cover gas is helium supplied in cylinders at 2400 psig. It is
purified by passage through filters, dryers, and oxygen traps (titanium
at high temperature). Purified gas is then sent to two distribution
systems, one for the fuel and one for the coolant salt. The total flow
is about 10 liters/min (STP) at about 40 psig. A rupture disk protects
the reactor against supply pressures greater than 50 psig.

The largest flow of gas is directed to the fuel-circulation pump, the
freeze-flange buffer zones, and the fuel drain tanks, where it is in di-
rect contact with the fuel salt. Gas that passes through the fuel pump
is circulated through a series of pipes where it is held and cooled for
at least 2 hr to dissipate heat from the decay of short-lived fission
products. Then it passes through & charcoal bed, where krypton and Xenon
are retained for at least & and 72 days, respectively, and through a fil-
ter and blower To the offgas stack. There 1t 1s mixed with a flow of
18,000 cfm of air, which provides dilution by a factor of approximately
10° and reduces the concentration of Kr85, the only significant isotope
remaining, to 10-7 uc/cc at the stack exit. The charcoal bed is a series
of pipes packed with activated carbon. It and a spare bed are mounted
vertically in a secaled, water-filled secondary container; either or both
beds may be used.

Fuel salt transfers require more rapid venting of the cover gas, but
the heat locad is low. A third charcoal bed is provided for venting those
gases before they are sent to the stack. Figure 1.15 is the offgas dis-
posal flowsheet.

The cover-gas distribution for the coolant system (also shown on
Fig. 1.14) supplies a small flow of helium to the coolant system, the
sampler-enricher system, and to the coolant pump. Gas from the coolant
system is vented directly through filters to the offgas stack, as indi-
cated in Fig. 1.15. Monitors will stop the flow to the stack on indica-
tion of high activity.
250 psig

DRYER
N

TO ATM.

OXYGEN
REMOVAL

UNIT
1200°F

\ F_{ 7

 

HELIUM
SUPPLY
TRAILER
2400 psig

 

 

 

Co

SUPPLY
. HEADER

T

ARRANGEMENT AT CONTAINMENT
CELL WALL TYPICAL FOR SUPPLY

FUEL PUMP SWEEP GAS

NTAINMENT

RUPTURE
DISC
350 psig

f—

UNCLASSIFIED
ORNL DWG. 64-600

Py LEAK DETECTOR SYSTEM

 

 

 

TREATED

HEL IUM
SURGE

TANK

 

 

 

| REACTOR

CELL

TO:

FUEL PUMP LEVEL BUBBLERS

FUEL DRAIN TANKS

Fig. 1.14.

ZZ( ENRICHER-SAMPLER
250 psig HEADER

GRAPHITE SAMPLER

RADIATION
MONITOR
f-—\TO STACK
RUPTURE DISC

50 psig
ey SPENT FUEL PROCESSING

40 psig LEVEL BUBBLERS
HEADER I
FUEL SYSTEM DRAIN TANKS
g COOLANT PUMP
AND DRAIN TANK
P FUEL PUMP SWEEP GAS

e e PUMP O SYSTEMS

 

Cover~-Gas System Flow Diagram.

LE
UNCLASSIFIED
ORNL DWG. 64-594R

RESTRICTOR\\

 

 

 

 

 

 

 

 

 

     

 

 

 

 

 

 

 

    

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

FUEL
PSMP | [ CHARCOAL BED =
HELIUM VOLUME s @r . VOLUME
SUPPLY HOLDUP LOOP | HOLDUP LOOP SPARE
L 1] I XF_cHarcoal Beo X
l AUXILIARY N
OVERFLOW 1 ™ CHARCOAL BED
TANK -
I | .
ACTIVITY | -
1 I MONITOR¥ |
I |
TYPICAL
DRAIN I g& <t
TANK I Bwkgwe STACK
—————— ACTIVITY
HEADER MONITOR
REACTOR CELL |
— —— s S ST P— — —— COMPONENT COOLING $ ABFSIE'FEI;E
‘—TJ— SYSTEM BLOWER J | i
30 in. { ][ } h
21,000 c¢fm FAN
ACTIVITRY*
' MONITO
COOLANT A
PUMP l
HELIUM
SUPPLY X I &L N
- —L%
COOLANT
SALT DRAIN * DRAINS REACTOR ON HIGH ACTIVITY
TANK

 

 

 

Fig. 1.15. 0Off-Gas System ¥Flow Diagram.

8¢
39

1.2.9 Sampler-Enricher

 

Small quantities of fuel can be added to or removed from the fuel
system by means of a sampling mechanism that is connected through the
top of the fuel pump bowl. As shown in Fig., 1.16, a cable assembly with
a motor-driven reel is used to lower and raise a small bucket through a
transfer tube into the salt pool in the pump bowl. The drive unit, which
is located outside and above the reactor containment vessel, is enclosed
in a special contaimment box (area 1C)., A gate valve provides access for
the containment box to the pump bowl through the transfer tube. A port

is also provided to remove the sample or to insert the enriching capsule

UNCLASSIFIED
CRNL-DWG 63-5848

 

REMOVAL VALVE AND
SHAFT SEAL

SN S T "‘___PER|SCOPE
CAPSULE DRIVE UNIT—\= 7 /# = A U LIGHT

._---p"/j:;{

 
  
  
   
  
   
 
  
    
   
      
  
 
  

 

P

,Z:J/CASTLE JOINT (SHIELDED
] WITH DEPLETED URANIUM)

LATCH-. |/
ACCESS PORT—t. '

AREA 1C o
(PRIMARY CONTAINMENT) -

 

 

SAMPLE CAPSULE 7T~ \—MANIPULATOR

T~ AREA 3A (SECONDARY
/ CONTAINMENT )

4 S~ SAMPLE TRANSPORT
' CONTAINER

AT LEAD SHIELDING

 

   
  
    
 
 
  
 
   
 
 
 
 

     

o
7‘{—;-j7" B
/l ’

< 7

 

o
g

 

 

 

== AREA 2B (SECONDARY
CONTAINMENT)

 

 

 

 

 

 

 

 

 

TRANSFER TUBE
{(PRIMARY CONTAINMENT ),

 

 

 

LATCH STOP —¢

B=BUFFERED ZONE

 

 

 

 

MIST SHIELD ‘ o B 2
CAPSULE GLIDE FEET

Fig. 1.16. MSRE Sampler-Enricher.
40

into the box. This special containment box is further enclosed in & secon-
dary containment area (area 3A), which also serves as a manipulator area
in which the sample is prepared for shipment. Access to this manipulator
ares 1s obtained with a transvyort container tube that is inserted through
a gas-buffered sliding seal and a ball valve.

All parts of the system that can be opened directly to the pump bowl
(the transfer tubc and area 1C) are designed for a pressure of 50 psig
at 100°F and comply with ASME Code Case 1273N-7. These parts are con-
tained in two separate vessels (arecas 2B and 3A) designed for 40 psig at
100°F and comply with ASME Code Case 1272N-5. Normal operating pressures
are 5 psig and atmospheric, respectively. All closures for parts opening
into the purp bowl are either welded or doubly sealed with a buffer gas
between the seals. Closures in the containment vessel that will be con-
taminated (area 3A) are also welded or doubly secaled and buffered,

When the system is not in cperation, there are three barriers between
the pump bowl and atmosphere: (1) the operational valve, (2) the access
port, and (3) the removal valve, Each of these barriers contains two
seals, with buffer gas between the seals., During sampling, interlocks
permit opening only one barrier at any given time. A barrier is con-
sidered closed only when the gas dpressure in the buffer region can be
maintained at or above a predetermined pressure. Each buffered region
is supplicd with helium through a flow restrictor, which 1s sized to pro-
duce a significant pressure drop at a hellum flow rate of 5 cc (STP)/min.
If the leakage rate is greater than this, the gas pressure cannot reach
the set point, and the barrier is not considered closed.

Fission-product gases released from the sample and from the pump
bowl during the time the operational valve is open are later cischarged
to the auxiliary charcoal bed. Interlocks prevent opening area 1C to
the charcoal bed until area 1C is isoclated from the pump bowl by the
operational valve,

Interlocks also prevent insertion operation of the cable drive unit
until both the operational and the maintenance valves are open. Neither
valve can be closed until the cable has been withdrawn sufficiently to
clear the valve seats. The access port cannot be opened until the pres-

sures on both sides of it are equal, to prevent damage to the equipment.
41

To prepare a sample for shipment to the analytical hot cell, the
gample is sealed inside a transport container tube. A double elastomer
seal prevents the escape of fission-product gases from the tube. The
tube containing the sample is drawn up into a lead-filled cask and secured
in it. The cask is then sealed inside & can, which is bolted to the frame
of a truck. Thus, the sample is doubly contained. The cask is secured
ingide the can so that the sample will remain in place and sealed in case

of an accident to the truck.

1.2.10 ZElectric Heaters

 

External heating of salt-bearing components of the reactor system is
necesgary (1) to prevent freezing of the salt, (2) to raise the reactor
temperature to a subcritical value for experimental convenience, and (3)
to heat the salts for reactor startup. Replaceable electric heaters were
chosen as the safest, most reliable means of supplying the large (~500 kw)
high-temperature requirement. Diesel-driven generators and two separate
TVA substations make a complete failure of the heating system extremely
unlikely. (Fig. 1.17 is a schematic diagram of the power distribution
system.) In the 1000 to 1500°F range, nearly all the heat is transferred
by radiation., This makes it unnecessary for the heating clements to be
in contact with the vessel walls and results in a safer system from the
standpoint of overheating and arcing damage. Each heater circuit has an
electric current meter to indicate proper functioning. Heat balances of
the system are taken at 2-hr interwvals.,

Thermocouples attached to the piping and vessels are located in re-
lation to the heaters so that maximum temperatures are indicated. The
thermocouples are scanned continuously to identify temperatures 100°F
above or below a chosen standard so that corrective action can be taken.

Different kinds of heaters are applied to different parts of the
regetor; most are kept energized continuously during reactor operation.
The core vessel, drain tanks, and flush tank are equipped with resistance
heaters which fit into wells surrounding the vessels. The piping and the
heat exchanger are covered with resistance heater assemblies that are

designed for easy removal. Reflective insulation is incorporated in the
 

NIESEL GENERATORS

MORMAL SOPPLY

Mfi/ 1500 KVA

 

GEMNERATOR BUS We 3 I

So0h Ny y

s
5

{75 AAA
CORLANT COOLING AU, 100 KVA

SALT WATER POWER LIGHTING

PUMP P TRANSE.

QENERATOR BUS N? 4

) ) ) )

 

Oy

O—_q

-
G

o ALTERNATE SUPPLY

VTV\ 2.8 KV/480 V.

Tva RBUS (NORMAL SUPPLY)

I
%) o) 5) I)

{0

BRADIATOR,
RS

FUBL PUMP
RLOWE HEAY EXCHAWGER
REMCTOR

MISC.
HEATERS

 

 

 

&

LIGHTIMG
DL BRAEAKER i\, — " !

 

TRIP POWER

 

REACTOR CONTROLS

AUTO
INSTRUMEBEWTATION o TRAMSFER
* S, SWITEH

QENERATOR BUS NT §
L «© *

i ! 1
o) ) )

 

 

 

&
’ AUX
FUEL cnoLIuGa POWER ) 125 KW- 250 V. N.C.
ohLT WATER o OUTPUT - MG * !
PuMP PUMP ———

— S G M
‘RC TRIP ‘

EMERGELCY ...._11,_./-—-1»—-(\,-"\,)-;--

— ey
|

 

J

;5__mnw_4mdfi}+

7___7;4_‘,Amj

250V
RATTERY

I
i3 .8 MV TRAMSFER
CONTROL POWER

 
 

- -
i Qf)
! 25 KW

- 120/ 240 V.- |
! . MC‘!04 #
i

 

 

DRAIN TANK FUEBL COOLAMT
. DRANW  RADIATOR
HEATERS PIPE HEATERS
HEATRRAS

Fig. 1.17. Single-Line

Diagram of MSRE Power System.

 
43

design. The coolant system and the radiator are heated by Calrod-type
radiant elements, with ordinary insulation, because remote malntenance

igs not necessary in these areas.

1.2.11 ILiquid Waste System

 

The MSRE radioactive agueous waste is expected to consist mainly of
intermediate level waste (10 uc to 5 curies per gallon). The quantities
of active wastes expected will be low enough to permit dilution in the
11,000-gal liquid waste tank if the allowable concentration 1s exceeded.
The waste will be sampled, neutralized if necessary, and pumped to the
Melton Valley Central Station through an underground pipeline.

As shown on the simplified flowsheet of Fig. 1.18, the system con-
sists of a stainless steel storage tank, a sand filter for clarifying the
shielding water in the decontamination cell, a stainless steel pump for
circulation of water through the filter and for pumping waste to the Cen-
tral Station, and an offgas blower to prevent pressurization of the waste
tank when Jetting into the tank.

The sources of radioactive wastes are listed below.

1. There will be fission-product gases from fuel processing. Since
the caustic scrubber is mainly for HF neutralization, the efficiency is
expected to be low for fission-product removal and the amount of activity
collected should not be large.

2. There will be activity from decontamination of equipment in the
decontamination cell, The active liguid can be pumped directly to the
waste tank or the activity can be accumulated on the sand filter and be
backwashed to the waste tank.

3., Washdown from the offgas stack and filters will contain a small
amount of activity.

4. In case of a serious spill or leak of radioactive material, con-

siderable activity could result during cleanup operations.
CAUSTIC ADDITION FOR NEUTRALIZATION

 

HOT DRAINS AND SINKS

 

CELL SUMP JETS AND SUMP PUMPS

 

 

LIQUID WASTE TANK
(44,000 gallons}

 

FUEL- PROCESS
CAUSTIC SC

 

ING SYSTEM
RUBBER

 

SAND
FILTER

 

 

 

!
N

|
H
|

|
|

|

 

 

UNCLASSIFIED
ORNL DWG. 64-5584

1 .

DECONTAMINATION CELL

 

“-—u—u.—.....-.__.-__—__——
—— e e — T, s, m—————

TO OFF-GAS
( 5 F FILTERS

 

 

 

—

 

 

OFF-GAS
BLOWER
TO CENTRAL WASTE STATICON
e INTERMEDIATE-LEVEL
) STORAGE TANKS

 

LIQUID WASTE
PUMP

Fig. 1.18. Simplified Flowsheet of MSRE Liquid Waste System.
45

2. CONTROLS AND INSTRUMENTATION

The control and safety systems of the MSRE are designed to provide
(l) mesns for safe operation of the reactor and the fuel and coolant salt
systems under normal conditions, (2) the flexibility needed to do experi-
ments with the reactor, and (3) reliable protection against the reactor
hazards usually encountered and against unacceptable damage to the equip-
ment by nuclear accidents or by freezing of the fuel or coolant salts.
Nuclear instrumentation and control and safety devices similar to those
of other reactor types are required. The use of a mobile fuel that melts
at high temperature requires additicnal control and safety systems for some
operations. Some related features reduce the demands on protective de-
vices.

The reactivity required to operate the reactor, starting from the
isothermal (at 1200°F) zero-power condition with noncirculating fuel, is
shown in Table 2.1. The excess loading is a strong function of the tem-
peratures of the fuel and of the graphite moderator. The overall tempera-
ture coefficient for fuel C and the moderator is —7 X 1077 (8k/k)/°F.

The excess loading for the clean reactor, critical at zerc power and at
a temperasture of 900°F, with noncirculating fuel, is 4.0ié'2 % 6k/k.*
The total worth of the control rods for operation with fuei C is calcu-
lated to be 5.7% 8k/k.

The loading of the MSRE is subject to rigorous administrative pro-
cedures reinforced with special control and safety instrumentation*¥* that
limits the rate of fuel addition and reverses the loading by draining
the reactor if potential hazards or hazardous conditions develop. Limita-

U235 in the reactor, that is, control of the

tion of the inventory of
shutdown margin, is effected by instrument or manual drain of the fuel

into the drain tanks.

 

*Determined from

(1.9f_é'2) % 8k/k + (1200-900)°F x (7 x 1079) S—%F&x 100 = (4.03'2)% 8k fic .

**3ee Tables 2.2 and 2.3 and Section 2.2.4.
46

Table 2.1 Reactivity Requirements

 

 

Reactivity,
% sk/k
Loss of delayed neutrons by circulation -0.3
(power in steady state)
Power deficit (at design-point power) 0.1
Samarium transient® and xenon poisoning ~0. 8
(equilibrium at 10 Mw)
Voids from entrained cover gas ~0.2
Burnup and margin for regulation 0.5
Uncertainty in estimates 1.0
+1.0
Total . C
ota 1 2_0.4

 

®Tabulation does not include reactivity required to
compensate for buildup of long-lived Sm149, which levels
off at approximately 1.1% Sk/k after 60 days of opera-
tion at full power. See MSRP Semiann. Progr. Rept.
July 31, 1963, p. 64, USAEC Report ORNL-3529, Oak Ridge
National Luaboratory.

The permanent nuclear instrumentation provides reactor operating
information (power =nd period) over the entire normal operating range of
the system. The initial loading will be preceded by many dry runs using
barren salt. These will afford a thorough check of the system, train the
operators, and verify the procedures. The initial critical experiments
and the regular instruments will be augmented by four channels of BFj
counting equipment, which will extend the counting range a decade or more
below that recuired for normal startups.

As is true for most reactors, transient responses in the MSRE are
strongly influenced by the various temperature coefficients of the fuel
and moderator and by the thermal and transport time constants of the sys-
tem. Over 1ts operating range the power density in the MSRE is low, and
the heat capacities of both the fuel and the moderator are large. Hence,
the thermal time constants of the reactor system are long. These condi-

tions establish maximum rates of load change without appreciable overshoot
47

in the amplitudes of such parameters as nuclear power and temperature pro-
file.

Typical parameters that influence the transient response of the MSRE

when loaded with fuel C are the following:*

Mean neutron lifetime 2.4 X 107% gec

Delayed neutron fractions

With fuel circulating and power in 0.0036
steady state
With fuel stationary 0.0067
Power densities in fuel at 10-Mw level
Maximum, on core center line 31 w/cm3
Mean 14 w/cm®
Temperature coefficients of reactivity
Fuel only (fuel C) ~3.3 x 107°? (8k/k)/°F
Graphite moderator only 3.7 x 1077 (8k/k)/°F

It was possible, by using analog simulation, to devise an adequate
control system which will provide smooth operation over the entire oper-

ating range of the plant.

2.1 Cocntrol Rods and Rod Drives

 

One of the three control rods and its associated drive unit are shown
in Fig. 2.1 as agssembled and installed in the reactor. The locations of
the rods within the core are shown in Fig. 1.6 (see sec. 1.2.1).

The control rod, Fig. 2.2, consists of approximately 36 poison ele-
ments in the form of hollow, cylindrical ceramic slugs of 70 wt % gado-
linium oxide and 30 wt % aluminum oxide clad with Inconel. These indi-
vidual poiscn elements are stacked around and supported by a flexible
stainless steel hose to form a continuous flexible control rod 1.14 in.
in diameter and 56 in. long. The use of flexible rods permits offsetting
the rod drives to obtain access to the graphite sample penetration and

protects against binding from misalignment or thermal expansion.

 

*¥Table 7.1 gives a more complete tabulation showing the influence
of fuel composition on nuclesr characteristics.
 

48

  

T \REACTOR CELL ROOF PLUGS

CONTROL ROD DRIVE
MOTOR HOUSING ——=

—.—CONTROL ROD DRIVE HOUSING

 

 

CONTROL. ROD
THIMBLE NO.3
CONTROL ROD
THIMBLE NQ. 2

~==-—CONTROL ROD THIMBLE NOA.

REACTOR ACCESS
PLUG FLANGE —=

& §\ TR SHELD

SALT QUTLET TO
CIRCULATING PUMP

 

 

 

 

 

C————

 
   

REACTOR ACCESS NOZZLE

 

CONTROL ROD—

140 iu-|

l?

UNCLASSIFIED

ORNL-DWG €4-6083

    
 

 

L/

 

 

 

 

 

 

 

~—.

C
)

REACTOR
VESSEL—=

 

 

  
   

   
  

_|
¢ CORE
|
|

| GRAPHITE
3 /_/iLATTICE BARS

 

  
 

FPig. 2.1. Control Rod Drive Unit Installed in Reactor.

O

 
POISON SLUG
SEE DETAIL

       
  
  
   
   
 
  

  

£-in. DIAM BRAIDED
WIRE CABLE.

AR FLOW

AlR FLOW

END FITTING WITH
INTEGRAL AIR NOQZZLE

N
"FLOW
RESTRICTOR

49

{ .
-8§|nx

LOCATION OF FIDUCIAL ZERO

UNCLASSIFIED
ORNL- DWG 63-8394

N

 

 

 

2-in. 0.0.

W
wsin !

- !’L 2”"7'* -7

KGR, 5

 

 

 

 

 

 

 

 

 

 

 

!

 

   

THIMBLE
_OUTER TUBE

 

 

 

rd
UPPER SEALING Gdp O3~ Al, O3 BUSHINGS 7
RING ... / &/ . LOWER SEALING
ey i . A e RING
NS TS ST
i /
S _-POISON SLUG
| INNER TUBE” TWICE SIZE
0.790-in. o
1D,
1.140-in.
0.0,
L

 

 

3

| NI AR K

S
2,

 

Fig. 2.2.

 

- -

1.560 in.

 

- -

.

 

Diagram of Control Rod.

Analysis indicates that with the reactor at full power, the control

rod poison elements will operate at a temperature of about 1500°F, if no

cooling is provided,
thimbles.

This temperature will not damage the rods or the

The rod drive design is conservative in that provision is made

for flow of cooling air, which may be routed down the inside of the flexi-

ble hose and the hollow inside of the rod and return up through the annular

space between the rod and the thimble, see Fig. 2.2.

The need for con-

tinuous cooling will be determined during early phases of operation.

An electromechanical diagram of most of the essential elements in the

drive train is shown in Fig. 2.3. The shock absorber, limit switches,
air flow system, and the pneumatic fiducial zero-positioning device do not
appear on this figure. The rod is scrammed by breaking the circuit-supply-

ing current to the electromechanical clutch.
50

 

~-—— FiNE

TO POSITION READOUTS
IN CONTROL ROOM

ROD POSITION
INPUT TO SAFETY SYSTEM

 

 

 

 

 

 

 

 

 

-a4— COARSE ————

    

 

_/w

SYNCRC NO.2
60° PER [NCH

 

 

OF ROD MOTION

 

 

 

POSITION
POTENTIOMETER.

 

 

 

 

 

 

 

 

SYNCHRC NOA
5° PER INCH
OF ROD MOTICN

f“*} ELECTROMAGNETIC
29, CLUTCH

 

 

 

UNCLASSIFIED
ORNL-DWG 63-8391

REDUCTION
GEARING

 

 

 

   

 

 

 

TO SAFETY
SYSTEM e
FAN
MOTOR
TACH
SERVO
MOTOR
S
i Al
REDUCTION “EQUAL RPM
GEARING j -
NCLUDES M
REVERSE :
_OCKING

OVERRUNNWNG///
CLUTCH 7

Fig. 2.3.

 

POISON ELEMENTS

HORIZONTAL
GRAPHITE BARS 3

CORE VESSEL—™

 

L 1-TO-1

 

AR FLO
TO COOL

LEXIBLE

TUBULAR ROD
SUPPORT. "

¥=0.5in./

THIMBLE —-1

 

==

\—

GEARS

1A

W
RQOD

sec

 

3

 

 

—_———— e )

 

 

   
  
 

PROCKET

DRI
(s
%9—3[

Le— SPROCKET
CHAIN

| GRID PLATE

 

Electromechanical Diagram of Control Rod Drive Train.
51

The reactivity worth of a rod as a function of depth of insertion in
the core ig shown in Fig. 2.4. Other features of the rods and drives
pertinent to reactor safety are listed below:

1. The rod velocity during normal motor-driven insertion and with-
drawal is 0.5 in./sec ($10%). Msximum stroke is 60 1in.

2. The clutch release time is less than 0.05 sec; the average drop
timel from a height of 51 in. is 0.79 to 0.81 sec, including release bime;
and the average acceleration during a 51-in, drop is 13 ft/secz. It has
been established, as indicated in Fig. 2.5, that satisfactory scram per-
formance is obtained with two rods dropping with an acceleration of
5 ft/sec2 and with a clutch release time of 0.100 sec.

3, The drive motor is a 115-v, 60 cps, single-phase, capacitor

start-and-run, positively reversible unit.

 

17, R. Tallackson, "Scram Performance of the Prototype MSRE Control
Rod," internal ORNL document MSR-64-7, Feb. 10, 1964.

UNCLASSIFIED
ORNI—DWG 64-—-623

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1.0
./—'/'_—-
P
! }
0.8 :
I {
|_
z |
o |
= EQUIPOISE-3 CALC. ;
4 0.6 | IL o
2
wZ
P )
& / |
> 0.4 ¢
S / .
»— i
g / |
x /
0.2 /,I
//
2
z
00——"”‘
0 10 20 30 40 50 60 70

DISTANCE INSERTED (in.)

Fig. 2.4. Reactivity Worth of Control Rod as a Function of Depth
of Insertion in Core.
52

UNCLASSIFIED
ORNL—DWG 64— {1109R

 

 

 

 

DISTANCE (in.)

 

 

 

 

s b— — e e e XX' .

60

 

 

 

 

 

 

 

Q 0.25 0.50 0.75 1.00 .25 1.50
ELAPSED TIME (sec)

CURVE A-REFERENCE CURVE QF SATISFACTORY SCRAM PERFORMANCE;
BASED ON ACCELERATION OF 5 ft/sec? AND RELEASE TIME OF 0400 sec

CURVE B-SCRAM PERFORMANCE FROM TESTS OF JAN.27-28,1964

Fig. 2.5. Control Rod Height Versus Time During a Scram.

4. The overrunning clutch transmits drive motor torque in the rod
insert direction, and the control system contains an interlock with the
safety system that turns the drive motor "ecn'" in the rod-insert direction
following a rod scram. This will provide a force in the order of 350 1b
to push the rod into the core if it is stuck.

5. A low-~angle, low-efficiency worm and gear between the servo motor
and the sprocket locks the drive train so that if an external torgue in
the direction tending toward rod withdrawal is applied to the drive
sprocket, 1t will not rotate. A conservative stress ectimate of elements
in the drive train from the worm gear to the sprocket chain gives a value
of 463 1b as the gafe static torque. This torque is produced by a dif-
ferential pressure of 450 psi acting on the projected cross-sectional area

of the control rod.
53

6. The electromechanical clutch is a single-plate, flat-disk unit
with the driven plate spring lcoaded: it has a stationary field winding.
The clutch is disengaged with no field current.

7. The coarse-position synchro rotates 300° for full stroke and,
hence, provides unambiguous information. The synchros transmit the angular
position of the drive-sprocket shaft and do not take into account small
changes of rod position caused by stretching of the drive chain and the
hollow, flexible, support hose. A pneumatic, single-point, fiducial,
zero-position-indicating device 1s provided that is activated by the change
in pressure drop of the cooling air leaving the bottom of the rod. When
the rod is at the zero position, the flow of cooling air from the radial
exhaust nozzle at the bottom of the rod is impeded by a constricting throat
near the bottom of the thimble and the sharp change in differential pres-
sure indicates the location of the rod with respect to the thimble to an
accuracy of 0.030 in. Figure 2.2 shows the nozzle and throat. This device
will be used for periodic determination of the relation between synchro
outputs and the position of the poison elements with respect to the thimble.

2. The shock sbsorber, Fig. 2.6, is of the spring-loaded "hydraulic"
type, but it is unconventional in that the working "fluid" in the cylinder
is hardened steel balls. The stroke ig adjusted between 2.75 and 4 in.
by varying the spring constants and the preloads on the buffer return
spring and the ball reset spring. The resultant deceleration during shock
absorption is seven times gravity. Tegts! of the prototype rod and rod
drive have included more than 1100 scrams from full withdrawal without any
malfunction of the shock absorber. This i1s greatly in excess of the duty
required in service.

9. The weight of the control rod plus the shock absorber produces
an unbalanced weight on the sprocket chain of 18 1b. Static tests? of
the prototype show that the internal friction of the drive unit plus the
thimble-to-rod friction is equivalent to an upward force of 10 1b on the
sprocket chain. Thig leaves an unbalanced force of 8 1b in the scram
direction. This is in good agreement with the acceleration given in para-

graph 2 above.

 

M. Richardson, private communication, May 12, 1964,
 

UNCLASSIFIED
ORNL-DWG 64-983

54

 

 

 

T T Y T T e e o et o ool ool Z T BBl d Bl 2T T T T T T T T T TTTT g LTI

e SR T AaH T ., M
| ;wfiwmhfl...fluunnm,pawqm,flmu ......... - @ == - | &\\ \\W

) Q8
= SR A 00000008 Attty

Ry 2o eiad = e \ >

 

 

 

 

 

 

 

 

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_c¢ccc¢c

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2N
\\/\
e e i A e e /\
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& 2 Q-
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B g & Q 3 o
2 z P 5 x 9 EZ y 2
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_m bu o o & w5 m.M 2 & m ¥ o~
4 Zz G g o £ 4 < & B @ =
E o : t 2 2 3 Wi & z 3
3 S g 2 i & 2 ooh 3 8 B E g

Control Rod Shock Absorber.

Fig. 2.6.
55

10. The upper and lower limlt switches at‘the ends of the total rocd
stroke are in independent pairs and are wired independently. All these
switches are located at the upper end of the drive unit to eliminate long
electrical leads adjacent to moving parts. This 1s also the most favorable
location from the standpoint of temperature and radiation intensity.

11. A photograph of a prototype thimble is shown in Fig. 2.7. It is
substantially the same ag the thimble that will be installed in the core
vessel. The bottom of the thimble is closed and (see Fig. 2.1) is only
0.% in. above the graphite stringers on the grid plate. Thus there is
ample protection against a rocd falling out of the core and initiating =
nuclear excursion. The prototype rod drive assembly, supplied by the Vard
Division of Royal Industries, Inc., Pasadena, Californis, is shown 1In

Mg, 2. 8.

2.2 Safety Instrumentation

 

The instrumentation and control devices classified as safety equip-
ment are those whose Tailure to perform their protective function when
required would result in an unacceptable hazard to personnel or unaccept-
able damage to the reactor or its major components.

The following principles were used in the design of the MSRE safety
syetem in order to obtain the maximum degree of reliability:

1. Redundant and independent channels composed of high-grade com-
ponents were provided for each protective function. Two degrees of redun-
dancy are employed: (a) two independent channels, either of which will
produce the required safety action, and (b) three independent channels
in two-out-of-three coincidence. The second system requires agreement
of any two channels to produce safety action. Removal of any channel for
maintenance or operational checks i1s equivalent to a safety system input
from that channel. In these circumstances the three-channel system re-
verts to a simple two-channel system,

2. Provisions were made for periodic on-liine testing of each channel.

3. Continuous monitors were provided to disclose certain component

failures and malfunctions, power failures, and loss of channel continuity.
SRR T T AR A R AR T e T R TR R T A A e

 

* UNCLASSIFIED
PHOTO 39863

56

 

 

 

 

 

 

 

 

 

 

 

 

Control Rod Thimble.

7

2

Fig.
'

L\"

 

 

0.

 

_“. CONTROL ROD DRIVE
- UNIT HOUSING

 

 

¥ THIMBLE ADAPTOR FLANGE

 

 

57

“7 UNCLASSIFIED
© PHOTO 62054

   
 
  
  
 
  

_ POWER- AND POSITION-
i INDICATOR PACKAGE

 

_CHAIN DRIVE -~

 

  

. TOWBLOCK AND =~
" ROD DISCONNECT -

   

 

 

 
 

 

 
e e e e

58

4. The use of all components in the safety system is restricted to
providing safe reactor Qpefatibn; e.g., modifications to the safety in-
strumentation for the confenience of data gathering by experimenters is
prohibited. Similarly, the addition to the safety system of extra or
auxiliary fUncfions that do not contribute to reactor system safety is
considered a dilution of safety system efféctiveness and is prohibited.
Typical alterations in this category are nonsafety interloéks and alarms.

5. Physical protection and separation of all components in each
channel (conduiting, closed cabinets, etc.) are provided.

6. All safety-equipment components in each channel were identified
as such. ’ ) | |

Protection; separation, and identification (items 5 and 6)'are in-
corporated throughout the system. Items 1 through 4 are discussed in
detail inithe following paragraphs. A simplified block diagram of the
MSRE nuclear and process safety system is shown in Fig. 2.9, which dis-
pléys both the various inputs or conditions used to detect or to indicate
the existence of unsafe conditions and the results obtained when these in-
puts indicate that a hazard, real or potential, exists. The safety sys-

tem input and output elements, their corrective actions, and the means em-

ployed to attain the required reliability are listed in Tables 2.2 and 2.3,

2.2.1 Nuclear Safety System

The nuclear safety system is required to decrease the reactivity re-
liably upon the occurrence either of a high neutron flux or a high reactor
outlet température. The primary safety elements for accomplishing this
are the electromechanical rod-release clutches (for a description of the
rod-drive mechanisms, see sec. 2.1). Deenergizing any clutch releases
the drive sprocket (Fig. 2.3) from the motor and brake, and the absorber
rod fglls into its thimble in the core. The instrumentation associated
with these clutches and the initiation of a rod drop (scram) is described
here. ’

A block diagram of the nuclear safety system instrumentation is shown
in Fig. 2.10. The ranges of the flux safety channels, along with those of
the flux channels used for control, are listed in Table 2.4. Details of

chamber installations are given in Section 2.3.

4

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

59
UNCLASSIFIED
ORNL-DWG 64-637
CONDITION CORREGTIVE ACTION
FLUX SCRAM
>15 Mw ALL RODS

REACTOR OUTLET _

TEMPERATURE >1300°F — v
OPEN -
- BYPASS VALVES \
¢ THE BYPASS VALVES AND ~ THAW DRAIN
THE VENT VALVES ARE e FREEZE VALVES REACTOR
REDUNDANT SETS, EITHER
WiLL PROVIDE PRESSURE OPEN
RELIEF IN DRAIN TANK *
DURING A FUEL DRAIN VENT VALVES
-
ANY ROD NOT ABOVE B CLOSE MAIN He SUPPLY
FILL POSITION 1 VALVE TO DRAIN TANKS
'FUEL PUMP BOWL CLOSE He SUPPLY VALVES
PRESSURE > 2 psig USED TO FILL REACTOR

 

 

 

 

 

 

 

 

 

 

 

FUEL PUMP BOWL He

 

 

 

 

 

PRESSURE > 50 psig

 

FUEL PUMP OVERFLOW
TANK LEVEL >20% FS

 

 

 

 

AGTIVITY IN REACTOR
GELL AIR

 

 

 

 

 

COOLANT PUMP OFF-GAS
ACTIVITY > 20 mr/hr

 

 

OPEN FUEL PUMP BOWL
VENT VALVE

 

 

 

CLOSE He SUPPLY VALVE TO
FUEL STORAGE TANK

 

 

 

i

CLOSE BLOCK VALVES
He LINES

 

 

 

 

 

 

 

 

 

 

He SUPPLY PRESSURE

 

 

 

CLOSE BLOCK VALVE IN CELL
EVACUATION LINE

 

 

 

< 30 psig

 

 

 

CLOSE He SUPPLY LINES TO IN-CELL
LEVEL INSTRUMENT BUBBLERS

 

 

 

ACTIVITY IN
QFF-GAS

 

 

 

JTEMPERATURE AT RADIATOR

 

OUTLET <1300 °F

 

 

LOSS OF -COOLANT

 

 

CLOSE BLOCK VALVE IN MAIN
OFF-GAS LINE TO STACK

 

 

 

CLOSE LUBE OIL SYSTEMS
VENT TO OFF-GAS VALVES

 

 

 

 

SALT FLOW

 

 

ACTIVITY IN He LINE TO

 

CLOSE RADIATOR DOORS

 

 

 

IN-CELL- BUBBLERS

 

 

 

 

REACTOR CELL

GCLOSE LIQUID WASTE
SYSTEM BLOCK VALVES

 

 

 

_ PRESSURE >2 psig

 

 

 

ACTIVITY N IN-CELL -

_CLOSE INSTRUMENT AIR

 

LINE BLOCK VALVES

 

 

COOLING WATER SYSTEM

 

 

~

Fig. 2.9.

f

’”

 

. CLOSE IN-GELL COOLING
WATER BLOCK VALVES

 

 

Functional Diagram of Safety System.

 
' : ‘ UNCLASSIFIED
: . : ORNL~DWS 64-638

 

 

 

CHAMBER VOLTAGE
‘ | MONITOR AND TEST
- Q-260%

CHAMBER VOLTAGE
MONITOR AND TEST|[ |

CHAMBER VOLTAGE

MONITOR AND TEST
e o e e i e o | e b Q-2601

HIGH
TEMPERATURE
TRIP

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

e | Q-2601
oo [—— " - \ r---r r 7
'] CHAMBER FLUX ! | | cHAMBER | | cHamser | \ FLux '
I | Hv, supPLY AMP, | ‘ | | HwvsuPPLY i | WV suppRLY MP, |
I Q-2602 | I | Q-2602 |
- ——t Lo e e b o o e e e o e e =)
DATA DATA DATA
LOGGER LOGGER LOGGER

Vo . ~

L L. | | | L I

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

FAST TRIP FAST TRIP FAST TRIP FAST TRIP FAST TRIP FAST TRIP FAST TRIP FAST TRIP FAST TRIP
COMPARATOR COMPARATOR COMPARATOR COMPARATOR COMPARATOR COMPARATOR COMPARATOR COM PARATOR COMPARATOR
Q-2609 Q-2609 Q-2609 Q-2609 Q-2609 0-2609 Q-2609  0-2609 Q-2609
ECC ECC EcC gcCc | gce gce
ROD REVERSE SAG BYPASS ROD REVERSE SAG BYPASS SCRAM ROD REVERSE SAG BYPASS ‘
: SCRAM SCRAM

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

v, o £z J
- . Ty Ny
’ | —l ! RELAY SAFETY
RELAY SAFETY ELEMENT | RELAY SAFETY
’ ELEMENT | ' ece Q-2623 : ELEMENT
Q-2623 | A T | Q-2623 _
Ecc | e | L £ec
- ALARM ] ] ALARM
N S 5 J - J [
COINCIDENCE COINCIDENCE COINCIDENCE
MATRIX MONITOR MATRIX MONITOR MATRIX MONITOR
~Q-2624 Q-2624 Q-2624
NO. 1 ROD - . |noz rop NO.3 ROD
CLUTCH . CLUTCH CLUTCH

 

 

 

 

 

 

 

 

 

Fig. 2.10. Block Diagram of Safety Instrumentation for Control Rod Scram.

~

09

 

 

 

 

O

 

 
 

.‘

 

 

-)

-t

Drain fuel

61 Nf
v C | :
o Table 2.2. MSRE Safety System Inputs
Condition or Situation Which Indicates - Causes of the Hazard, the Consequences, and v ) ‘
a Real or Potential Hazard the Corrective Action 1' - Supplementary Information
: ; ‘
- ' : - :
I.  Excess reactor power; 9 > 15 Mw 1. Causes _ 1. Redundancy: Three independent channels provide flux
\ : ' a. Uncontrolled rod withdrawal - information
b. Premature criticality during filling (excess 2. Testig Response of each channel is tested, with the
U23% due to partial freezing) exception of the ion chamber response to £lux changes
¢. Cold slug
3. Monj._toring. System design provides continucus check on
) d.  Unimowvn or unldentified mechanisms circuit continuity and amplifier operation
‘ 2. Consequences . L 4. ‘- Safétx only?: Input information used solely for safety
o , Excessively high temperatures; first in the fuel salt and to restrict reactor operation within safe limits
loop, and ultimastely in the coolant salt loop, with L vt e '
resultant damage to equipment and, if unchecked, loss 5. Thig system employs two-out-of-three coincidence
. of primary containment - Refer to Section 2.2 and Fig. 2.10
v 3. Corrective action | -
a. Scram all rods _ : !
b. Open vent valves; relieves filling pressure in !
o : fuel dra.in ta.nk (see .1b above) - 9 S i
-IT. Fuel salt outlet temperature 1. Causes [ Redfinda.ncz: - Three indepehdent channels
greater than 1300°F Bee I above; also situations in which power genera-. 2. Testing: Response of each individual input channel may
tion exceeds rate of heat rejection be, tested ‘
2. Consequences - 3. Monfi.toring' A thermocouple break or detaclment from
'See T sbove pipe wall produces upscale reading and sa.fety a.ction
_ . in that channel .
3. Correct;ve actions 4 ‘SaiJet onlv?: Yes
a. Scram rods : ‘ o
b. Drain fuel 5. This system employs two-out-of-thrge coincidence
6. Refer to Section 2.2 and Fig. 2.11
" III. Pump bowl helium pressure greater 1. Causes 1.” System design provides 5.0 £t of overflow capacity, and
than 35 psig power level scram and fuel overflow tank level systems
&. lzzpidruifil;:ckzgee;cpmgi;; Oioizein ga':;'z b{e:;{g::i (see I and IV, this table) protect against fuel salt
inggerm:t ion’ 8 & PP exba.nsion caused by power excursion; safety system holds
bypass valves (Fig. 2.14) open during operation, in
b. TFallure of hellum off-gas letdown valve, ROV- which case the pressure rise will be moderate until all
522A)., combined with faillure of 40-psig inlet the overflow volume is filled
Y - helium reguletion system and rupture disk - B
c. Unknown or unidentified mechanisms 2. Redundancy: Two pressure channels, one in the pump
: bowl and one in the overflow tank
2. Consequences
3. Testigg: Teet procedure’ checks entire channel, with ex-
“ a. If unchecked, dama.ge to pump seals, which are :
rated at 100 psig, and possible overstressing of . .celption of transmitter bellows (see Fig. 2. 15)
primary loop ' 4. Monltoring: ILoss of helium flow to bubblers caused by
b. Exceed capacity of off-gas system, with resulting v either low inlet pressure or line blockage is alarmed
threat to contalnment .
¢c. Possible loss of primary containment Sefety only?: Yes
i 6. Draining (see Figs. 2.9 and 2.14 and Table 2. 3) ensures
3. Corrective actions impediate opening of the bypass valves to back up ad-
&. Drain fuel ministrative control. Venting to the auxiliary char-
b. Vent pump bowl to the auxiliary charcoal bed cofenl bed is less effective but useful backup :
IV. TFuel pump 6verfl%w tenk level 1. Cause 1. This backs up (anticipates) cause la in IIT (this table)
greater than 20% full scale on . : . -
Jevel instruments a. Overfill (malfunction in fill system, misopers- 2. Redundancy: Two independent systems provide level in-
tion, ete.) formation -
. Xpan 1 b; t -
b Efi:s?s:e 1 :Iilgnlclafagg‘ere)caused Yy high tempera 3. T__e_SM: Test procedure checks entire channel with ex-
ception of transmitter bellows (see Fig. _2.15)
2. Comsequemces 4. Monitoring: Loss of helium flow to bubblers caused by
- Loss of capacity to handle fuel expansion or overfill either low inlet pressure or line blockage is alarmed
(. 3. Corrective action 5. Saefety only?: Yes
 

Table 2.2 (con%inued)

 

Condition or Situation Which Indicates
a Real or Potential Hazard

'~ Cguses of the Hazard, the Consequences, and
. 1 the Corrective Action

Supplementary Information

 

V. Helium pressure in fuel pump bowl
greater than 2 psig during fill

_ VI. Reactor cell air activity greater
than 20 mr/hr

VI1I.

Coolant pump off-gas actlivity
. greater than 20 mr/hr
VIII. Control rods not ahove "f£ill"

position during reactor filling

Caus¢s

a. chess £1lling rate or overfill
emperature excursion during £111 (see I and II,

’

this table)

c. -%aqunction in helium letdown system (lines 522,

324, and PCV-52241, etc., see Fig. 2.14)

d. (losing of HCV-533, normally maintained open by
gdministrative control, during the filling opera-
fion

Conseguences

of itself, a hazard; & release of positive pres-
may produce a sudden rise in fuel level during
L and, possibly, & nuclear excursion

 

Corre¢ctive action

Draiq fuel

Caus¢ 8

'Rupture or leak in the primary containment

 

 

Consé uences

nation of secondary contalnment and increased
posglibility of contamination of area in the event
tha} secondary containment fails

   
 

Corrective actions

a&. Drain fuel

.b. Close in-cell cooling aeir syatem vent (to stack)

valve

 

Leak}in primary heat exchanger

Consegquences

a. This is.a loss in primary containment

b. (ontamlnation of coolant salt, coolant salt loop,
€ components therein, with possible contamina-
tion of area in the event radiator or other cool-
ant system componehts fail

Corréctive action
Draiu‘fuel

Caumgg
Administrative

Conse¢quences

. Failpre to have rods cocked during f£ill is a loss

of protection against the fill accident which can
leed to a damaging temperature excursion

Corrective action

Prohjbit £illing

[

 

This uses same instrumentation as III (this tdble)

1.

2.
3.
4‘

'Testigg:

Redundancy: Two channels are used; one in the pump
bowl and the other in the overflow tank

Testigg See III, this table ~
Monitoring: See III, this table

Safety only?: Yes

Redundancy: Two independent channels, either will pro-
duce safety actlon

Testing: Complete testing of each channel is possible

Monitoring:
Safetx onlz?:

Certain system failures produhe alarms

Yes

During normal operation the coolant salt pressure in . : ,
the heat exchanger is greater than that of the fuel *
salt; in these circumstances the contamination of the
secondary ls that resulting from back diffusion through
the leak

Redundancy: , Two independent channels, elther will ini-
tlate safety acticn

Complete testing of each channel is possible
Certain system failures produce alarms

Monitorigg:
Sefety only?:

Yes
i

Redundancy: Action by apy two of the three rods affords
protection
Testing: This may be tested prior to f£illing

Monitoring: A system test immediately before filling
meets requirements

Safety only?: Seme as I (this table)

 
 

A

Condition or Situation Which Indicates

~ & Real or Potential Hazard

 

 

 

63

Table 2.2 (continued)

Causes of the Hazard, the Consequences, and
the Corrective Action

 

coedn

Supplementéry Information

 

IX. Supply pressure less than 30 psig
in helium line 500 which supplies
helium to all reactor cell com-
ponents, drain tank cell, and

fuel processing system

X. Activity greater thsn 20 mr/hr

in off-gas line

XI. Activity greater than 20 mr/hr

in helium line %o reactor cell
level sensore (bubblers in pump

bowl end overflow tank)

Reactor cell pressure greater
than 2 psig

XII.

=

1.

Causes
Loss of supply helium caused by:

a. Empty tank

b. Previous overpressure which operates relief valve
and breaks rupture disk

¢, Malfunction of pressure-regulating valve PCV- 5OOG
which mainteins supply at design-point value of
40 psig

Consequences

Possible loss of secondary containment (see 1b above)

Corrective actions

T ————

Block all hellum lines to reactor end drain tank cells

Causes

a. Charcoal beds not operating cbrrectlyf(overloaded,
see III sbove) or pump seal rupture allowing dis-
charge of activity to lube-oil system

b. Activity in the coolant salt loop

Consequences

Radioactive gases discharged up stack

- Corrective actions

a. Close off-gas block valves
b. Close lube=-0il systems vent valves

Causes

Reversal of flow in these lines (or back diffusion)
from pump bowl, drain tanks, overflow tanks

Consequences

Radioactivity inside the helium piping in normally
safe areas. This activity will be contained as
long as piping is not breached

Corrective action

Close in-cell block valves

Causes v

a. Maximum credible accident

b. Malfunction of cell pressure control system

Consequences !

Not, of itself, & direct hazard; during normsl opera-
tion the reactor cell is maintained at a negative
pressure of —2 peig (13 psia) to ensure inflow in the
event of a leak; the existence of positive pressure
is eyidence that a malfunction or misoperation exists;
a loss of secondary containment is a possible result

Corrective actions

a., Close instrument air block valves
b. Close liquid waste system block valve (from réac-
tor cell sump to waste tank)

4.

. Testigg

'Redundancy:

Safety only?:

Redundancy: Three independent channels, any two will
initiate safety action

Testigg: Complete testing possible
Monitoring: Testing meets requirements

‘Safety only?: Yes

a
.

)
i

i

s'

|
Redundency: Two independent channels, either will ini-
tiate safety action ,

'Complete test;ng possible
Monitoring:

Saféty only?:

Certain system failures are alarmed

Yes

 

x}

Floy rate, in either direction, is limited by capil-
laries; check valves, two in series, are used to back
upiblock valves

Redfindancz Three independent channels; any two will

, initiate safety action

Testigg: Complete testing of each channel is possible
Mofiitorigg: Not applicable

Yes

 

Sefety only?:

‘.
~l

Three independent chennels; any two will
initiate safety action

‘Testigg: Complete testing is possible
Monitor;gg: Not appliceble, testing meets requirements

Yes

il
 

Table 2.2 (continued)

 

Condition or Situation Which Indicates

a Real or Potential Hazard

auses of the Hazard, the Consequences, and
i ' the Corrective Action

*

Supplementary Information

 

XIII.

XIV.

XV,

Rediocactivity greater than 100
mr/hr in in-cell cooling water
system

Loss of coolant salt flow

Low coolant salt tempersture,
measured at radiator outlet

Cauj
Max;

teT

es

imum credible accident, which ruptures water sys-

Consequences

a. |Local contamination by radicactlve water lesk
b. |Loss of secondary containment, with accompanying
‘leontamination if radiocactive material is dis-

charged from the cooling-water system
Cau_i;gg
a. [Coolant pump stoppage
b. |[Line break C
¢. {Unscheduled coolant salt drain
d. [Plug in line or rasdiator
Congequences

Exc#pt for case VI1II ébove; no hazard exists; a flow

logs at full power will freeze the radiator in
2 min if no corrective action is teken

Corrective action

Drop radiastor doors

Causes

a. {Malfunction of load control system (complex of
doors, blowers, and bypass damper)

b, |Cessation of power generation in core from any

cause (scram, drain, rupture in primary contain-
ment, ete.)
c. |Loss of coolant flow (see XIV above)

Congequences

No flazard; warns that potential'radiator freezeup
may be developing (see XIV above)

- Cortective action

Drop radiator doors

.Redundancx: Three independent channels; any two will
initiate pafety action

Testing: Complete testing of each channel is possible-

Monitoring: Certain system failures initiate alarms
or produce safety action

Safetx only?: Yes

. ‘ . . ,

Redundancy: Two direct flow channels which receive
information from a common primary element, a venturi,
-plus two independent pump speed channels (Fig. 2.13)

Testing: Partial testing possible; a test which in-
cludes dropping radiator docors will perturdb operation;
input elements not tested '

 Monitoring: ' By surveillance and comparison

Safetx only?: Yes

|

Redundancy: .Threé independent channels; any two will
initiate safety action

Testing: A complete test of each input channel is pos-
sible; complete testing requires dropping doors, which
disturbs operstion

Monitoring: A thermocouple break or detachment from
pipe will produce safety action in that channel

Safety only?: Input channels used solely for safety |

 

3

 

 

-
 

 

 

65

Table 2.3.

 

ct
S
"

i
+

MSRE Safety System Output Actions

e st b A L 14 im0 g . 11

 

Safety Action

Results Produced

Supplefientary Notes -

Tnitiating

Condition

 

I.

II'

IIT.

VII.

( VIII.

Rod scranm

b ‘Close freeze valve cooling air

Resactor efiergency drain

1. Close freeze valve cooling air
" valves HCV-919A and HCV-919B

2. (a) Open bypass valves HCV-544,

545, 546

(b) Open vent valves HCV-573,
575, 577

3. ‘Close helium supply valve
PCV-517A1

valves HCV-909, HCV-910

5. Close helium supply valves
HCV" 572A-1-, HCV" 574Al »
HCV-576A1

Radiator door scram

Close helium valvyes HCV-572,
HCV-574, HCV-576

Close heiium supply. valve
PCV-517A in main header serving
all drain and flush tanks

Close helium supply valve HCV-SBO
to fuel storage tank

Close valve HCV-516 in helium
purge line to fuel pump

Close valve HCV-512 in helium
purge line to coolant pump

.Reduces temperature at which reactor is critical;
 system is shutdown until new (lower) critical

temperature 1is reached

.Provides shutdown margin and transfers fuel salt
- to a safe locatlion

Thaws freeze valve FV-103, main drain valve at
bottom of reactor vessel

.Bqualizes fill pressure in drain tank and pump

bowl pressure

Reduces pressure in drain and flush salt tanks by
venting to charcoal beds

Shuts off pressurizing helium (40 psig) in main
supply header to drain and flush salt tanks;
this valve is closed during reactor operetion

Thaws freeze valves FV-105 and FV-106, which
admit fuel salt to drain tanks 1 and 2

Shuts off pressurizing helium (40 psig) to indi-
vidval drain and flush salt tanks; these valves

~are closed during reactor operation

Shuts off air flow acress radiator to prevent
freezeup

Preserves containment; shuts off supply of pres-

surizing helium to individual drain and flush
tanks and blocks these lines to prevent escape
of radiocactive gases

Preserves containment; shuts off 40-psig helium
to drain and flush tanks and blocks escape of
radicactivity or reverse flow from the drain
tanks

Preserves containment

Preserves contalinment; blocks line 516 and pre-
vents escape of radicactivity from containment

Preserves containment; blocks line 512 and pre-
vents escape of radiocactivity from containment

Redundancy obta

 

Action by any two of the rods is sufficient;

rods tested at %ach shutdown

N

Complete emergency drain system may be
eration 1f test period

tested during

does not exceed 5 min

These valves ar
must cperate

Bypass valves ax
by administrati
closed to test

The vent and b:
drain tank fo

 
   
 
  
 

I a redundant palr

and with checkzvalves-

Administrative dpntrol is used to keep one
open during operation and
there are two empty drain

of these valve%
to ensure that
tanks to receiwe fuel

See II.3 above;:

closed and p
2 psig - g

Complete testin% possible but will perturb

pover generation

i

Redundahcy obta: ed in combination with V

(below) and with check valves

|
|

Redundancy obtajned in combination with IV

(above) and with check valves

1Hfiw_wvwe;efi§§

i

{

Q .
Bagic contalnment requirements met by this
valve and check valves in series; ESV-516A1

and 516A2 provide redundancy

Redundancy; 2.bérriers, provided by check

valves plus hesit exchanger

.

{ a redundant palr; only one

kept open during operation

e control and hence may be ‘4' Radicactivity 1

8s valves serving each

ed in combination with
individuel supply valves (see II.5 below)

%ay be tested during opera-
tion if PCV—511A is closed and if pump bowl
pressure reduced to less than 2 psig; PCV-
517A may be tested if these valves are
bowi pressure is less than

 

1.  Flux signal, po
Mw

2. Fuel salt outlet

greater than 13

1. High reactor cu

2. Fuel pump bowl
than 50 psig

3. Fuel salt overfl
greater than 2

tainment cell

5. Redloactivity i
coolant pump

6. During filling,

r greater than 15

   
   
   
  
  
  
    
  

temperature
O°F

et temperature

essure greater
w tank level
secondary con-
T

off.gas from

el pumfi bowl

pressure in excess of 2 psig

7. ‘During filling,
drawn

-

Low coolant sald
2. Low coolant salt

rods. not with-

temperature
flow

 

l. Low helium suppl
{less than 30

2., Fuel pump bowl
excess of 2 psi

3. All shim rods ng
drawn

1. Low helium suppl]

Yy, pressure
ig)
essure in

t with~

y pressure

(less then 30 peig)

2. Fuel pump bowl p
excess of 2 psi

3. A1l shim rods no
drawn

Low helium supply pr
than 30 psig)

Low helium supply pr
than 30 psig)

Low helium supply pr
than 30 psig)

regsure in
£

t with-
pssure (lees

pssure (less

basure (less

 

 

 
 

66

< Table 2.3 (continued)

 

XII.

XIII.

XIV.

XVII.

Safety Action

Block helium supply lines to bub-
bler (level instruments in reac-
tor cell)

1. Close valves HCV-593BL,
HCV-593B2, HCV-593B3

2. Close valves HCV-599El,
HCV-59932, HCV-599B3

3. Close valves HCV-595Bl,
HCV-3595B2, HCV-595B3

Close walve HCV-511A1 in helium
supply line (40 psig) to coolant
drain tank

Close HCV-557C; off-gas line from
charcoal beds, coolant salt pump,
coolant salt drain tank, and fuel

end coolant pump lube oil systems-

Close lube-oil systems vent velves
PCV-510A2 (cool pump system) and
PCV-513A2 (fuel pump system)

Close valve PCV=-565A1 in yent line
from reactor cell evacuation line

N

Close in-cell cooling water block
valves:

1. FSV-837-A1

2. FSv-846-A1

3. FSV-846-A2

4. FEV-847-AL i

5 . FSV— 844"Al

6. ESV-STmA

Close liquid waste system'block
valves

1. HOV-343A1 and HCV-343A2

2. HCV-333A1 and HOV-333A2

Close instrument air line block
valves

Opefi fuel pump bowl vent valve to

“auxiliary charcoal bed

‘Results Produced

Preserves containment; prevents reverse flow of

helium in lines to bubblers in tqe pump bowls
and in the overflow tank

 

Preserves containment

i
i

Preserves containment; blocks helium flow which
normally passes out off-gas stack

1

Preserves containment; blocks helium flow which

normally passes out off-gas stack
| | \

Preserves containment; blocks component cooling
alr flow which would normally pass out through
off-gas stack S-1

Preserves contalnment

Blocks outlet line from drain tank cell space
cooler

Blocks outlet line from reactor cell space
cooler No. 1

Blocks outlet line from reactor ce}l space
cooler No. 2

Blocks outlet line from thermal shdeld -and fuel
pump motor

Blocks inlet line to thermal shiel? and fuel
pump motor ‘

Blocks vent line from surge tank |

Preserves containment; blocks reaétor cell and
drein tank cell sump ejector disc rge lines to
liquid waste stor&ge tank

Preserves secondary containment

Preserves containment

Supplementary Notes

Redundancy provided’by two check valves in
series with each block valve

Redundancy firovided by charcoal bed plué

valves PCV-510A2 and PCV-513A2 (see XIT .
below)

Redundancy provided by valve HCV-557C (see
XI above) and by pump seals; these valves
- used for both control and safety

Redundancy provided by reactor vessel; this
valve forms part of secondary containment
barrier

' Redundancy of valves 1 to 6, inclusive,

provided by ESV-ST-A (6, below)

This valve provides redundancy per above

These valves normally closed during opera-
tion; 1 and 2 are both redundant pairs

Redundancy provided by pfimary containment

vessel walls

Initiating Condition

1. Low heliunm sfipply pressure (less

than 30 psig)
2. Radiocactivity in lines to fuel

salt level instruments (primary

sensors)

Low helium supply pressure (leés-fhan

30 psig)

Radloactivity in off-gas line

Radiocactivity in off-gas line

Radioactlvity in reactor cell air

Radiocactivity in cooling water system

Reactor cell pressure greater than
2 peig

Reactor cell pressure greater than
2 psig

Fuel pump bowl pressure exceeds 35

psig
67

Teble 2.4, Operating Ranges of MSRE Nuclear Instrument Channels

 

 

Fumber of Normal Operating
Channel Type Detector Type Channels Range
Wide-range Fission chamber 2 Source to five
counting times full power
Linear power Compensated ioniza- 2 15 w to 15 Mw
tion chamber
Flux safety Noncompensated 3 200 kw to 20 Mw

ionization chamber

 

The sensors used in the flux safety channels are uncompensated neu-
tron-~sensitive ionization chambers. The linear Tflux amplifier in each
channel produces an output signal of O to 10 v that is proportional to the
ionization chamber current. The gain of the flux amplifier and the cham-
ber locatlon is Tixed so that at design-point power the lux-amplifier
output is 5 v; the setting of the scram point of the fast trip comparator
is fixed at 7.5 v. During the early phases of MSRE operation, the scram
point will be reduced by changes in either gain of the amplifier or chamber
position or both. These changes are not routine and are not at the dis-
cretion of the operator; they will be accomplished under administrative
control and with suitable testing after they are made.

The temperature channels, described in the next section, monitor the
reactor outlet salt temperature over a range extending tc 1500°F. The
scram point is adjustable about the nominal 1300°F presently intended but
only under administrative control.

A scram channel combines one temperature and one flux channel to pro-
vide a trip 1if edither temperature or flux exceeds its scram point. The
three scram channels thus Tormed are then arranged so that signals from
two of the three scram channels are required to release the safety rods.
Relay contacts in the relay safety element are used as the logic elements,
with a separate two-of-three matrix being used for each rod. The coinci-
dence matrix monitors are used to display the operation of the relay con-

tacts in the matrices during operation, as well as during tests.
68

The flux and temperature safety channels are tested and monitored in
service in the following ways:

1. A voltage ramp may be applied manually to the system that causes
a current to flow through the leads from the ion chamber to the input of
the flux amplifier. Channel response is checked by observation of panel
meters.

2. A steady-state current may be applied manually to the input of
the flux amplifier. Channel response is observed on panel meters or, if
the applied current is large enough, it will trip the channel undergoing
test and the trip so produced may be verified by observing the operation
of relevant relays in all three relay matrices. The operation must reset
the scram circuit after a test or, in fact, after any scram signal. Reset
cannot be accomplished so long as the scram signal exists. On-line test-
ing of each temperature channel is effected in a similar manner, except
for the method of introducing the test signal (described in detail in a
following section). Since the flux and temperature safety channels are
in two-out-of-three coincidence, testing will not produce a scram.

3. A voltage loss at the ion chamber terminals will produce =z safety
trip in the channel. This monitoring action is automatic and continuous,
and a trip so produced is indicated by a lamp on the front panel.

Throughout the nuclear safety system the three instrumentation chan-
nels have been isolated from each other and from circuits used for control.
Theilr components are ldentified as éafety devices and, as safety devices,

are subject to the strictest administrative control in their maintenance.

2.2.2 Temperature Instrumentation for Safety System Inputs

Both rod scram and radiator door scrams require reliable temperature
input information. Figure 2.11 gives the details of one of the three
identical channels used to provide the required three independent tempera-
ture signals to the nuclear safety system shown in Fig. 2.10. The same
type of system is used for the temperature input signals to the safety
equipment that scrams the radiator doors.

The interconnecting wiring of each channel is in its own conduit,

and conduit runs are identified as part of the safety system. ZXach
69

UNCLASSIFIED
ORNL-DWG 64-626

CURRENT
ACTUATED SWITCHES

  
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

TEST ASSEMBLY: T SaF roay  CONTROL ROD
a. ELECTRICAL OSYSSTEENTY L REVERSE
r HEATER SEE FIG. 2.40 T0
| | b. TEST THERMQ- DATA
| |~ COUPLE LOGGER
| | (SEE NOTE 1.) URRENT
c
) | /ACTUATED
[ T SWITCH
Q
EmF_ TO 7 ISOLATION 7
CURRENT | AMPLIFIER 100 Q
MEASURING CONVERTER —
THERMOCOUPLE Lo ot
t ° 400 0
L0
* o & I—— 10-50 ma
10 70 50 me METER

 

 

 

 

CHANNEL NO.Z2

N0

SAME AS ABOVE

CHANNEL NO.3
NOTE:

{. POLARITY SHOWN FOR TEST
THERMOCOUPLE 1S THAT USED TO
SPARE TEST SYSTEM RESPONSE TO A
: TEMPERATURE INCREASE

 

 

e REACTOR OUTLET PIPE
LINE NOQ. 100 AND RADIATOR
OUTLET PIPE, LINE 202

Fig. 2.11. Typical Temperature-Measuring Channel Used in Safety
bysten.

channel is monitored, logged, alarmed, and pericdically tested during op-
eration.

Monitoring of the safety thermocouples is provided by the burnout
feature of the emf-to-current converters and comparative survelllance of
the three channels. If a thermocouple breaks or becomeg detached from the
pipe wall, the burnout protection will cause the channel to fail toward
safety, i.e., upscale. Wire-to-wire shorts or wire-to-ground shorts at
points away from the thermocouple junction are detected by comparing the
readings of the three channels. Operational testing of the individual
channels is accomplished by use of a thermocouple test assembly (see Fig.
2.11), which consists of a small heated thermocouple assembly connected in
series alding with the reactor thermocouple. When the thermocouple heater
is off, the test thermocouple is at the same temperature throughout its
length, and its net generated emf will be zero; however, when the heater

is on, an emf will be generated that will add to the emf of the reactor
70

thermocouple and cause the alarms and control circuit interlocks associ-
ated with that channel to operate. The use of two-out-of-three coinci-
dence circuitry permits individual channels to be tested without initiat-
ing corrective safety action. The thermocouple test unit described above
satisfies the requirement that operational testing shall not disable a
safety channel during the test and has the advantage of simplicity, re-
liability, and low cost.

The radiator thermocouple test is similar, except that the test
thermocouple is connected in series opposition so that heating the thermo-
couple reduces the net emf and simulates a temperature reduction in the

radiator outlet pipe.

2.2.3 Radiator Door Emergency Closure System

The radiator doors are dropped automatically if conditions indicate
that there is danger of freezing the salt in the radiator. An analog
computer calculation has shown that a loss of coolant salt flow, with
the reactor at full power, would freeze the salt in a radiator tube in

3 The calculation assumed completely stagnant salt

not less than 44 sec.
in the radiator and made no allowance for time required for flow decelera-
tion or for natural circulation if the particular situation permitted it.

Two types of input information (see Table 2.2) are used to initiate
closure. The low-temperature signal, measured at the radiator outlet, is
the primary indication that remedial action is required, regardless of the
cause. On loss of flow, where temperature measurements become meaningless,
the doors are dropped on low-flow signals from (1) the venturi meter in the
coolant salt loop and (2) the pump speed monitors. (The temperature input
information safety channels are described in the preceding section. ) Also,
individual radiator tubes are equipped with thermocouples which are con-
nected to the temperature scan and alarm system.

Coolant salt flow is measured using a single venturi in the coolant

salt pipe loop. With the exception of the venturi, two input channels are

provided. 1In a typical channel, differential pressure is converted into

 

33. J. Ball, "Freezing Times for Stagnant Salt in MSRE Radiator
Tubes, " internal ORNL document MSR-63-13, April 19, 1963.
71

an electrical signal by a NaK-filled transmitter followed by an emf-to-
current converter whose output is used to actuate alarm switches that
serve as trips to actuate the clutch and brake in the drive mechanism.

Coolant pump speed is measured by an electromagnetic pickup mounted
very close to the pump shaft coupling. Teeth, similar to gear teeth, are
cut on the periphery of the coupling, and the speed pickup generates a
voltage pulse as each tooth moves past it. The count rate, as determined
by the speed monitor, is thus an indication of pump speed and is used as
an input for alarm and safety.

The two flow signals and the speed signal are fed to a two-out-of-
three coincidence matrix such that 1f any two reach unsafe values they
will cause the doors to drop.

The two flow channels are monitored by comparison, one with the other.
A loss of flow signal is simulated by shunting a resistor across one leg
of the strain-gage bridge in the pressure transmitter and observing the
action of the relays in the coincidence circuitry.

The pump-speed channel is fested by using the calibration switch on
the speed monitor and observing relay response in the coincidence cir-
cultry. Oscilloscope display of the input pulses affords a check on the
integrity of the sensors and the wiring to the speed monitors. Low speed
is annunciated. The combined system is diagrammed in Figs. 2.12 and 2.13.
Interlocks are used to turn off the blowers when a safety signal to close
the radiator doors is received. The radiator heaters are maintained "on"
at all times when there is salt in the coolant loop. Heater current is

monitored by operating personnel as a routine administrative procedure,.

2.2.4 Reactor Fill and Drain System

 

A simplified diagram i1s shown in Fig. 2.14 of the reactor vessel,
the drain tanks, the interconnecting piping, and the control elements re-
quired to fill and drain the reactor system with fuel salt in a safe,
orderly way. The control rods are ecssential to a safe filling procedure
but are omitted from this diagram in the interest of simplification.

The reactor is filled by applying helium pressure to the gas space in

the selected drain tank and forcing the molten fuel salt up and into the
72

UNCLASSIFIED
ORNL-DOWG 63-8390

 

ELECTRO-

ELECTRO- /MECHANICAL BRAKE

MECHANICAL / OVER-RUNNING
CLUTCH

 

Iuj IUJ// SHEAVES —— CLUTCH: - d’//FLYWHEEL
A\ie

| K K | B

| £ \

 

 

 

 

    
 

=
X0

 

 

 

 

ORIVE MOTOR WITH
GEAR REDUCER. SINGLE
MOTOR USED TO DRIVE
BOTH DOORS~—

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   
    
   

SPRING SHOCK
ABSORBERS

 

 

A8

 

NOTE:

THIS ARRANGEMENT 15
TYPICAL AND IS USED FOR
BOTH DOORS

RADIATOR DOOR
1770 b

Fig. 2.12. Radiator Door Emergency Closure System.

core vessel. As a typical example, if the reactor is to be filled from
fuel drain tank No. 1 (FDT-1), pressurizing helium is admitted via lines
517 and 572 (see Fig. 2.14). The filling pressure is controlled by pres-
sure-regulating valve PCV-517A. The pressure setting of this valve is
controlled by the operator. The upstream capillary in line 517 limits the
maximum flow rate in this line., Valves HCV-544 and HCV-573 in pipes which
connect the gas space in FDT-1 to the pump bowl and to the charcoal beds
are closed., Since the net pressure head available to produce flow de-
creases as the fuel level rises in the core vessel, the fill rate becomes
progressively slower as the fill proceeds if constant inlet helium pres-
sure is maintained and if the pump bowl pressure remains constant.

Helium pressure in the pump bowl is the second component of the dif-
ferential pressure that drives fuel salt into the core vessel. A sudden
reduction in pump bowl pressure during the reactor filling operation would
cause an unscheduled rise in salt level in the core. If, at this time,

the reactor were on the verge of becoming critical, an unexpected nuclear
 

 

 

 

 

COOLANT
PUMP -
MOTOR :
o MAGNETIC PICKUPS PUMP, SPEED]| S Z#
INSTALLED ;// . - r—’ MONITOR
SPARE, [veer}
— ] o
| PUMP SPEED
MOTOR TO PUMP _ MONITOR
SHAFT COUPLING VENTURI LINE 201

 

 

  

 

 

pume sowL (___

 

 

7

 

 

 

 

 

 

 

 

 

LINE 202

 

HEAT -
EXCHANGER

 

 

THERMOCOUPLES —____ |

 

 

 

 

  

DIFFERENTIAL |
PRESSURE TRANSMITTERS .

 

 

 

 

 

 

SAFETY _ |
CIRCUITRY .
$10R Sp Sy OR. Sy
* INQEZE?QENT COINCIDENCE ,
A : Fy RELAY -
et {0)2 FLOW — ul MATRIX
F (b} SPEED £,
il —e]

 

 

 

 

__—RADIATOR

 

. CIRCUITRY

SEE
FiG. 2.1

 

 

 

 

 

 

 

 

 

U

 

 

 

 

 

 

 

THERMOCOUPLE

 

 

 

 

 

oo
CURRENT
ACTUATED

SWITCHES

 

 

 

]

  

 

 

 

 

 

3
EMF TO CURRENT
CONVERTERS |

 

 

Fig. 2.13.

 

|

 

 

 

 

iy
H
|

     
 
   
 
  

 

‘--—*—}3
A"

 

 

 

 

 

 

w

 

UNCLASSIFIED

ORNL- DWG 63-8389

 

 

 

 

\'.

 

SIGNALS

COtNCIDENCE]

RELAY

 

 

MATRIX

 

T
INDEPENDENT [ ~ 15 ]
TEMPERATURE {_Tg_ (2 OuT OF 3)

(—

 

 

NOTE:

EMERGENGY GLOSURE PRODUGCED BY INP

SIGNAL COMBINATIONS AS FOLLOWS:
(a) S, OR S, + F,
(b) S, OR S+ F,
(e) Fy+ Fy

(d} ANY TWO OF THE THREE TEMPE,RATUR'E

INPUT SIGNALS

 

. 4

ALARM
- .

 

Pump-Speed Monitoring System.

L= 70 READOUTS
—= LOGGER ETC,

 

 

UT

 

.

 

 

CLUTCH

SEE FIG. 2.12;
)

CLUTCH

 

BRAKE

UPSTREAM DOOR

=2

3

E

 BRAKE

 

DOWNSTREAM DOCR

 

 

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

74
System Conditien
PresFill Gpocete
Yeive No. o Norm Draln Inte
with  Clre, Nermst  Fill feam Fill from Fill from Ewerp.
Helium Ha Opsrstion FOT#1 FOT#2 FFT  FOTAl  FDT#2 FFT Dewin
PCY S, Fill prons o x x 0 0 o x X X
HCV 873, Fill, FOT »1 x X x o x x x x X x P
HCV 574, Fill, FOT #2 x x % x o x X x X x ‘ FUEL PUMP
HCY 378, Fill, FFT X x X X X 0 X X X X f—\ ‘ BOWL OFF-GAS
HCY 344, By pass o o o X 0 ° o X E o - .5 (3%2) 320+ HOLDUP AND
HCV 345, By po 0 x E E o ' P f— H i
HCV 546, : :.:: g g :. 0 0 : E g o o INLET}HELIUM ) @ COO_LER
NCY 573, V. X x x x x x o. o LINES - - l i J_'_i 140 liters
, Yeat ‘ SEE|FIG. 217 =TT 1
, Vo X x X PT
HCY 577 Ve X ox o x xax . % : > 539 600 : — HCV 523
HCV 533 Al o x 0 o o x X X x : $ (589)
°
A S T A | ——e— || | ] _
HCY 910 Al o o o X o X X 0 X X | ; ‘AR
HC 519 Asn' o X x X x x x x x x 1 : | OVERFLOW (
Hev 528t 0 0 o o 0 o . o o o o ‘ | ! ! Loy TANK
pevsna gl 5 5, 0 o 0 o o o o o INPUT TO — —= EMERGENCY DRAIN
e S xke % o cx X 8 1% . | SAFETY SYSTEM | __ OPENWCV___ . _ . -
FV 106 X X OwX* 0O XX ° X x o ! SEE FIG. 2.15 533 "
iy 07 x x X x X X X X ; ‘
:: 108 : : : x X X X X x X X BYPASS LINE =20 fs—s?
FY 109 : X x X X X X X X X - N
X -:lou‘:o-o;-n;‘! = sither span or cloasd, . . '_
*During nermel oparation wither, (but net beth) ¥V 105 or FV 106 is apen. . . )
;HCV 518 AL B eponed after comploting fill and betore sporation to ensure thet drein line 103 is clesred, )
. THCY 523 clesed enly when H-‘rln‘ solt gut of averilow tenk, BYPASS
. BYPASS
INITIAL !
LOADING V REACTOR
CONNECTION M) VENT =5 VESSEL
' LINE HCV (
HGV HoV ST
. 546 3
HCV ° &19
x HCV  HCV
} Py FV103 919A 9198
{103) 7
; FV103 ; io
TO FUEL . FV106 FVIOS
1 FLUSH TANK @ v I
: }
Fvi08 TO FUEL _ FV107
e
D . (108)- TRANS. AND <ep——DD)=u(107)
TORAGE TANK
FV 109 B . Y S
o "I” 10 o, HeV T At 0
WASTE SALT - = ' 9104 933X1 7 —l
DISPOSAL . FEa
578 G
li L gl @ r""—'-"“. 2 ’———”—*-.
: | | 1 PT 1
| | & ] | e T = |
FUEL w w | 908 Al FUEL |
| Sl TANK |
lDRAIN > Jla | No2 |FLUSH,
CAPILLARY ‘ Tfig:‘ Sl So | ‘ TANKl 1
FLOW RESTRICTOR o\, | Nt o T z ! o | |
HELIUM WE st | i 3 > 576 S
x HIx & X
SUPPLY, w g
40 psig :
{(575)
FREEZE VALVE

5198

5194

 

Fig. 21

 

 

 

 

 

 

COOLING AIR

14. Reactor Fill and Drain System Valving

 

   
 
   

UNCLASSIFIED
ORNL-DWG 63-8395

PCV
522 A4

TO
CHARCOAL
BEDS

TO AUX.

(G61)~= CHARCOAL
BEDS

f FREEZE

VALVE

Lcoon.ms

AlR
75

excursion could not be ruled out. The safeguard provided is tc allow
filling only when the positive pressure in the pump bowl is equal to or
less than 42 psig and thus keep the maximum possible increase in net Till-
ing pressure within safe limits. Two causes Tor a sudden decrease 1n pump
bowl pressure are considered. First, opening valve HCV-533 after reactor
filling has started would vent the pump bowl to the auxiliary charcoal
beds, which operate at close to atmospheric pressure. The resultant pres-
gure change in the pump bowl would be -2 psi. Administrative contrecl isg
used to maintain HCV-533 open just before and during filling. Second, a
rupture or leak would allow the escape of pump bowl helium to the reactor
cell stmosphere, which is held at =2 psig (12.6 psia). This condition
would cause a maximum pressure decrease in the pump bowl of 4 psi. If

the fuel pump bowl pressure exceeded 2 psig, the safety equipment (1)
would initiate a drain by opening the bypass valves, HCV-544, -545, and
-546, and equalizing helium pressure in the drain tanks and pump bowl and
(2) would shut off the supply of pressurizing helium. This is the helium
valve condition shown on Fig. 2.14 and is required by normal operation
with the pump bowl at 5 psig; therefore, the 2-psig channel need not be
disabled during normal operation. Additiocnal safety considerations in-
volving pressure in the fuel salt system and the instrumentation used for
measuring helium pressures is discussed in the following section.

During reactor operation helium pressure in the pump bowl is main-
tained at 5 psig by means of throttling valve PCV-52241 (Fig. 2.14),
which controls the flow rate of off-gas from the pump bowl. This valve
is actuated by a signal from a pressure transmitter, PT-5224, on line
592, which admits helium to the pump bowl.

Excessive pressure (50 psig) in the pump bowl is relieved by opening
HCV-533. During reactor operation, with the bypass lines open, one or
more of the drain tanks would be subjected to this pressure. Considering
the gas volumes involved, a 50-pslg overpressure in these circumstances
is of questionable credibility. The input signal used to open HCV-533
originates in either of a redundant pair of pressure transmitters, PT-592B
or PT-589B, on inlet lines to the pump bowl and overflow tank, respec-
tively. Thege same input channels provide the 42 psig safety signal dis-

cussed in the preceding paragraph.
76

As a Turther safeguard during filling, the system requires that all
three control rods be partially withdrawn in order to pressurize the drain
tanks. The weigh cells on the drain tanks provide information used to
monitor the filling rate and the total amount of fuel salt moved into the
core vessel.

The possible filling accidents are discussed in Section 7.1.4.
Briefly, these accidents are (1) premature criticality during filling
caused by an overly high concentration of uranium brought about by selec-
tive freezing in the drain tank; (2) premature criticality during filling
of low-temperature (200°F) fuel salt of normal concentration, whose normal
critical temperature is 1200°F; and (3) premature criticality during fill-
ing of normal fuel salt at normal temperature with all control rods fully
withdrawn. Protective action is the same for all three cases: (1) the
control rods are scrammed, and (2) the reactor vessel is drained to ensure
permanent shutdown. Rod scram is produced and the vent valves HCV-573,
-575, and =577 are opened by the excess flux signal. ©Since the criti-
cglity would be occurring before loop circulstion was attained, the cut-
let temperature sensors would be ineffective. The safety system also
invokeg a reactor emergency drain to enhsznce ccntainment if there is evi-
dence that radiocactivity is escaping from the primary fuel loop. These
subsystems are covercd in g subsequent section.

Emergency drainage is effected by the following actions (see Fig.
2.14):

1. The freeze valve in line 103 which connects the reactor to the
drain tanks is thawed. Thawing is accomplished by closing valves HCV-919A
and -919B, a redundant pair, to stop the flow of cooling air. The systen
is designed with a heat capacity sufficient to thaw the plug ixn 1> min.

2. The helium pressures in the fuel drain tank and the uniilled
portion of the fuel salt loop are equalized by opening bypass valves HCV-
S4de, =545, and -546.

3. Vent valves HCV-573, -575, and -577, which release pressurizing

helium in the drain tank to the off-gas system, are opened.
7'

4. Pressure regulating valve PCV-517A1 in line 517/, the inlet header
that supplies pressurizing helium tc all the drain tanks, is closed and
shuts off the supply of pressurizing helium.

5. The drain tank pressurizing valves HCV-572, -574, and -576 in
the helium supply lines are closed to halt further addition of filling
pressure.

Acticons 2 or 3 are immediately and independently effective in revers-
ing a fi1l1l, and hence the valves are redundant. Similarly, PCV-517A2 and
the individual pressurizing valves, 4 and 5 above, form series pairs and
provide redundancy. For example, it can be seen from Fig. 2.14 that,
taking fuel drain tank No. 1 as typical, pressure equalization and vent-
ing are accomplished by opening HCV-544 and HCV-573, respectively. Ad-
ministrative control is used to ensure that, when the reactor is filled
with fuel salt, both drain tanks, FDT-1 and FDT-2, are empty and that one
of the freeze valves, FV-106 or FV-105, is thawed. The tank condition is

monitored by reference to the weigh cell and level instrumentation on each

tank.

2.2.5 Helium Pressure Measurements in the Fuel Salt Loop

One channel of the instrumentation which measures helium pressure in
the fuel pump bowl and the overflow tank is shown in Fig. 2.15. The com-
ponents and their installation are designed to meet contaimment require-
ments. Since the secondary containment is held below atmospheric pres-
sure and since the temperature therein is not closely controlled, it is
necessary to use a variable-volume reference chamber to maintain atmo-
spheric pressure on one side of the measuring bellows in the transmitter.

Two channels are used: one in the pump bowl and the other in the
overflow tank. These normally operate at the same pressure. Any safety
signal from either channel initiates the appropriate safety action. This
is discussed in the preceding section and outlined in Tables 2.2 and 2.3.

The system may be tested periodically during reactor operation by
(1) Observing system response to small operator-induced pressure changes
and (2) by shunting the torque motor in the pressure transmitter. These

tests will establish that, in the channel undergoing test, the lines from
NOTES:

i, OPERATION OF THE TEST SWITCH SIMULATES THE APPLICATION
OF PRESSURE TO THE TRANSMITTER AND GCHECKS ALL
COMPONENTS IN THE SAFETY CHANNEL EXCEPT THE
CONVERSION OF PRESSURE TO FORGE BY THE TRANSMITTER
BELLOWS.

2. THE 2 psig SAFETY CHANNEL IS IN THE ACTUATED MODE
DURING NORMAL REACTOR OPERATION

 

 

UNCLASS!FIED
ORNL-DWG 64-627

TO SAFETY
SYSTEM
SEE F!G. 214 AND
TABLE 2.2
,SINGLE ALARM
SWITCH, SET AT
40 psig
~DUAL ALARM SWITCH
7 SWITCH SET TO PRODUCE
2 SAFETY INPUT SIGNALS

    
 

   

\

 

aLarM |
—

 

 

 

 

 

 

 

 

 

 

VENT TO
STACK CONTAINED. ZERG 1 IF PRESS IN LINE EXCEEDS o
e i | . . .
- , AN
psig REFERENGE J%N‘ | oW 2psig AND 50 psig J
\ . CHAMBER. RATED | g2 N TO DATA
N AND TESTED AT 100 psi | S0 psig; 2ps LOGGER FROM FUEL SALT
SLACK M ! ps19 T OVERFLOW TANK PRESSURE
DIAPHRAGM ~ l . o TRANSMITTER
L TEST SWITCH , o ~
rrfTer,, 1 | SEE NOTE f o] S 5~ METER
E A [=10TO 50ma = 4000 ‘~/)1
' 1 E R ~ =10 TO 50 ma
2eh T v
T WELD SEALED PRESSURE " ISOLATION
AMPLIFIER

 

INLET
HELIUM *

 

592>

 

 

TO PRESS
TRANSMITTER
USED FOR CONTROL

 

TO OVERFLOW TANK,

 

INLET

TRANSMITTER WiITH PRESSURE-
TO-CURRENT TRANSDUCER-AMPLIFIER,
RATED AND TESTED AT 250 psig

REFERENCE LINE TO
FUEL PUMP BOWL

TO FUEL PUMP

 

 

INSTRUMENTATION
SAME AS ABOVE

 

HELIUM

Fig. 2.15.
the Primary Loop.

BOWL GAS SPACE

Typical Instrumentation for Measuring Helium Pressures 1in

84
79

the pump bowl (or overflow tank) are clear and that the electronic equip-
ment and associated wiring is capable of operating the output relays.
Since it takes 5 min to thaw drain valve FV-103, this time can be used to
observe the response of the thermocouples on the freeze valve as 1t heats
up but before actual salt flow begins. The valve can then be refrozen
before an actual drain is initiated. In such a test, valve HCV-533 will
be opened, and the regulated pressure (5 psig) in the pump bowl will be
lost for the duration of the test. The response of HCV-533 can be noted
by observing the actuation of the position switch on the valve stem. This
test procedure checks the entire safety channel, except the ability of
the transmitter measuring bellows and the associated linkage to transmit
the pressure.

Monitoring of these channels is accomplished by indicating and alarm-
ing the pressure downstream from the hand throttling valves in each helium
line serving the primary elements (the bubbler tubes) in the pump bowl
and in the overflow tank (see Fig. 2.17 in sec. 2.2.7). A downstream
flow stoppage by a blocked line is indicated by a high-pressure alarm;
low flow caused by a loss of upstream (supply) pressure, an upstream
blockage by foreign matter, or a hand valve closure is indicated by a
low-pressure alarm.

The fuel level in the overflow tank is measured by the differential
pressure across the helium bubbler probe which dips into the fuel salt in
the tank. The degign criteria for testing and monitoring these differen-
tial-pressure-measuring channels are the same as those which guided the
design of the pressure channels described in the preceding varagraphs.

The reference chamber is not required. Two channels are employed for
safety and either will initiate a reactor drain if the fuel salt level in

the overflow tank exceeds 20% of full-scale level indication.

2.2.6 Afterheat Removal System

 

The drain tank afterheat removal system, typically the same for both
drain tanks, is shown in Fig. 2.16. Once placed in operation the evapora-
tive cooling system is designed to be self-regulating and to operate with-

out external control., Reliable operation of the afterheat removal system
80

requires (1) that the feedwater tanks contain a supply of cooling water,

(2) that an ample supply of cooling water is available to the condensers,

and (3) that the system includes reliable valves to admit feedwater to the

   

FROM
COOLING
TOWER

 

COOLING WATER
TO OTHER
COMPONENTS

TO

CONDENSATE

STORAGE TANK

HCv-882
c

<

 

      
 
  
 

FROM
PROCESS
WATER MAIN

TOWER

      
    

70
COOLING =—@56T——

 

DRAIN

TANK
CONDENSER
NO.1

   

 

 
   
  

 

 

 

 

 

 

 

Fig

 

UNCULASSLEFIED
ORNL—-DWG 63—8388

TO
OFF-GAS
STACK

RUPTURE

g/ DISKS

   
 

MOTOR
OPERATED
VALVES

HAND
VALVE

 

 

 

560 TO VAPOR CONDENSING
| SYSTEM
o a
812 = TO DRAIN TANK
~ CONDENSER NO.2
FEEDWATER
TANK NOA <
®
878 <
— CONDENSATE 3
T ¥ ESV-806 0
PN

 

 

 

T 7 et A
L LT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

iy
]
| |

| A STEAM
| x' ORUM |
i 1N f
=
E -—g‘j - ——-—-‘-HA \
::v' S
DRAIN TANK S
CELL= — ~wy k

i’/‘/

; FUEL é

- DRAIN
- TANK |
7 .,

. 2.16. Afterheat Removal Systemn.
81

steam drums. Administrative control is relied on to ensure that the feed-
water tanks contain water. Valve ESV-806 opens to admit water automati-
cally to the steam drum when the salt temperature in the drain tank exceeds
1300°F and recloses at some lower temperature to be determined experimen-
tally. This valve is in parallel with manual valve LCV-806A, and the pair
are a redundant means of valving water to the steam drum., Normally the
condensers are cooled by tower cooling water, but diversion valve HCV-882Cl
provides an alternate supply of condensing water. Loss of tower water is
detected by pressure switch PS-851B, which operates diversion valve HCV-
882C1 and supplies condenser cooling water from the process water main.
Since it takes over 12 hr for excessively high afterheat temperatures
to develop after a drain, there is sufficient time to effect a transfer
to the other drain tank in the event of a failure or malfunction. This
is also ample time to make connections from the ccoling water system to

a tank truck in the event that the normal water supply is i1noperative,.

2.2.7 Containment System Instrumentation

 

Contaimment requirements are met by providing at least two independent
reliable barriers, in series, between the interior of the primary system
and the atmosphere. For example, the two-barrier concept is fulfilled by:
1. Two independent, reliable, controlled block valves with independent

instrumentation.

2. One controlled block valve plus a restriction such as a charcoal bed
which will limit the escape of activity to the stack to less than the
maximum permissible concentration.

One controlled block valve plus two check valves.
solid barriers (vessel or pipe walls ),

One solid barrier and one controlled block valve.

One solid barrier and one check valve.

o W oW
=
O

The general considerations outlined in this section apply to the
instrumentation and control equipment used to operate the block valves.
Rlock valves are not located at such a distance from the containment pene-
tration that the lines become tenuous extensions of the containment ves-

sel., Valves and other devices used in lieu of solid barriers will be
82

routinely tested and demonstrated to be capable of maintaining leakage
below the specified tolerance when closed.

Helium Supply Block Valves. A diagram is presented in Fig. 2.17

 

that shows the helium supply lines tc the primary containment vessel and
the associated valves used for contrcl and blocking these supply lines
against the escape of radicactivity from the primary system as a result
of reverse flcw or back diffusion. Two types of input signals are used
tc initiate block-valve closure (see Tig. 2.9)., The first, a reduction
in the supply pressure from its normal value of 40 psig to 30 psig actu-
ates pressure switches and closes all the inlet helium bleck valves. The
second, excess radiatior in any of the helium lines supplying the level
probes (bubblers) and pressure-measuring instruments in the pump bowl and
overflow tank, closes the block valves in these lines. A reduction in
helium supply pressure in line 500 from its regulated value of 40 psig
indicates a leaky rupture disk or leaky piping and a loss of primary con-
tainment.

In-service testing cf the loss of pressure channels is accomplished
by cpering the hand valves on the lines to the pressure switches on the
main supply pipe (line 500), one at & time, and observing the action of
the relays irn the twc-oui-of-three coincidence matrix in the control room.
Actuzl block-valve closure is not tested with this procedure. Low- and
high-pressure alarms are prcvided on line 500; these will actuate before
the safety system pressure switches close the blcck vaives.

Additional testing of the solenoid block valves in the helium bubbler
lines is provided by clesing each velve individually and cbserving the
pressure change downstream of the hard throttling valve,

The three radiation-monitorirg safety input channels, RM-596A, B,
ard C, are tested by exposing ezch irdividual radiation element to a source
and noting the response ¢of the cutput relays which produce valve closure.
Since two-cut-of-three coincidence 1s used, this test willl not perturb the
system; nelther dces it provide a valve-clcsure test.

Off-Gas System Monitoring and Block Vslves. The off-gas systemn,

 

Figs. 2.18 and 2.19, is monitored for excess radiation in four places:
  
 

 

 

83

 

 

TO FUEL

 

 

 

 

 

 

 

 

 

   

. TO
psig ALARMS

-

 

 

 

 

 

¥

UNCLASSIFIED
ORNL-DWG 63-8398

TO COOLANT SALT -
PUMP BOWL BUBBLERS

PRESSURE GAGE - AND :IPRESSURE SWITCHES WITH HIGH AND

TO LEVEL PROBES,
(BUBBLERS) IN FUEL
SALT OVERFLOW

| TANK.SEE FIG.2.14

-

TO LEVEL PROBES

{ BUBBLERS) IN FUEL
SALT PUMP BOWL.

- SEE FIG™2.14

 

 

   

 

 

 

. PUMP— > ° @ LOW PRESSURE ALAI‘\;M 1S TYPICAL OF ALL BUBBLER LINES
CESV % ESV : _ '
516 A1 § § si6a2 — & MoV
] * ¥ 59983
BI&—D<I—IX [ = (600 !- »
siom ¥ o f”’ / g~ Hev o
TO LUBE OIL < P D ) * 159982 |
X SYSTEMS FOR FUEL AND [P / X~ |
e o POV L COOLANT SALT PUMPS r - = HeV |
sI3A [ ! i ) * - s9981 |
x . oV . >} - == @ i K ’
I £t SHAl ggA?%O%:NNJ ' - f * q: so3e1 |
511 I~ e - . D<= (598 Y s 1
T | e o |
' * 1 59382
312 T~ - TO COOLANT PUMP >pg——G2 : pk L=
> _ * el W50 |
: : ETN ]
(50D -GoD /I X
- CLOSES : 7 o P
BLOCK VALVES -SEE NOTE 3 | (deos ) (522 )(522)
SEE NOTE | f : A B
Y 2/ ]
" TO FUEL STORAGE L 2 OUT OF 3 \ @ A
TANK AND CHEMICAL COINCIDENCE @ ‘ @

PROCESSING SYSTEM

SNUBBER

      
 

SNUBBER

r'_‘ 3 INDEPENDENT
~

b

 

SNUBBER

    

VENTED TO
OFF-GAS SYSTEM

DUCT (832 (TYPICAL)

EXCESS FLOW
RUPTURE DISK
@ [ s0esig

SIGNALS

' THREE CHANNELS
FOR INPUT SIGNAL.
PRESSURE SWITCHES
ACTUATE AT 35 psig

 

— — ——y —

 

 

 

 

 

—
HELIUM SUPPLY

REGULATED AT
40 psig -

 
 
   

 

 

>

PCV | .
; i 6054

 

MATRIX

 

 

 

PCV
*i S17TA
X

 

 

 

 

    

 

 

REFERENCE CHAMBERS SHOWN ON FIG. 2.5

 

 

 

 

 

%
4
1 1
1
INDIVIDUAL TESTING | CAPILLARY ,
OF EACH CHANNEL | RESTRICTOR v / AND ARE SECONDARY CONTAINMENT BARRIERS
BY OPENING HAND | * | f ‘ = - g
VALVES T mov DTN (572 TO FOT- .
- I 574 A1 | \ ‘ , A
2 OUT OF 3 E]; q & 2
COINCIDENCE Aoy RN &3 = TO FOT-2 |,
MATRIX @ 5T6A1 :
*
|F ——i—x =N G790 T0 FFT
ESV ‘
5194 :
CLOSES * 4 53¢ '
, 5198 TO REACTOR DRAIN .
BLSSJEcg VALVES < X pr——T"—T\———= LINE [NO.103) PURGE i
/ |
STACK ™ SECONDARY '
CONTAINMENT
"
’ |
RELIEF © NOTES: _ L
VALVE 1, AN EXCESS RADIATION SIGNAL FROM 2. A LOW HELIUM SUPPLY PRESSU)?E

PCV -
J,500¢

 

FROM
HELIUM
STORAGE

Fig. 2.17. Helium Supply Block Valving.

RE 596, (ANY 2 OF 3) CLOSES HELIUM
SUPPLY VALVES , HCV 593 B1, B2, B3,
HCV 599 Bi, B2, B3 TO BUBBLERS IN

FUEL PUMP BOWL.

4, THIS LINE, 588 GOES TO AUXILIARY

SIGNAL LINE 500, (ANY 2 OF 3]
PRECS. SWITCHES) CLOSES HELIUM
BLOCK VALVES MARKED WITH ).

ASTERISK, THUS % |}
i ,

CHARCOAL BEDS, LINE 56%, VIA DRAIN
TANK VENT VALVES. SEE FIGS. 2.14 AND 2.15

3. LINE AND ASSOCIATED
PXM'S IS A ZERO psig REFERENCE
LINE WHICH MAINTAINS
CONTAINMENT AND
TRANSMITTERS.
 

g T

 

 

 

 

 

 

 

 

 

 

 

      
   
    
  
  
   
  
 
   

 

 

 

 

 

 

) A ‘ , ( ' UNCLASSIFIED
L . - . ) ORNL-DWG 638397
- (560)
© O IN-GELL COMPONENT T —mwesToy
, COOLING AIR HEADER - /
' CC —
. . . \ .
=51 _ % oy REACTOR CELL
, - AIR COOLER : EVACUATION LINE
, NO.1 . I 565 Al y / ;
' A""v‘v"""““‘v - x
cor | Cee—— v | :
‘ NO.2 p ' ’ >
T PRESSURE BLOWERS e ‘
HELIUM FROM | ! :
FUEL DRAIN TANKS "= D -— i GHARGOAL BEps M D<HNH><H-EED | | OFF-GAS |
- AND SAMPLER-ENRICHER | e ‘ . STACK~—__
- : / : 1 ) .
HELIUM FROM FUEL, SALT CIRC. G221 Tank T oA B0 L >G5
PUMP, : L NOS. 1 AND2 J'F , A
| s @
: 13)
HELIUM FROM COOLANT SALT ~— G- N>

.GIRC. PUMP. ‘

     

ABSOLUTE:
FILTER

 

 

 

 

@7 -~

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

  
  

 

 

- : _ Sem—e—|——n
HELIUM FROM | HEY J
COOLANT DRAIN TANK | 557Ct
AND LUBE OIL SYSTEMS - 55D
NS
SAMPLER-
ENRICHER
I @D
- v
0 &0
3 T
- ) \
‘ :\ | | , AE 0 2
: 1 ) 3 £3 :: 1
J _ VENTILATION AIR ] / _ . S
: FROM: / ' B | < :
. : 1. LIQUID WASTE CELL "-:!a ;\J_h v N ,93'\\-/ I o THIS LINE USED ° .
' 2. DECONTAMINATION CELL I TT ¢! =~ . i ONLY DURING MAINTENANCE /
3. REMOTE MAINTENANCE [ — '
PUMP CELL
4, HOT STORAGE CELL |
5. FUEL STORAGE TANK : : VENTILATION
CELL J AIR' FROM REFESER F!Ss.o%qs o
: : SE VICE TUNNEL - FOR ILS OF LU
6. SPARE CELL OIL SYSTEMS
IN-CELL COMPONENTS o
rrrrre @39 Y
/ a
| oo c o |
H ANK
1 CELL 3
4B 1 > >~ v 000 coxoarrr
REACTOR [ELL /
VAPOR ;
_ PROCESS WATER —— CONDENSING ]
; TANK NO.2 °
VAPQ - FILTER 250
CONDEN ' LTE ctm
.o TAN
: vT

 

 

FROM RUPTURE DISK ON
THERMAL SHIELD COOLING
WATER LINE, SEE FIG. 2.22

F\ig. 2.18. Off-Gas System Instrumentation and Valving.

 

 
LOG

AND +——m

 

ALARM

~

COVER GAS
PRESS CONTROL
SYSTEM

85

 

 

 

 

 

 

   
  
 

 

UNCLASSIFIED
ORNL-DWG 63-83293

SIGNAL TO CLOSE
FROM RADIATION MONITORS
RM-557 A48 |

—— N3
-

 

 

 

 

 

 

 

 

 

 

 

 

 

| QUIGK
™~ | oisconnecT 99 -
THIS VALVE BLOCKED \\\T——————-— | A
WHEN HELIUM SUPPLY POy | ooy |
HELIUM |
SUPPLY (513) <} I
40 psig I
ALARM
.i..
LOG OIL SUPPLY | D
5 TANK FOR
MON FUEL SALT
RE Pugf1 G ~——708)-
QT8 = —(715)——<4——=—— EXCESS FLOW
RETURN FROM
PUMP OUTLET
ALARM
TC CIL
G1ar D PUMPS
713 e FOP-1
FOP-2
NOTE :
BOTH LUBE OiL SYSTEMS
ARE SIMILAR. NUMBERS
ENCLOSED THUS (O
ARE EQUIVALENTS IN
v COOLANT SALT PUMP
COOLING QUICK LUBE OIL SYSTEM
WATER DISCONNECT
Fig. 2.19. Lube 0il System Off-Gas Monitors.
1. Line 557, which carries off-gas from (a) all charcoal beds, lines

562, and 557, (b) helium from the coolant salt pump, lines 526, and

528, (c¢) helium from fuel and coolant salt pump lube-oil systems,
lines 534, and 535, and (d) helium from coolant salt drain tank and

line 560.

cates 1 (b) abovel.

Line 528, which carries helium from the coolant salt pump bowl [dupli-
3. Line 565, the reactor cell evacuation line.

5
4. Line 927, the main inlet line to the stack fans.

As shown on Fig. 2.9 and Tables 2.2 and 2.3, these different radiation
monitors do not have a common output. A high radiation signal from RM-
557A or RM-257B in line 557 blocks helium flow from the lube oil systems,
Fig. 2.19, by closing valves PCV-210A2 and PCV-513A2. It also closes
HCV-557C1l and blocks line 557, which is, in effect, a header carrying the
flows listed above. Blocking of the lube-oil-system off-gas is, there-
fore, redundant. A high radiation signal from the radiation monitors in
line 565 closes HCV-565A1 and blocks this line only and thus prevents cell
evacuation. A high radiation signal from the radiation monitors in line
528 carrying helium from the coolant pump initiates a drain. This particu-
lar monitor is deemed necessary becauge it provides an indication that
there exists a leak in the salt-to-salt heat exchanger (from the fuel salt
loop to the coolant salt loop). Such a lezak could put fission producte in
the coolant salt; however, when the coolant salt circulating pump 1s
running, the coolant salt pressure in the heat exchanger tubes is greater
than the fuel salt pressure in the shell., The fission-product gases would
be carried from the free surface in the ccolant salt pump bowl into line
528. Thig activity would be read by RM-557A or B or both shortly after
being flret read by RM-528A or B cor both. Blocking as described above
would then fellow snd the protecticon afforded against this heat exchanger
leak is thus redundant.

The systom ig tested by inserting a radiation source in the radiation
monitor shields and observing (1) the radiation-monitor indicator, (2) the
control circuit relays, and (3) the action of the control valves which
provide blocking. The three pairs of radiation monitors, RM-5Z&A and B,
RM-55%A and B, and RM-565A and B, all operate such that an excess radia-
tion signal from either channel in a pair will produce safety action.

Insertion of a radiation source in any one monitor will cause all
block valves asscclated with that particular monitor to close. Testing
of RM-565 or RM-52¢ will initiate a reactor drain, and unless the test

—

is completed within 5 min, the time required to thaw freeze valve 103,
87

an actual drain will be established. The other channels may be tested
for longer periods of time without affecting operations.

The stack gas monitor (item 4 above) on line 927 monitors off-gas
activity from all the other sources Just before the filtered off-gas
enters the stack blowers. Ixcess activity is alarmed in the control room
and another alarm signal is transmitted to the ORNL Central Monitoring
Facility.

In-Cell Liquid Waste and Instrument Air Block Valves. As indicated
by Fig. 2.9, the in-cell (secondary contaimment) liguid waste lines are
blocked if the reactor cell pressure exceeds 2 psig (17 psia). The normal
operating pressure is 13 psia. Excess cell pressure is a symptom of sys-
tem malfunction or, at worst, the maximum credible accident (see sec. 8.5).

The system and its operation are shown on Figs. 2,20 and 2.21. Three
independent pressure-measuring channels with trips (pressure switches)
are employed as inputs. The pressure-switch outputs go to a two-out-of-
three coincidence matrix. If any two input channels indicate a positive
cell pressure equal to or more than 2 psig, the block valves in the lines
from the reactor and drain tank cell sumps to the liquid waste tank are
closed., The instrument air lines into the containment vessel are blocked
at the same time. Since each pressure-switch contact is connected to a
different power source (see Fig. 2.21), any single power failure will not
produce blocking, and no single channel or‘component failure will disable
the systemn.

The pressure switches are mounted close to the cell wall and are
accessible during reactor shutdown. The pressure gages and hand valves
are mounted on a test panel located in an area which is accessible during
power operation. Fach pressure switch is connected to a separate tap on
the reactor containment vessel. Iengths of lines between the snubber,
hand valve, gages, and the switch are as short as is consistent with the
location requirements. Wiring in the three channels is run in conduits
separate from each other and from control-grade wiring.

The switches are connected toc a two-out-of-three solencid matrix in
the manner shown in Fig. 2.21. The circuit shown is a simplified schematic.

A relay will be required to operate other block valves,
SERVICE

 

 

 

 

 

 

PRESSURE
INDICATOR

SINTERED

 

 

 

UNCLASSIFIED
ORNL~DWG €3-8396

NITROGEN
———— SUPPLY
50 psig MAX

 

 

HAND
VALVES

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   
   

 

 

 

R DISK SNUBBER '
SECONDARY PRESSURE SWITCHES
/ CONTAINMENT, (- @ (ACTUATE AT +2 psig)
s 0@ " o3 SEE FIG, 2.21
\; REACTOR [ \\ /
(669) | CELL @ \
@_ ) 1 NOMINAL | F \ 2 OUT OF
""" T OPERATING B ~{ge-g/— — | 3 COINCIDENCE [~ —
' ceveL ! 1 ; - i PRESS.= | o 7’| RELAY MATRIX
» J [ 1 -2 psig (3 psia) / |
o TRANSM. (| 1 DRAIN | / A1 et / '
I TANK CELL A Qx 7| Pume /
. —bfl-—m {4 : RCF }
: s
— Gt ‘ JET TO CLOSE
100 psig | & PUMP o \SUMP |
AIR = o . : ‘
Teve, | : = c osss‘ss_ocx
‘ 1 .
PROBE (BUBBLER) m“ Sump oy LT T T VALVES IN INST,
’ - : FCV ’
PCV 3431 i 24322 ~ AIR LINES
33128 »< bt
<] e:l:’
- 332 = LIouiD.
! WASTE -
TANK
11,000 gal

/

 

 

 

,

Pig. 2.20. In-Cell Liquid Waste and Instrument Air Block Valving.

88

 
89

UNCLASSIFIED
ORNL-DWG 63~-8392

 

 

 

 

25 kva
AG EMERGENGCY
M-G SET DC POWER
+
PRESSURE
LINE FROM
REACTOR _—.| SWITCH —_—
CELL / (TYF’ICAL) / \
¢ m PSS PSS ; m PSS
. RC-B RC G \ . RG-F
w T/ T
RC- — - RC"
ESV ESV ESV ESV ESV ESV
1A1 142 181 182 1G4 1C2
-AC DC
NEUTRAL NEGATIVE

80-psig
EMERGENCY ey X
AlIR HEADER - ENERGIZED

' L
Pl : DE-ENERGIZED
25 kva 1A4 TVA 5 o
ESV
cov {0 e
A VALUE FLOW,
i TYPICAL OF ALL
B ESV'S
1A2 VENT v
—

TVA +DC
7. ESV
Esv (1] e F—y

-

Pl

 

 

 

 

 

 

 

 

183
COINCIDENGE
MATRIX OF
+DC Pl 25kva --—— VALVES OPERATED
fAS .y BY PRESSURE
Ecsv [tq; }@ ‘L'bJ‘ sAo SWITCHES
1C1

 

 

 

BLOCK VALVE i N
HEADERS TO

REACTOR AND

DRAIN TANK i

CELLS i —

VALVES
INST AIR
LINES

 

 

 

 

Fig. 2.21. Pressure Switch Matrix Used with Instrument Air Line
Block Valves.
90

The pressure channels will be tested by slowly opening the hand valves
in the nitrogen supply lines, one at a time, and observing pressure changes
in the pneumatic circuits that operate the block valves (for example, see
Fig. 2.20). If the nitrogen valve in the left hand line is opened, pres-
sure switch PSS-RC-B will be actuated and the actuating pressure may be
checked with the gage, PI-RC-B. As shown in Fig. 2.21, actuating this
pressure switch deenergizes two solenoid valves, ESV-1A1 and ESV-1A2, in
the six-valve matrix. The line carrying PI-1A5 will be vented tc the atmo-
sphere, and PI-1A5 will read zero.

The other two channels are tested in a similar way by observing pres-
sure gages PI-1A1 and PI-1A4, respectively. The tests do not actuate the
block valves, since the coincidence matrix requires agreement of any two
of the three input channels to vent the block valve headers to atmospheric
pressure.

In-Cell Cooling-Water Block Valves. A simplified flow diagram of

 

the in-cell cooling-water system, with instrumented block valves, is
shown in Fig. 2.22. The signal tc close the block valves 1s provided by
an excess radiaticn level in the pump return header, line 827. Three
independent sensocrs, 3=-827A, B, and C, are used that operate in two-out-
of-three coincidence; i.e., block valve operation 1s initiated when any
two of the input channels indicate excess radiation in line 827. This
system permits the loss of one sensing channel by malfunction or for
maintenance. In these circumstances, a signal from elther of the re-
maining channels will actuate the block valves. Operation of the system
is apparent from Fig. 2.22. Testing is accomplished during operation by
manually inserting a radiation source in the monitor shield and observing
that the indicated radiation level increases and that the proper relays
operate. The construction ¢f the sensor shield is such that each radia-
tion monitoring channel can be tested individually. Since the monitor
contacts are arranged in two-out-of-three ccincidence, testing of indi-
vidual channels does not close the valves. The valves will be tested
individually by operating a hand switch and observing that flow stops

in the line under test. The complete system can be tested during shutdown.
NOTES:

. ON EXCESS RADIATION SIGNAL FROM ANY

 

  

 

 

 

 

 

 

 

 

 
   
 
   

 

 
 

 

  
   
 

 

 
  
 
 
 
   

 

 

 

UNCLASSIFIED
ORNL- DWG 64638

 

 

 

 

 

   
 
   
 
    

 

 

 

 

 

 

 

 

2 OF 3 RADIATION MONITORS ALL VALVES BIES ;E;QL_%E__ . AN SOPPY
MARKED WITH ASTERISK, THUS (D% ARE CLOSED M 43 psig . 0 GAS
| T 828 COOLER
42 psig IN SPECIAL
| I \| © EQUIPMENT
2. RUPTURE DISK AND PRESSURE SWITCH PROTECT | ROOM
GAS PUMP
THERMAL SHIELD FROM RUPTURE IN THE EVENT | T'TJEA‘?FTE?D \\ | oL COOLERS
OF EXCESS WATER PRESSURE | cooLer /| A b
| - | - THROTTLING "/ ;W'
R —_—— e T/ vawes
51 psig J \
FROM COOLING / 3
TOWER /
PUMP Y ‘= o
NO. @ / z g
®2D o
wn
—@23 o
: o % o
\ 8 £ \3
F’@’;’ (835 58 psig PRESSURE gy :
SWITCHES o Z
SEE NOTENO.2 T - 25 3 o
RUPTURE _Lay © COOLANT
CHEMICAL 5K L SALT PUMP
ADDITION FROM SPECIAL FUEL PUMP . i ; ; MOTOR
o EQUIPMENT M \ i L £
T ROOM  COOLANT SALT
SPRING SO
LOADED ‘ -
HAND MOTOR l—m
VALVE —»
S NUCLEAR ]
ks, @ @ INSTRUMENT Ty
PENETRATION % N C .
C [ L= DRAIN
— | ‘ r' TANK
3 INDEPENDENT RADIATION L L Lo CELL
MONITORS RE 827 A, B, E,\R’E‘I\RE) y
ro- "AND G <] T AR | -
2 OUT OF 3 St X COOLERS < =7
r COINCIDENCE {————=——~~ v X R
| MATRIX
| 3 RELIEF .
; VALVES ]
L~ 10 BLOGK vaLvEs ~ 100Psa ] 1
SEE NOTE NO. 1 } THERMAL /
v SHIELD ~23 psig
{84&r
; P :], SECONDARY
- | (84— v CONTAINMENT
: @D ’, §
0 . !
WASTE TANK he e e . A

Fig. 2.22.

 

 

In-Cell Cooling Water System Block Valving.

 

T6
92

Valve ESV-ST-A serves as a backup to FSV-837, FSV-846, FSV-841, and
FSV-847 and provides redundant blocking in the system. The thermal shield
is protected from excess hydraulic pressure by interlocked blocking of the
inlet, line 844. The pressure switch on this line closes valve FSV-844A1.
The rupture disk protects against overpressure if a temperature rise takes
place when the thermal shield is isolated by the block valves or other
flow stoppage. Discharge through the rupture disk passes into the vapor-

condensing system and is contained.

2.2.8 Health-Physics Radiation Monitoring

The purpose of the radiation monitoring system is to provide personnel
protection throughout Building 7503 from radiation hazards due to airborne
and fixed-gource activity. The monitors composing this system are placed
at strategically located points throughout the building (sce Fig. 2.23).
A1l the monitors that measure area activity and airborne activity produce
signals which are transmitted to a central annunciator and control panel
locgted in the reactor auxiliary control room. These monitors also pro-
duce a building-evacuation signal based upon a coincidence of combined
monitor operation.

The criteria fcr the number, location, and arrangement of the Health
Phyeics menitors were established by the ORNL Radiation Safety and Con-
trol Office assisted by the reactor operations group. The locations of
the monitors in Building 7503 are shown in Figs, 2.23 and 2.24.

The contiruous air monitors (seven required) are located within the
building so that the air-flow patterns will sweep air past these units
tc menitor for airborne zctivity. Twe monitcrs are in the high-bay area,
one in the ccntrol rcom office area, and three In the basement working
area. A seventh instrument is a mobile unit for special operations.
Eight fixed mcnitrcns for lcow-level gamma detecticn are located in the
same general arrangement as the constant air monitors.

Very sensitive beta-gamma mornitcrs (GM-tube count-rate meters) ére
used in contamination free areas for clothing and body surveys. Two of
these devices are used for process monitoring, one in the cooling water

room and one in the vent house.
 

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    
 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

  

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

, ‘ . . i
\ ‘ N
m‘ 4
' |
N \ 93 . i
. : . ‘ UNCLASSIFIED
. . . f . ORNL-DWG 64-6209
T
! .
- o
. T . . BLDG. 7509 |
‘ ; ‘
N — pY i
fis} fi) \ |
. ¥ T I \8 T
] [ | :
) _ : & ‘ : ‘ DOWN et ;—1. L ||
=N L o P P A 5 |
L \2 u - — GUARD PORTAL g MAIN CHANGE ! |
. . FIRE HOSE o PANEL"B" - ROOM o
o—— 'A\__&—&& /7' °\ - . HOT CHANGE 4 4 ‘ : :
FIRE THROWOVER SWITCH FIRE ‘ __Roow : :
; , EXTINGUISHER bl EXTINGUISHER ik :
: o _ : - “FIRE T N _ = e | FILTER P
e . : EXTINGUISHER : ‘ e T : -
. g o LIGHTING A : - eI :
'. - PANEL"G" = ' . i
- o ‘ HOT CHANGE  T~LIGHTING PANEL"J" ' .
O— . — : oow 3 [T T | o
, C ' - 10-ton CRANE v : *: |
' FTT T[T TTT FIREHOSE= i ~ 0o ‘
N lFUIELI | o .
HOT STORAGE | 1ennSFER ! . : 0 _ :
Lo | ! R R R C o ‘
\ | I CELL, J _ _ _ o
S LL L e ojojooofg]ofo]olo a0 o
WFTTTW ol J ' o
1 v
b 4 | _~RADIATOR PIT - :
NATION. Srogy | | DRAIN TANK CELL & PENTHOUSE TOP :
Lt |y srorace | - : ELEVATION. < .
{11 1eew | ' 86111 Oin. ADSORBER -
ALl | 30=ton 0R ,
= CRANE «__ i
: - MAINTENANCE
ot ELEZ CONTROL ROOM Ly
FLOOR ELEVATION . .
! : 8621t Oin. -”- WEST SPACE COOQLER '
~ EAST SPACE COOLER '
A
)
_ , : SWITCH HOUSE : 1
‘ | ' o ’ ‘ DOWN—T ' F . ‘
: L o : EAN HOUSE i M AR MONITER |Q-2240
. * _ _ | - - i mmme MONITRON Q-H54
] . @ CRM (BETA AND GAMMA) Q-2091
_ W EVACUATION HORN
w | 1 ! @ ALARM BEACAN LIGHT
‘ o i 48 HAND AND FOPT MONITOR Q-t939
'.
o , T
. o . : i
A , ' - : |
D : i DIESEL HOUSE | Y
: ; !
) PLAN-ELEVATION 852t Oin.
- ‘ | ' PUMP NO. 2
: . gwmp NO.
! ' \
: e—COOLING TOWER
g
C | , L Fig. 2.23. Reactor Building (7503) at 852-ft Elevation Showing Locations of Monitors. -
F ’
b o

 

 

 

 

 
UNCLASSIFIED . S

 

 

 

 

    
  
  
 

    

 

 

 

 

 

 

 
   
 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

" ORNL-DWG 64—-6210
N -— -
® @ ® ® ®
uP - .
O—F ] : ' » -
BATTERY DOWN DOWN
ROOM FLooz’gNEL
533 # ’}E lfih ’
IN BREAKER e
&ng:grioc NO.2 . FIRE EXTINGUISHER
=S/'a7 l=0IST. NO. 2 - :
®&——uh o o D40-ton CRANE o b
DIST. NO. 1 @V e .
MG NO.2 AND NO.3 R
MOTOR AND GENERATOR ————— ——— MAIN BREAKER NO.{ \
48 volt DC
BATTERY CHARGING SYSTEM
DIST. PANEL 1A
©—) & [a
SPECIAL
MAIN TRANSFER EQUIPMENT O
SWITCH TO it x
EMERGENCY ROO
1 | GENERATOR
@— : 1 -
" | |
o I e e M Il
’:‘ S 1Y 7
Loy stacken BLOCK

COOLING WATER

 

 

 

o . L EQUIPMENT ROOM—T[
‘ »
Bl AR MONITOR. G-2240 - -
amme MONITRON Q-54 , — . . ,
@) CRM (BETA AND GAMMA)} Q-20Q1 . FAN HOUSE _
W EVACUATION HORN . ‘ RAMP DOWN
M ALARM BEACON LIGHT '

 

 

 

 

 

 

 

'PLAN-ELEVATION 840 ft Oin.

Fig. 2.24. Reactor Building (7503) at 840-ft Elevation Shdwing Locations of Monitors.
95

A selected group of constant air monitors and monitrons form the two
input channels, contamination and radiaticn, respectively, which actuate
the building-evacuation system. The output signals from the monitors in
each channel are transmitted to a central annunciator panel located in
the auxiliary control roomnm.

Excessive radiation signals from two monitors in either the radia-
tion or contamination channel will cause the evacuation siren to sound.

At the same time, four beacon lights at the corners of the building will
begin flashing and a signal will be sent to Guard Headquarters. Also, a
manual evacuation switch is available on the reactor console.

On the constant air mcnitors and the monitrons, there is an inter-
mediate alarm. When a monitor reaches this set point, the indicator lamp
on that monitor input module is lighted and a lccal buzzer sounded. This
buzzer is located in the HP annunciator panel. At the same time this
buzzer sounds, an annunciator in the control rocm indicates that a health
physics instrument 1s in the alarmed state. This alarm signal is also
produced by an instrument mslfunction, since both the constant air moni-
tors and monitrons produce a signal upon instrument failure.

An administrative prccedure involving both reactor operations and
health physics is used to ensure adequate protection during reactor main-
tenance periods without cauvsing false evacuation signals. This is imple-
mented by two key-locked switches, one in the main control room and the
second in the maintenance control room (see Fig. 2.39 in sec. 2.6). Oper-
ating either of these switches prevents the monitors from actuating the
alarm sirens and beacon lights and switches any evacuation signal to Guard
Headguarters. The established procedure requires the Chief of Operations
to call Guard Headquarters and to evacuate the area of all personnel except
those essential to the particular maintenance operation involved. The
Chief of Operations then throws the key-locked switches and disables the
alarm sirens. The manual evacuation switches can be used to override the
alarm cutout should evacuation be required. The individuval monitor alarms

and the main annunciator still continue to function.
96

2.3 Controcl Instrumentation

 

2.3.1 Nuclear Instrumentation

 

The normal operating ranges of the nuclear instrumentation are listed
above in Table 2.4, and the safety channels were described in Section 2.2.
The nuclear instrumentation used for reactor control consists of two wide-
range counting channels and two linear power channels, as described below,

In addition, provision is made for two temporary BF; counter channels
(Fig. 2.25), which will be used for the initial critical experiments and
other low-level nuclear tests. These counters will be located in vertical
holes near the inner periphery cof the annular thermal shield, as shown in
Fig. 2.26. Alsc, extra tubes are available in the neutron instrument pene-
tration, as described below.

All permanent neutron sensors and a portion ¢f the associlated electri-
cal and electronic ccmponents shown on Figs. 2.10, 2.27, and 2.28 are lo-
cated in the nuclear instrument penetration, a 36-in.-diam, water-filled,
steel tube. The upver end of the tube terminates on the main floor (852-ft
level), end the lower end penetrates the thermal shield at the horizontal
midplane cf the reactor core. The penetration is inclined at an angle of
42° from the horizontal; its design and location are depicted in Figs. 2.26
and 2.29.

Fach sensing chamber is located and supported within the penetraticn

by an individusl aluminum tube. The chamber housings are lcose fits in the
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Pig. 2.25. Typical Low-Level BF3 Counting Chamnel for Initial Criti-
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UNCLASSIFIED
ORNL—DOWG 64-984

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INSTRUMENT
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Fig. 2.28. Linear Power Channels and Automatic Rod Controller.

tubes so that chamber positioning can be accomplished by mcving the chamber
axially within the tube.

A float-type water-level indicator is used that will alarm on low
water level. Water supply to the tube is controlled manually by a hand
valve. An overflow pipe near the top of the penetration limlts the upper
water level.

The normal water level provides ample biological radiation shielding.
In addition to the level alarm, monitron No. 2 (see Fig. 2.23) will alarm

on an excessive radiaticn level in the region near the top end of the
 

100

241t 2in,

 
 
 
 

NEUTRON INSTRUMENT
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UNCLASSIFIED
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Nuclear Instfumentation_Penetration.

geneous Reactor Experlment (HRE-2).

Wide-Range Counting Channels.

A 51m11ar installation was used successfully in the Homo-

v

The entire operating range of the re-

actor is monitored by two counting channels, using automatic positioning

of the fission chambers to extend the limited range of the more usual ar-

:'rangement.4 The block diagram of one such channel is shown in Fig. 2.27.

The technique mekes use of the variation of neutron flux with detector

 
101

position, which is nearly exponential in most shielding configurations,
The function generator is used to make the detector-position signal more
nearly proportional to the logarithm of the neutron attenuation. Adding
this signal to a signal proportional to the logarithm of the counting rate

gives a resultant signal proportional to the logarithm of the reactor

power. Computing a suitable derivative then yields the reactor period.
The chamber is moved by a drive mechanism under the control of a small
servo system, whose function is to attempt to maintain a constant counting
rate of 10% counts/sec.

When the reactor is shut down, the counting rate is much less than
10% counts/sec, and the servo drives the detector to its innermost posi-
tion. As the reactor is started up, the counting rate increases and
causes the indicated reactor power to increase correspondingly. As the
startup proceeds to higher power, the counting rate reaches 104 counts/sec,
and the servo withdraws the detector, keeping the counting rate constant;
the change in detector position then increases the indicated reactor powver.

The servo need only be fast enough to follow normal maneuvers; any
transients which are toc fast for the servo will change the counting rate,
and the channel will read correctly in spite of the lagging servc. The
advantages of this technique are the absence of any necessity for range
gwitching or any other action by the operator, the wide range covered,
and the elimination of the 'dead time' induced when the chanber is with-
drawn quickly, as in other methods. Further, operation is always at an
optimum counting rate when sufficient flux is available.

Tinear Power Channels and Automatic Rod Controller. There are two
linear power channels in the MSRE. Each channel consists of a compengated
ionization chamber, power supply, and linear picoammeter, with remote
range switching. Either of the two channels can be used to provide the
flux input signal to the automatic rod controller, and both can be used
for monitoring reactor flux level over a range of about six decades. A
block diagram of the two flux channels and the automatic rod controller

is shown in Fig. 2.28.

 

“R. E. Wintenberg and J. L. Anderson, "A Ten Decade Reactor Instru-
mentation Channel,  Trans., Am. Nucl. Soc., 3(2): 454 (1960).

 
102

The automatic rod controller has two modes of operation. When the
reactor is in the "start" mode (see discussion of modes in the next sec-
tion), the automatic rod controller is a conventional on-off flux servo.
The set point, in terms of neutron flux, depends upon the range of the
picoammeter selected by the operator. Continuous adjustment is available.
The maximum set point attainable in this mode corresponds to a power of
approximately 1.5 Mw. At such low power levels the effect of power on
temperature 1s slight; however, the operator must adjust the thermal load
to keep the system temperatures within acceptable operating limits. When
the reactor is in the "run" mode, the automatic rod controller becomes &
reactor temperature controller, with the temperature set point adjustable
by the operator,

When the servo controller is operating as a temperature controller
itg principal function is to augment the natural temperature coefficient
of the reactor and to provide the reactivity changes required to compen-
sate the small [-107% (Sk/k)/Mw] power coefficient of reactivity. The
average reactor power will, over a period of time, tend to follow the
load demands with only small changes in average core outlet temperature,
The thermal capacities of the system are large and the power density is
low; this, plus the long transport lags in the system, result in long time
constants and delayed coupling between the power demand and the power being
generated in the core. Analog simulation hags shown that without supple-
mentary control, the reactor's response to changes in load are slow and
characterized by flux and temperature overshoots and oscillations. The
rod control servo, operating to hold core outlet temperature constant,
makes up for these deficiencies in inherent control at a rate which causes
a minimum of system perturbation.

The servo controller,5:6 see Fig. 2.30, compares measured reactor

flux, ¢, with the measured temperature rise, AT, in the core and computes

 

53. J. Ditto, "Preliminary Study of an Automatic Rod Controller Pro-
posed for the MSRE," internal ORNL document MSR-62-96.

®E. N. Fray, S. J. Dittc, and D. N. Fry, Design of MSRE Automatic
Rod Contreller, "Instrumentation and Controls Div. Ann. Progr. Rept.
Sept. 1, 1963," USAEC Report ORNL-3573, Oak Ridge National Laboratory.
103

UNCLASSIFIED
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SUMMING AMPLIFIER (TYFICAL),

 

 

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T = FUEL SALT TEMPERATURE, CORE INLET
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Fig. 2.30. Computer Diagram for Servo-Controller Simulation.
104

a flux demand signal which, in turn, is used to produce the appropriate
regulating rod motion so that reactor flux is adjusted to meet the load
demand at rates equal to the rates of demand changes. The controller
includes a slow reset element that corrects (slowly reduces) the error
signal on a long time scale. The rate of reset is such that it has little
effect on the transient error signal brought about by load changes. This
reset element acts so that slow drifts and calibration errors of the tem-
perature and flux sensors dc not affect the servo's ability to make the
reactor follow changes in load. If, however, the outlet temperature
thermocouple drifts, the reactor outlet temperature will follow the drift
and system temperatures will be raised or lowered accordingly. Such a
change will be apparent to the operator since system temperatures are
monitored frequently and in many places.

Figure 2.31 shows the result of an analog simulation of the response
ol the reactor to step changes in load from 10 to O Mw and back to 10 Mw.
The temperature of the salt leaving the reactor shows nc appreciable de-~
viation from set pcint, except after t = 250 sec, when the controller is
forced by its limiting action toc hcld the power at ~0.5 Mw with no heat
loss. The behavior of varicus system temperatures as the temperature set
point is linearly decreased and increased at 5°/min, while the load is

held ccnstant at 10 Mw, is shown in Fig. 2.32.

2.2.2 Plant Centrel

Because of the conflicting control reguirements under different oper-
ating conditions, the control system has been designed to operate in sev-
eral modes selected by the cperator. These are shown in the tlock diagram
in Fig. 2.33. The "prefill" mode is that mode of operation during which
the reactor is empty and certain manipulations of the helium and salt sys-
tems are required. This mode provides interlocks to prevent filling of
the reactor while allowing transfer of fuel or Tlush salt between storage
tanks., Circulation of helium for cleanup and for heating or cocling can
also be maintained. The "operate” mode is used for f£illing the reactor
and for all operations associated with a full reactor. In the "operate”

mode the transfer of fuel or flush salt between tanks is prohibited.
TEMPERATURE (°F)

PCWER (Mw)

1250

1225

1200

1475

1150

1125

1400

1075

1050

1025

7

 

Fig. 2.31.

UNCLASSIFLED
ORNL—DWG €64—-620

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Results of Analog Simulation of

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SALT TEMPERATURE (°F)

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Results of Analog Simulaticon of

Reactor Response to Ramped Changes in Outlet Tem-
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106

UNCLASSIFIED
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Fig. 2.33. BSimplified Master Plant Control Block Diagram.

Inasmuch as the reactor may be empty, yet conditions for the prefill mode
are not satisfied, or full and the conditions for the operate mode are not
satisfied, the third mode is defined as the mode when neither of the
other two exists and is called the "off" mode. When the off mode exists,
none of the usual operations involving pressurization, circulation of
helium or fuel, or manipulation of the control rods is permitted.

For nuclear operation of the reactor it is necessary that the reactor
be full, and the fuel should be circulating. Under these conditions the

reactor system may be operated in either the "start" mode or the "run"
107

mode. (These are, in a sense, submodes of the operate mode.) The start
mode is used for all power levels up to gbout 1 Mw, at which point thermal
effects become significant. The run mode is used for power levels between
1 Mw and the design point of 10 Mw.

Control of the nuclear operation of the MSRE includes control of
heat generation and of heat removal. To attain steady state, it is clear
that the two quantities must be made equal. For low-level operation the
reactor neutron flux is consldered to be the most significant reactor
parameter, although system temperatures must be held within safe limits.
ITnasmuch as at power levels lesgs than about 1 Mw the maximum rate of
change of system temperature with no heat loss is of the order of a few
degrees per minute, the use of flux control at these power levels does
not pose much of a problem in adjusting the cooling capacity of the radia-
tor to control temperature. Therefore, when the reactor system is in the
start mode, where the maximum power allowed is 1.5 Mw, the flux and ther-
mal load are independently controlled. The use of an automatic flux con-
troller aids in such independent control by effectively decoupling the
flux from any temperature feedback brought about by load changes or by
changes in electrical heat input from loop heaters.

In the power range above 1 Mw, the problems of independent ccontrol
of power and of temperature become more difficult. The control system
for high-power operation of the MSRE, with its capability demonstrated
by analog analysis, permits the operator tc adjust the heat removal rate
by manipulation of the radiator compconents while an automatic rod con-
troller is used to maintain the desired reactor salt outlet temperature.
The automatic rod controller has a temperature setpoint that is adjustable
by the operator at a maximum rate of a few degrees lahrenheit per minute.
In addition, the control of the load has been programmed so that the opera-
tor need only actuate a single switch to increase or decrease the load
over the entire range at a rate compatible with the temperature control-
ler's capability. In practice, the automatic rod controller changes its
mode of operation from flux to temperature when the reactor operating mode

is changed from start to run.
108

The capsability of the automatic rod controller to make the reactor
closely follow its load at all power levels from 1 to 10 Mw, with inde-
pendent control of reactor outlet temperature and load, will be a great
asset to the reactor operator. However, manual modes of operation have
been provided for greater flexibility in the experimental progran.

Rod Control. A simplified block diagram of the rod control system
is shown in Fig. 2.34. In general, manual withdrawal of the three shim
rods is permitted if the reactor period is not shorter than 10 sec and
if no automatic rod insertion ('"reverse") signal exists. Reverse and the
conditions producing it are discussed in a subsequent paragraph.

In addition to manual operation of the rods, there is an automatic
rod control, or "servo, ' mode of operation. In the servo mode, one of the
shim rod drive motors is controlled by signals from the automatic rod con-
troller described in the preceding section. In such operatiocn, the opera-
tor exercises control cver the position of the servo-controlled rod (called
the shim regulating rod) by adjustment of the other rods.

The regulating rod limit assembly (Fig. 2.35) is an electromechanical
device used to limit the stroke of the servo-controlled rod. Figure 2.36
is a diagram of the associated circulitry. This diagram is a simplified
version of the actual control circuits in that it omits administrative and
supervisory contacts, upper and lower rod limit contacts, etec.

Referring to Fig. 2.35, 1t can be seen that, by means of the mechani-
cal differential, the net rotation of the limit switch cam is the resultant
of two components: (l) the rotation produced by the balance motor and (2)
the rotation of the shim locating motor. When the reactor is in the servo
control mode and if the operator is not relocating the position of the
regulating rod span, the balance motor, driving through the clutch (the
brake is disengaged) and the differential, rotates the limit switch cam
through an angle that is directly proportional to the linear motion of the
servo-controlled rod. It is assumed for the sake of discussion that the
reactor 1s being operated in steady state and that no fuel is being added.
Burnup and poisoning will cause slow withdrawal of the servo controlled

rod unlessg the reactor is manually shimmed by the other two rods. If no
o 1-»-1

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

     
   
 
    
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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i POSITION OF A MANUALLY OPERATED SWITGH 1 ‘
S \ |
ROD 1 NOT AT - ROD § NOT AT ROD 2 NOT AT ROD 3 NOT AT RESULTING CONDITION OR ACTION
SHIM WITHDRAW - REG. ROD WITHDRAW WITHDRAW
LT INSERT LIMIT LIMIT -LIMIT !

 

 

 

 

 

 

 

   
    

WITHDRAW
REG. ROD

 

 

 

WITHDRAW
ROD NO.2

 

Fig. 2.34.

Simpli]

 

 

 

      

WITHDRAW
ROD NO.3

 

e e s

e T ————

arme

 

NOTE:

* THESE CONDITIONS REFER TO POSITION OQF THE

 

 

SERVO CONTROLLED ROD WITHIN TS AUTOMATIC
SPAN AS DERIVED FROM-THE REGULATING ROD
LIMIT SWITCH ASSEMBLY.

fied Rod Control Block Diagram.

 

 

 

 

 
ROD DRIVE MOTOR,
SERVO OR MANUALLY
CONTROLLED

 

 

 

 

 

 

 

 

JA

. y
LOWER
LIMIT SW ITCH—-”"’g;

[
" CONTROL ROD |

 

 

 

 

 

 

 

 

 

 

SHIM LOCATING
MOTOR, OPERATOR
CONTROLLED

GEARBOX

NOTE: AS REDUCED TO PRACTICE THE
CONTROL ROD MOTION IS
REPRODUCED IN THE CONTROL ROOM
AND LIMIT SWITCHES ARE ACTUATED
B8Y THE REGULATING ROD umMIT
SWITCH ASSEMBLY. SEE DIAGRAM.

——————

REGULATING ROD
SPAN Ay,

  

REGULATING ROD
LIMIT SWITCHES

3% AVAILABLE TO SERVO
CONTROLLER WITH
REG. ROD SWITCHES

L ATy,

 

1

 

 

SYNCHRO
POSITION

    
  

  

UNCLASSIFIED
ORNL-DOWG 64-622

REGULATING ROD LIMIT SWITCH ASSEMBLY :
"IN INSTRUMENT CABINET

 

 

 

 

SHIM LOCATING MOTOR

- ROD POSITION OPERATOR CONTROLLED
TRANSMITTER BY CONTROL ROD ACTUATOR

SWITCH. USED DURING

   
  

 

    
   
  
 
   
  
  
 

 

 

 

 

 

= SERVO OPERATION ONLY
COARSE r- |
: csms.ox ! I I | .
| Hl—
- g SYNCHRO _
ROD DRIVE MOTOR: | | CONTROL  'BALANCE e L
. OPERATED BY ROD CONTROLLER | | | TRANSFORMER MOTOR
ONLY WHEN IN SERVO OPERATION
AND BY ROD ACTUATOR SWITCH | | | cLutch-BRAKE: CALIBRATION
IN MANUAL OPERATION - o. IN SERVO OPERATION THE DIAL
. CLUTCH 15 ENGAGED AND ~_. .
; V THE CAM 1S COUPLED TO SLIP CLUTCH
R I UNT I b} tie eaLance woror ]
- 1 i b. IN MANUAL OPERATION GEAR BOX FOR -
THE OUTPUT SHAFT IS
) | | LOCKED AND THE SPAN CHANGE GEARS
| | BALANCE MOTOR
| DECOUPLED (?D'
I REGULATING ROD
| LIMIT SWITCH
| | | cam anp umir switcres LOWER
| | WITH MECHANICAL STOP ,
| | REGULATING ROD UIMIT REGULATING: ROD
WITCH UPP
| | SWITCH UPPER POSITION TRANSMITTER
I CLOCK CAM ROTATION (SYNCHRO)
I | equiv. TO:
| | 0. WITHDRAW ROD NO.{ |
I b, INSERT LIMIT SWITCH
/ b SPAN { DECREASE y,) |
- ' |
| ]
@ @ _ SEE FIG.2.36 FOR
COARSE - FINE CIRCUIT DIAGRAM
. REG ROD POSITION
SHIM ROD INDICATOR {RELATIVE
POSITION TO SPAN LIMIT
INDICATORS -

 

\  SWITCHES)

 

 

 

. 34
. . —‘

CONTROL CONSOLE

' Fig. 2.35. Regulating Rod Limit Switch Assembly.

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.« ) : . . ) ORNL~DWG 64—-6082
’ HSv AC '
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4 _1_SERVO CONTROLLER _
: rr—— " .
_lexiea : ' __I_ $22D —L_stop ~ KIBEF _I_ INSERT" ROD NO.4 oM REG. ROD
¢ : %= OPENS ON AUTOMATIC "REVERSE® MANUAL ] SHIMNO.{  "REVERSE" . SSEMBLY LIMI - :
. i . " " - UPPER SWITCHES LLOWER K17T5A
: : ; ‘ , REVERSE ACTUATE . LMIT L “NaL Tmir _1_K \
_ . - _ "INSERT" S k1838 L. Ki76A 7 . < L. Ki82a T 'WITHDRAW
. S . _ ‘ T _lzkesiE INSERT [ WITHORAW ~ | ——J ' T~ "INSERT“ LimiT | LIMIT SWITCHES
& : _ . . ! > OPENS WHEN SHIM REG. ROD NO. 1 ROD NO.{ . SWITCHES
; , o : ROD LIMIT SWITCH ) ;é’msac
— ASSEMBLY CAM AT K183D ;é '
_LzKatToE : I KAI7OF LOWER LIMIT St 1
71~ CLOSED WITH OPEN WITH 1% Iv
SERVO “ON" SERVO "ON" , -
. | : ' _L_kazao0c . _ _lzkB170
| .- ‘ o T CLOSED WHEN SHIM-REG. | 71~ CLOSED WITH I' _ |
sion | . - _ ROD ASSEMBLY CAM AT | “SERVO" ON ! 2%(240 | K244 o
_ UPPER LIMIT o ) o
' : , . SHIMNO.{ ' ' 1 SERVO CONTROLLER | <% i
"“ _l_ Ki38aA _L_KiTaA ACTUATE IR ~T~ WITHDRAW.ROD NO.1 . : | ,
f;?g‘igp,?.t‘“-' ‘T GROUP WITHDRAW . wITHDRAW : ' : - _ INSERT | WITHDRAW | WITHDRAW
S20B] SHIMS NO. 2 _ [/ K240F _lzK240a S | : I ¥ _
- AND NO.3 7T~ OPENS WHEN SHIM 7~ OPENS WHEN SHIM o - _ DRIVE MOTOR _ _ | SHIM LOCATING MOTOR
| : $218 | ACTUATE REG. ROD LIMIT SWITCH REG. ROD LIMIT ROD NO.4 | ~ | SEE FIG. 2.35
5224 ‘ ‘ CAM AT UPPER LIMIT SWITCH CAM AT i Lol Posimion || J _
fiz GROUP ROD  ° | L _L -~ o | uppEr LMIT } TRANSMITTER [[ ™
1 KANT? . '
WITHORAW |/ ky70H : L OPéuovcmH _Lz karoo | i S MECHANICAL DIFFERENTIAL
(MANUAL} . >A— CLOSED WITH | servo “on" . 7T CLOSED wiTH é
SERVO "ON" _ _ SERVO "ON" :
: ‘ : 0 NOTES
! i - . ' :
_ , : . SHIM LOCATING
e ,‘; [t ome] ~ [ hop o s womon] | woron rorarion
71~ OPENS WHEN SHIM REG. - £ ki83c ' : ‘ {INSERT)
o - AT LOWER LIMIT : ‘ ’ o AND VICE VERSA FOR COUNTERCLOCKWISE CAM ROTATION
| o NO.1 INSERT O |
KI7?5 ' ‘ KiT6 K182 ‘ K183 | § ' ' 2. REFER TO FIG. 2.35 FOR DETAILED DIAGRAM OF CAM DRIVE
. THIS RELAY, WHEN ENERGIZED, THIS RELAY, WHEN o THIS RELAY, WHEN ENERGIZED, " THIS RELAY, WHEN ENERGIZED, , ‘ o
' _ K174 OPERATES SHIM LOCATING ENERGIZED, OPERATES OPERATES SHIM LOCATING OPERATES ROD ORIVE MOTOR : 3. SWITCH AND RELAY CONTACT ASPECT (GPEN OR CLOSED) ON - /
- MOTOR N REG. ROD LIMIT ROD DRIVE MOTOR IN MOTOR IN LIMIT SWITCH INSERT IN “INSERT" DIRECTION CAM - THIS SKETCK IS FOR FOLLDWING CONDITIONS :
SWITCH. WITHDRAW DIRECTION ROD WITHDRAW DIRECTION, DIRECTION, CAM ROTATES ROTATES WITH SERVO "ON m . 1. SERVO "ON", ZERO EXROR SIGNAL
CAM ROTATES =~ CAM ROTATES WITH : ' fl 3 2. REACTOR SYSTEM IN|"OPERATE~RUN" MODES
) o SERVO “ON O : _ 3 ‘ 3. REGULATING ROD BETWEEN LIMITS
GROUP ROD REG. ROD LIMIT. ~ ROD NO.1 REG. ROD ROD NO. 1 { ' ,
WITHDRAW SWITCH ASSEMBLY WITHDRAW LIMIT SWITCH INSERT : _
- WITHDRAW - . INSERT ‘ ' ‘

Fig. 2.36. Regulating Rod Control Circuit.

e g i T L eatR,

 

 

 
112

shimming taekes place, the limit switch cam'wiil continue to rotate clock- '(]' -
wise, and unless‘the shim locating motor is actuated to produce counter- |
clockwise cam rotation the upper shim regulating/rod limit switch»wili be
opened, its associated relay No. K240 will drop out, and contact No. K24O0F
(see Fig. 2.36) will open. This wili prevent the servo system withdraw
relay from transmitting power to the rod withdraw relay (No. K176). Note
that the servo system is not turned off and that it continues to exert _
control in the "rod insert"‘direction; i.e., if for any reason the servo
calls for the insertion of negative reactivity with the Servo rod at the
upper lifiit of the regulating rod span, the rod insert requirement wiii

be transmitted to relay K183, which causes the rod drive motor to ifisert
the rod. This rod No. 1 insert relay, K183, cross interlocks the No, 1

rod withdraw relay, K176, such that an "insert" request from any source
véverrides all "withdraw'" requests. The operator must now shim the reactor.
If he chooses to shim by using rod No. 1, the servo controlled‘rod, he

does so by means of the individual rod control switch (819) used for manual
| bperation. This closes contact S19A and energizes K175,.the relay which
causes the shim locating motor to turn the cam counterclockwisé. The limit
switch closes, contact K240F closes, and the servo is in command of rod

No. 1. The servo now withdraws rod No. 1 until the upper limit is again
reached or until shimming requirements are met. OSince the éhim relocating
motor moves the cam at a slightly lower speed than does the balance motor;
the actual withdrawal of rod No. 1, in these circumstances, is a series

of start and stop movements until the span of the regulating rod is re-
located. The alternate method of shimming is to withdraw either or both

| manually controlled rods Nos. 2 and 3 until the servo automatically returns
rod No. 1 to a suitable place within the regulating rod span. '

Should the servo controller malfunction and ask for a continuous out-
of-control withdrawal of rod No. 1, the withdrawal will only persist until
the upper regulating rod limit is reached. The on-off power amplifier re-
~lays in the servo controller do not exert direct control of the rod drive
motors (see Fig. 2.36); instead they control the "withdraw" and "insert"
relays (Nos.,Kl76 and K183), and & servo malfunction has to Be accompanied
by faiiure of other control-grade interlocks to become 6f consequence. ' (’
113

Should the regulating rod limit switch mechanism misoperate in con-
Junction with a servo controller malfunction such that the cam fails to
rotate as the rod 1s being withdrawn by the servo, the control system
interlocks call for a "reverse” (group insertion of all three rods) for
any of the following conditions:

1. Reactor power above 1.5 Mw in start mode.

2. Reactor power above 12 Mw,

3. Reactor period less than 10 sec in start mode.

4. Reactor outlet temperature above 1275°F.

5. Excess fuel level in the overflow tank.

"Reverse' by the control system is independent of the condition of the
servo controller and is also invoked should the operator inadvertently
withdraw either shim rod so as to produce any of the above listed condi-
tions. '"Reverse," while not a safety system action, is used to keep re-
actor system parameters within well-defined operating limits and away from
other limits, such as scram level power, which are classified as safety
limits.

If one of the shim regulating rod limit switches remains actuated or
if the cam fails to move off the mechanical stop, rod No. 1 is inhibited
from further motion in one direction only. The operator can provide con-
trol with rods 2 and 3 and can switch over to manual control until the
trouble is cleared.

The span, Ayy on Hig. 2.35, of the regulating rod limit switch assem-
bly and, hence, the reactivity available to the servo controller 1s ad-
Justed by means of change gears in the cam drive train. The mechanism 1s
not directly avalilable to operating personnel, and all changes in span are
subject to administrative control,

Load Control. Control of the heat removal rate of the MSRE is
achieved by controlling the effective area of an air-cooled radiator and
the mass flow rate of alr through the radiator. Control of the effective
area 1s accomplished by adjusting the positions of a pair of docors, one
on either side of the radiator. These doors may be moved individually or
together by a single motor through a clutch and brake arrangement. As

indicated in Section 2.2, the doors are closed automatically to prevent
114

freezing of the coolant salt. Control of the mass flow rate is accom-
plished by the use of one or two main blowers and by adjusting the posi-
tion of a2 bypass damper.

In principle it i1s possible to control the heat rejection rate manu-
ally; however, the very long time constants in the system and the complex
interactions between the varicus control devices have led to the develop-
ment of a programmed sequence for increasing or decreasing load. This
sequence consists of a series of steps involving movement of radiator
doors, switching blowers on and off, and changing a differential pressure
contrcller set point to regulate the position of a bypass damper. The
operator need only operate a single load demand switch to increase or
decrease load over the entire range and may interrupt the sequence at any
load between 1 and 10 Mw. During these lcad changes the automatic rod
controller holds the reactor outlet temperature at its set point. The
reactor power rollows the load with a time lag resulting primarily from
the thermal capacities and transport lags of the salt systems. TFor added
flexibility in experiments, manual control of the various load control
devices has also been included, with sultable provisions to limit tran-
sients during transfer from manual to programmed operation.

Interlocks and Circuit Jumpers. Throughout the plant there are many

 

control devices designed to assist the operator to attain orderly opera-
tion. These provide interlocks that prevent certain undesirable operations
under specific conditions but are not necessary for safety. DBecause the
experimental nature of the reactor demands considerable flexibility in the
control system, an arrangement has been provided for bypassing these con-
trol interlocks. A Jjumper board located on the main control board is used
to facilitate alteration of interlock circuits and to enhance administra-
tive control of such changes.

It is also desirable, when the reactor is not operating, to run opera-
tional check tests of the vital components that are subject to safety sys-
tem control. A typical example is a prestartup test to verifly that the
radiator doors operate correctly. The safety system maintains the doors
closed for either the condition of low salt flow or low salt temperature

(see Fig. 2.9) and would prevent such a test in an empty, nonoperating
115

gsystem. 1In order to permit such essential tests, the safety system design
permits bypassing the necessary safety system contacts by using the jumper
board. Jumpering of safety system circuitry is subject to stringent ad-
ministrative control; i1.e., permission of the Chief of Reactor Operations
must be obtained before using a jumper. The safety system Jjumpers and
associated circuitry, Fig. 2.37, fulfill design criteria, as follows:

1. Safety circuit isolation and separation are maintained., This is
accomplished by the "Bypassing Relay' on Fig. 2.37.

2. The Jjumpers (plugs) ére readily visible to supervisory and opersa-
tions personnel in the main contrel area.

3. The circuit and its condition (whether or not energized) in each
string of contacts is displayed to personnel in the main control area by
means of Iindicating lamps.

4, The presence of a safety system jumper is annunciated.

5. If any safety system contact is bypassed by a jumper the control
gystem cannot be put in the "operate” mode (see Fig. 2.33).

6. Failures of components in the Jumper board circuitry will not
Jeopardize operation of the safety circuits.

Control interlock (nonsafety system) Jumpering meets the criteria of
items 2, 3, and 6 above.

Figure 2.37 is a much simplified version of a typical safety circult
in that 1t shows only one relay contact in the safety relay circuit. In
an actual circuit there are several contacts, all of which may be wired
for Jjumpers.

In such a string of contacts with indicating lamps to show contact
condition, the remote possibility exists that the lamps could bypass enough
current around an open contact to keep the safety relsay operated. This
situation requires that the lamp neutral be open so that the current path
through the lamps passes to neutral via the safety relay. All safety con-
tact-indicating lamp circuits contain a dropping resistor and a silicon
diode in series with the lamp so that normal current through one lamp is
much less than required to maintain the safety relay operated. 1In the
event of an open lamp neutral, the diodes are "back-to-back"” in the sneak

circuit through the safety relay. A typical silicon diode will pass only
116

UNCLASSIFIED
ORNL-DWG €4-6080

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

JUMPER BOARD SAFETY SYSTEM
ON MAIN CONTROL BOARD RELAY CABINET
LS /99@754464//» ;7?25609?99006664447779066644/790</
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NOTE: THE JUMPERS AND THE INDICATING LAMPS ARE ON THE MAIN
CONTROL PANEL AND VISIBLE FROM THE OPERATOR'S CONSOLE.

Fig, 2.37. Diagram of Safety System Bypassing with Jumper Board.

a fraction of a milliampere of reverse current. Since relay holding cur-
rents are over 100 mamp, the relay will not be prevented from dropping out.
The Jjumper board presents a graphic display in full view of the opera-
tor of all of the circuitry subject to bypassing. The indicating lamps are
located on the board between each contact subject to Jumpering and provide

immediate information as to circuit condition (whether energized or not).
117

A circuit jumper, requiring administrative formality for its insertion, can

not become a forgotten clip lead.

2.4 Neutron Source Considerations

MSRE fuel salts provide a substantial inherent (a,n) source.! When
the core vessel is filled with fuel salt containing sufficient uranium for
criticality, this intrinsic source produces more than 10° neutrons/sec.

The resulting fission rate is such that statistical fluctuatlions are neg-
ligible and the reactor's behavior is accurately described by the well-
known kinetic equations. Kinetic calculations have established that this
initial fission rate is high enough to make tolersble the worst credible
startup accident. This accident is caused by the uncontrolled, simultane-
ous withdrawal of all three shim-safety rods and is described in detail in
Section 7.1.2.

It can be concluded that the inherent (a,n) source 1s adequate for
safe reactor operation. Since this source 1s intrinsic with the fuel salt,
instrumentation is not required to verify its existence.

Subecritical testing will recuire an external [other than (a,n) in the
fuel salt] neutron source and highly sensitive detectors. This external
neutron source is also used during normal, routine operation for monitor-
ing the fill and startup procedures and for partial shutdowns and, thereby,
provides information required for consistent, orderly operational routines.
The control system is interlocked so that a count rate of at least 2
counts/sec is required during fill and startup. During partial shutdowns
the fuel salt i1s maintained at the reactor operating temperature in the
reactor primary loop, and the shim-safety rods are inserted. The external
source is used to monitor the reactivity deficit.

The source is not located inside the core vessel because (1) typical
MSRE temperatures exceed the melting point of antimony and its oxides and
the required cooling system would be unduly complex, (2) accessibility and,

hence, source removal would be restricted, and (3) an additional in-core

 

7P, N. Haubenreich, "Inherent Sources in Clean MSRE Fuel Salt," USAEC
Report ORNL-TM-611, O=2k Ridge National Leboratory, Aug. 27, 1963.
118

penetration is undesirable. The locations of the source and the sensitive
BF3 counters are shown on Fig. 2.26, An antimony-beryllium source pro-
viding 107 neutrons/sec is adequate for all purposes.8

The foregoing evaluation of source requirements takes no credit for
the large photoneutron source that will be present after the reactor has

been operating at power for a relatively short time.

2.5 FElectrical Power System

 

A simplified version of the power distribution system for the MSRE
is shown in Fig. 1.17 of Section 1. Normally all electrical power is sup-
plied by TVA. Two main 13.8-kv supplies are available. These main supply
lines originate at geographically separated locations. Switchover from
the "normal" to the "alternate" 13.8-kv bus is done automatically.

Standby power is provided by three diesel-engine-powered generators,
and the switching operations required to put the diesels on the line are
manual, as 1s diesel startup. Uninterrupted ac and dc power for the vital
instruments, the control system, and emergency lighting and the power to
operate the larger circuit breakers is supplied by a motor-generator-
battery system that operates continuously. Normally the 250-v battery
flcats on the output of the motor-generator and is kept charged. Direct-
current requirements are supplied from the battery, and a 25-kw motor-
generator set powered by the battery provides ac power. This 25-kw motor-
generator set is normally connected to its load and is run continucusly sco
that no switchover is required in the event of a TVA outage. This elimi-
nates any momentary interruption that would be produced by switching time
and the time required for the 25-kw motor-generator set to pick up its
load. ©Such an interruption would, among other things, scram the control
rods unnecessarily.

The power distribution system was not designed for the purpose of
insuring uninterrupted, full-power operation of the MSRE. In the event
of a TVA outage, the reactor power level will be reduced until the standby

diesels are started and are producing sufficient power to run all pumps

 

8J. R. Engel, P. N. Haubenreich, and B. E. Prince, "MSRE Neutron
Source Requirements" (report to be igsued).
119

and blowers. The reliable storage battery supply described in the preced-
ing paragraph has sufficient capacity to operate for spproximately 2 hr.
This is ample time to start the diesels. If, after a reasonable pericd of
time, it 1s not possible to restore sufficient electrical power, either
from diesels or TVA, to operate the reactor system in an orderly fashion,

the reactor may be drained.

2.6 Control Room and Plant Instrumentation Layout

2.6,1 Main Control Ares

The main control area, shown in Figs. 2.38 and 2.39, contains 12 con-
trol panels and the control console. The 12 panels comprise the main con-
trol board, which is in the form of an arc centered on the console.

The main control board, Fig. 2.40, provides a full graphic display
of the reactor system and its associated instrumentation. These panels
contain the instrumentation for the fuel salt system, the cocolant salt
system, the fuel and coolant pump oil systems, the off-gas system, the
Jumper board, and a pushbutton station for the various pumps and blowers.
The primary recorders for the nuclear control system are alsc on the main
board. The instrumentation on the console consists of position indica-
tors, control switches and limit indicators for the control rods, radiator
doors and Tission chambers, and other switches required for normal and
emergency operation of the reactor. Routine control of the reactor is
accomplished from the main control area. Control of some of the auxiliary
systems is also accomplished here, with remaining portions controlled

from the local panels shown in Figs. 2.38 and 2.41.

2.6.2 Auxiliary Control Area

 

The auxiliary control area, Fig. 2.39, consists of eight auxiliary
control boards, five nuclear control boards, one relay cabinet, one thermo-
couple cabinet, four diesel and switching panels, and a safety relay cabi-
net. The eight auxiliary panels contain the signal amplifiers, power com-
puter, switches and thermocouple alarm switches, auxiliary indicators not

reguired for immediate reactor operation, the fuel and cooclant pump
 

120

ORNL~DWG 64-808

 

 

 

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: ~ e __ BUILDING 7509
1 \ b (orrice)
HEALTH PHYSICS - :
- ROOM DOWN INS“{ggMGEENT
STORA
Tl L) Al 4
N | | /
: OFFICE OFFICE OFFICE — . .
' DOWN ) INSTRUMENT
SHOP -
} MAIN :
"F CHANGE. :
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Ll - :
DATA ROOM CHANGE I | PENETRATION -
SEE FIG.2.43 ay  ROOM L __ROOM
CONTROL ROOM %OT CHANGE ROOM
REACTOR -
CELL ENRICHER
PANELS
\
‘
DRAIN TANK CELL
i ) VENT
fl HOUSE
ROOF
v REMOTE
MAINTENANCE 1
e CONTROL ROOM _
ELEVATION 862ft Qin.
N
Fig. 2.38. Main Floor Layout of Building 7503 at 852-ft Eleva-
tion. . "
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- UNCLASSIFIED
7 ORNL—DWG 64-630 .
N
OBSERVATION @ -— fi‘) '
cnu.cmr\ . :
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THERMOCOUPLE
CABINET
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POWER
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= . PANELS
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AUXILIARY PANEI.S _ _
DATA AND DIESEL AUXILIARY CONTROL AREA NUCLEAR
SUPERVISOR PANELS /
ROOM 4"%‘1 ! '\ s‘/-'..
oaan |5I4I3I=I'I |
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Fig. 2.39.

 

Main and Auxiliary Control Areas.

 

 
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Fig. 2.40. Main Control

Board.

 

 

UNCLASSIFIED
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122

UNCLASSIFIED
ORNL-DWG 63-8399

up

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DECONTAMINATION ” FUEL m
CELL PROCESSING Cl WATER
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SWITCH P
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BLOWER
BUILDING C%QTTERR/ HOUSE
VENTILATION PANEL
 SYSTEM
FILTER HOUSE COVER GAS
DIESEL TROL PANE bo
DIESEL Jj_con OL PANELS Wn

BUILDING 7503~ELEVATION B840 ft Oin. LEVEL

Fig. 2.41. Layout of Building 7503 at‘840¥ft Elevation.

microphone amplifier and noise level indicator, and the substation alarm
monitors. The relay cabinet contains oniy the relays for the main control
circuitry. The thermocouple cabinet contains a patch panel with pyrometer
jacks. All thermocouples in the system except those on the radiator tubes
are terminated in this cabinet. |

The five nuclear panels contain some of the instrumentation for the
process radiation system, the oil system, and the sampling and enriching
system. The health physics mbnitofing alarm system, the high level gammé-
chamber electrometers and switching panel, and the reactor control nuclear
amplifiers, servo amplifiers, and other equipment for the nuclear control
system are also located here. Four panels contain the instrumentation for

the operation of the'emergency diesel generators. The distribution panels

w

 

 
123

for the control circuits are located on the south wall of the auxiliary

contrcl room.

2.6.3 Transmitter Room

 

The transmitter room, Fig. 2.42, consists of nine control panels, one
solenoid power supply rack, and a transmitter rack. Panels 1 and 2 are
for the leak detectors system, and panels 3 and 4 are for fuel drain tanks
1 and 2, coolant drain tank, fuel flush tank, and the fuel storage tank
welgh instrumentation. Panels 5 and 6 are the fuel and coolant pump bub-
bler-type level control panels, panel 7 is the freeze valve air control
panel, panel 8 is an installed spare panel, and pancl 9 is for the sump
bubbler system. The solenoid power supply rack contains most of the sole-
noids for the fuel salt system and the power supplies and distribution

panels for the electric transmitters and differential transmitters. The

UNCLASSIFIED
ORNL-DWG 64-629

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Fig. 2.42. Transmitter Room.
124

transmitter rack contains the remote amplifiers for the electric signal
transmission system and also the current-to-air transducers.

A1l electrical and pneumatic lines that connect transmitter room
equipment to input signals from the reactor, drain tank, water room, and
all other areas outside the building enter through sleeves in the floor.
Output signals from the transmitter room equipment are conducted via
feeder cable trays to the main cable tray system, which runs north-south

in the building on the 840-ft level.

2.6.4 Field Panels

Additional panels are required for the auxiliary systems. These (see
Figs. 2.38 and 2.41) are located near the subsystems which they serve.
Four panels for the cooling oil systems are located in the service room
and service tunnel; one coocling and treated water system panel is located
in the fan house; three panels are located on elevation 852 ft at column
line C-7 for the sampling and enriching system. Two cover gas system pan-
els are located in the diesel house. Two containment air system panels are
provided. One is located at the filter pit, and the other is located in

the high bay area.

2.6.5 Interconnections

 

The control areas and the controlled systems are interconnected by
cable trays and conduit. This system includes the main cable tray system
running north-south in the building, with branch trays and conduits Jjoin-
ing the main tray. These main trays carry all tubing for the pneumatic
system, thermocouple leadwire, and the ac and dc control wiring. Each
category of signal runs in a different tray to avoid mixing of the signals.
Risers connect the panels and cabinets in the control area on elevation
852 ft with the main trays. No exposed trays are in the main or auxiliary
control rooms. Safety system wiring is completely enclosed in conduit
and junction boxes and is separated from control and instrumentation

wiring.
125

2.6,6 Data Room

The data room (Figs. 2.38 and 2.43) is designed primarily to house
the data-handling system, but it also serves as the information process-
ing center for the reactor. The room contains storage space for all types
of instrument data and log sheets. It also containsg the data-plotting
equipment used with the data-handling system and other special data dis-
play and processing equipment.

This room 1s located in the north end of the reactor building next
to the main controcl room. Therefiare entries from the hallway and the
main control room. A large glass window is located so that the main
control panel can be seen from the room.

The data-handling system is installed along the west and north walls
of the room, Fig. 2.43. The data system console is located so that the
reactor control panel can be seen by an operator at the consgole. At least
two desks and several tables are located in the room for use of reactor
operations and analysis perso?nel. Data-plotting equipment and record
storage cabinets are placed at cconvenlent locations. A typieal process
computer system is shown in Fig. 2,44 .

The input signal cables for the data system are brought up to the
bottom of the room in cable trays and conduits. Holes are cut in the
floor beneath the signal input cabinet for routing the cables to the
cabinet. Cable trays and conduit are also provided beneath the floor
for signal lines to and from the main and auxiliary control areas.

The data room is air conditioned by the system which is used for

the control room and offices.
 

©

126

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ORNL-DWG 64-628

 

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Fig. 2.43.

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Layout of Data Room.

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UNCLASSIFIED
ORNL-DWG 64-631

   
 
 
  
   
 
 
 
 
 
   
 
  

PROGRAMMING AND MAINTENANCE STATION

FLEXOWRITER
HIGH SPEED PAPER TAPE PUNCH
HIGH SPEED PAPER TAPE READER

 

 

 

 

BASIC COMPUTER 5YSTEM

CENTRAL PROCESSOR
CORE MEMORY

DRUM MEMORY
INTERRUPT SUBSYSTEM
MASTER INPUT-OUTPUT CONTROL SUBSYSTEM
PROGRAMMING AND MAINTENANCE PANEL

 

 

 

 

 

 

 

INPUT-OUTPUT SYSTEM

ANALOG INPUT SUBSYSTEM

ANALOG OUTPUT SUBSYSTEM

CONTACT CLOSURE INPUT SUBSYSTEM
CONTACT CLOSURE OUTPUT SUBSYSTEM
COUNTER SUBSYSTEM

OPERATOR-PROCESS COMMUNICATION STATION

LOGGING TYPEWRITER
OPERATOR’S CONSOLE
CONTROL PANEL
DISPLAY
INPUT SWITCHES
FUNCTION SELECTOR
FUNCTION MATRIX
ALARM PRINTER

Fig. 2.44. Typical Process Computer System (TRW-340).

LZT
128

3. PLANT LAYOUT

3.1 Equipment Arrangement

The general arrangement of Building 7503 is shown in Figs. 3.1 and
3.2. The main entrance is at the north end. Reactor equipment and major
auxiliary facilities occupy the west half of the building in the high-bay
area. The east half of the building contains the control room, offices,
change rooms, instrument and general maintenance shops, and storage areas.
Additional offices are provided in a separate building that is east of
the main building.

Equipment for ventilating the operating and experimental areas is
located south of the main building. A small cooling tower and buildings
containing supplies and the diesel-electric emergency power equipment are
near the west side of the maln buillding.

The reactor primary system and the drain tank system are installed
in shielded, pressure-tight reactor and drain tank cells, which occupy
most of the south half of the high-bay area. The reactor cell is 24 ft
in diameter and 33 ft in height. It is surrounded by a 30-ft-diam steel
tank, and the 3-ft annular space is filled with a shielding mixture of
sand and water. The top of the cell is covered with removable shielding
blocks that extend 3 1/2 ft above the floor elevation of 852 ft and are
sealed with a welded membrane.

Ad joining the reactor cell on the north side is the drain tank cell,
which extends 5 ft deeper than the reactor cell to provide gravity flow
from the reactor to the drain tanks. This cell is rectangular (17 1/2 X
21 1/2 f£t), with 3-ft-thick heavily reinforced concrete walls lined with
stainless steel. Access to the equipment is through the removable roof
blocks, which are arranged in two layers with a sealing membrane between.

Several other shielded cells are located in the north end of the
high-bay area, as shown in Fig. 3.1. They provide shielding and venti-
lated isolation for the fuel processing cell, the waste cell, and two
cells for equipment storage and decontamination.

Fig. 1.3 (see sec. 1.2) shows how the components of the fuel circu-

lating system are arranged in the reactor cell. The reactor tank and the
 

 

 

 

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UNCLASSIFIED
ORNL DWG. 63-4347

   
   
 
    

   
 
  

OFFICE

  

FUEL PUMP -~
BATTERY LUBE
ROOM

    

MAINTENANCE
SHOP :

  

 

CHEMICAL

SERVICE
LABORATORY

TUNNEL COQOLANT PUMP

   

  
 

 

  
  
    
  
   
    

   
    
 
  
  
     
   
 
   

  
 

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‘ FUEL DRAIN THERMAL T BLOWERS
TANK NO. t- SHIELD
- [ REAGTOR
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C.. . ANNULUS

*ELEC. SERVICE AREA BELOW .
RADIATOR
. BLOWERS

    
 

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i Fig. 3.1. First Floor Plan of Reactor Building.

4

OlL. SYSTEM

10 FILTERS '
- AND STACK

CELL VENTILATION

AND BLOCK
VALVE

COOLANT SALT
DRAIN TANK .

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UNCLASSIFIED ,
ORNL DWG. 64-397 :

      
 
    
    
   
 
   
   
 

3 AND 10-TON
CRANES

 
 

CI0LANT
SALT PUMP

SHIELD

LIQUID WASTE CELL

DRAIN TANK CELL

RADIATOR
BYPASS DUCT—

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DRAIN TANK /

  
   
 
   

FUEL

ACTOR VESSEL
USH

RMAL SHIELD

    

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ANK
NO. 1

ORAIN LINE
FUEL DRAIN
TAN" NO. 2

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Fig. 3.2. Elevation Drawing of Reactofr Building.

 

octT

 
131

thermal neutron shield are located south of the center of the containment
cell, and they rest on the floor. The primary heat exchanger is positioned
north of the reactor vessel and above it in elevation. The freeze flanges
on the fuel and coolant salt lines attached to the heat exchanger are
placed with adequate working space for remote-maintenance operations.

The fuel-circulating pump is mounted east of the reactor vessel and
above it and the heat exchanger, sc that the free surface of salt within
the pump bowl 1s the highest point in the system. Underneath the bowl
is the overflow tank and surrocunding it is an electric furnace. All ser-
vice piping flanges and electrical Jjoints are located close to the pump
bowl so that remote replacement of the pump bowl will be easier.

The arrangement of the drain tank cell is illustrated in Fig. 1.3.
The flush salt tank is at the south end of the cell, and the two fuel
salt tanks are north of it. Space is allowed to the east and west for
remote cutting and brazing equipment in case either of the tanks must be
removed. In the northwest corner of the cell 1s a 25-kva transformer,
which supplies the power for resistance heating of the 1L l/2-in. drain
line connecting the reactor vessel to the tanks.

The coolant cell abuts the reactor cell on the south. It is a
shielded area, with controlled ventilation, but it is not sealed.

The coolant salt circulating pump is mounted high in the coolant
cell, as shown in Fig. 1.3. The radiator is at a lower elevation, to
the west, and confined in the air-cooling duct. The coolant drain tank
is underneath the pump and radiator, at the bottom of the cell. The
blowers that supply coocling air to the radiator are installed 1in an exist-
ing blower house along the west wall of the coclant cell.

Rooms containing auxiliary and service equipment, instrument trans-
mitters, and electrical equipment are located along the east wall of the
reactor, drain tank, and coolant cells. Ventilation of these rooms is
controllied, and some rooms are provided with shielding.

The high-bay area of the building over all the cells is lined with
metal to minimize inleakage. Ventilation is controlled and the area is
normally operated at slightly below atmospheric pressure. The effluent

air from this area and from all other controlled-ventilation areas is
132

filtered and monitored before it is discharged to the atmosphere. The
containment ventilation equipment consists of a filter pit, two fans, and
a 100-ft-high steel stack. This equipment on the south side of the main
building is connected to the building by one ventilation duct to the bot-
tom of the reactor cell and another along the east side of the high bay.
The vent house and charcoal beds for handling the gaseous fission
products from the reactor systems are near the southwest corner of the
main building. The carbon beds are installed in an existing pit that is
filled with water and is covered with concrete slabs. The vent house
and pit are also controlled-ventilation areas. Gases from the carbon beds
are monitored continuously for radiocactivity and are discharged into the

ventilation system upstream of the filters.

3.2 Biological Shielding

 

The MSRE work areas are divided into five types based on expected
radiation levels: (1) areas with high radiation levels that prohibit
entry under any circumstances, such as the reactor and drain tank cells;
(2) areas that can be entered a short time after reactor shutdown, such
as the radiator area; (3) areas that can be entered at low reactor power
levels, such as the special equipment room; (4) areas that are habitable
at all times; and (5) the maintenance control room, which is the only
habitable area on the site when certain large-scale maintenance operations
are being performed.

The MSRE shielding i1s designed to permit prolonged operation at 10
Mw without exposing personnel to more than a few mrem per week. The areas
which will be entered routinely and will have unlimited access will be
essentially at the normal background level for the Oak Ridge vicinity.
However, there are several class 2 and class 3 areas which will have
activity levels considerably above background and which will be entered
only occasionally. These are located near the reactor or drain-tank cell
penetrations. One such limited access area is the coolant cell, in which
the background activity might be as high as 100 mr/hr. The blower house
is also a limited access area, since the radiation level may be about 20

mr/hr near the blowers. Although the special equipment room 1s considered
133

a limited access area, the radiation field should not exceed about 10
mr/hr. The south electric service area, another limited access portion
of the building, will have a generally higher radiation level of approxi-
mately 200 mr/hr. All these estimates are based on operation of the re-~
actor at the 10-Mw power level. On the infrequent occasions when these
areas must be entered, radiation surveys will be made so that the working
time and exposures can be kept to safe levels,

When the reactor is subcritical, all areas, except Tthose of type 1
(the reactor, drain tank, and fuel processing cells), may be entered a
few minutes after the reactor is shut down and after a radiation survey
is completed. 1In general, it is eXpected that access can be on an un-
limited basis unless "hot spots' are found.

In any direction from the reactor vessel, a sufficient thickness of
shielding material has been provided to reduce the radiation level in
habitable areas to background levels. In some directions the shielding
exists as several widely separated barriers. The general arrangement is
described below, beginning at the reactor vessel.

Reactor Cell. The reactor vessel at the south side of the cell is
surrounded by a stainless steel tank with a 16-in. thickness of iron and
water. This thermal neutron shield is located within the secondary con-
tainment vessel which, in turn, is within another tank to provide a 3-ft-
wide annular sgpace, which is filled with megnetite sand and water., The
outer tank is surrounded by & cylindrical, monolithic concrete wall 21
in. thick, except in the area facing the coolant cell, 1In this area
special shielding materials are installed to give equivalent protection.
There are no equipment or personnel access openings other than those
through the ftop of the cell.

The top of the reactor cell is flat and is covered with two 3 1/2-ft
layers of large, removable, concrete blocks, as shown in Fig. 3.3, The
bottom layer is high-density (sp gr = 3+) concrete, and the upper layer
is ordinary concrete. The joints in the lower layer of blocks are filled
with steel plate inserts. A stainless steel sheet (1/8 in. thick) is
sandwiched between the two layers and is welded to the edge of the tank

to provide a gastight seal,
134

UNCLASSIFIED
ORNL DWG. 64-598

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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CELL WALL
SECTION "xXx"

Pig, 3.3. Arrangement of Shielding Blocks on Top of Reactor.
135

The service penetrations through the reactor cell walls pass through
sleeves filled with magnetite concrete grout or magnetite sand and water.
Where possible, these lines have an offset bend. The penetration of the
30-in.-diam air exhaust line through the bottom hemisphere of the con-
tainment vessel required special treatment because of the size of the
opening. A shadow shield of a 9-in. thickness of steel is provided in
front of the opening inside the cell, and a 12~in.-thick wall of stacked
blocks is erected outside the cell at the foot of the ramp to the coolant
cell,

Coolant Cell., The top and sides of the coolant and coolant drain
tank cell provide at least 24 in. of concrete shielding as protection
against activity induced in the ccolant salt while the reactor is pro-
ducing power. The large openings provided between the coolant cell and
the blower house for the cooling air supply to the radiator, however,
make it difficult to shield the blower house from this induced activity.
Space has been provided for additional shielding in the form of stacked
blocks, should they be found necessary after full power is reached.

Drain Tank Cell. The drain tank cell has a minimum thickness of

 

3 ft for the magnetite concrete walls facing accessible areas. The top
of the cell comsists of a layer of 4-ft-thick ordinary-concrete blocks
covered by a layer of 3 l/2-ft—thick ordinary-concrete blocks. The pipe
lines penetrating the cell walls have off'sets, and the smaller pipes are
cast into the walls. Shielded plugs are provided for the larger penetra-
tions.

Other Shielding. The 1/2-in. offgas line from the reactor cell to

 

the charcoal beds is shielded by 4 in. of lead as it passes through the
coolant drain tank cell. Barytes concrete blocks are stacked to a thick-
nesg of 5 ft above the line in the vent house, and 17-in.~-thick steel
plate is provided above the line between the vent house and the charcoal
beds. The charcoal beds are covered with two 18-in.-thick by 10-ft-diam
barytes concrete blocks and an additional 30-in. thickness of stacked
barytes blocks.

The walls of the filter pit for the containment ventilation system

are 12 in. thick and the roof blocks over the filters are 18 in. thick.
136

The thicknesses of the walls and tops of the auxiliary cells are
given in Table 3.1. Additional shielding is provided by blocks stacked
on the west side of the fuel processing and the decontamination cells.

Fauipment in the fuel-circulating and drain tank systems will be
repaired or replaced with remote-handling and -viewing equipment. A
heavily shielded maintenance control room with viewing windows is located
above the operating floor. This room will be used as a protected place
to operate remotely controlled equipment when several roof shielding plugs
are removed and radiocactive equipment 1s to be transferred to a storage
cell.

Fguipment in the coolant cell cannot be approached when the reactor
is operating, but the induced activity in the coclant salts is sufficiently
short lived to permit The ccolant cell to be entered for direct mainte-

nance shortly after reactor shutdown.
Table 3.1.

Descriptions of Auxiliary Cells

 

 

 

 

Location Concrete Wall Thickness Thickness
N £ Cell Inside Dimensions Fioor of Top
ame oL e N-5 E-W N-8 X B-H Flevation N s B W Blocks
Columns Columns (in.) (in.) (in.) (in.) (in.)
Tuel processing cell =5 A-B 12 £+ 10 in. X 14 £t 3 in. 831 £t O in. 18 4y 182 128 48
Decontamination cell 3-4 A-B 15 £t Q0 in. x 14 Tt 3 in. 832 £t 6 in. 18 18 18 12 30
Liguid waste cell 2-3 A-B 13 £t 0 in. X 21 £t 0 in. 828 £t 0 in. 18 18 18 18 30
Remote maintenance cell 2-3 C-D 13 £t O in. X 2L £t O in. 831 £t O in. ig 18 18 18 30
Hot storage cellP 34 C-D 15 ft 0 in. X 14 ft 3 in. 832 £t 6 in. 18 18 122 18 30
Spare cell 45 C-D 13 f4 6 in. X 14 ft 3 in. 831 £t 0 in. 18 12 12 18% 30
aPlus additional stacked blocks as required.
bThe hot storage cell is lined with ll-gage stainless steel to the

clevation of 836 ft 6 in.

LET
138

4. SITE FEATURES*
4,1 Location

The Molten-Salt Reactor Experiment is located in the Roane County
portion of the Oak Ridge area of Tennesgee. The site is owned and con-
trolled by the Atomic Energy Commission. The MSRE is in the Melton Valley
area of the Oak Ridge National Laboratory southeast of the Bethel Valley,
X-10 area, in Building 7503, which formerly housed the Aircraft Reactor
Experiment (ARE). Building 7500, which formerly housed the Homogeneous
Reactor Experiment No. 2 (HRE) and which is 2000 ft west-northwest of the
MSRE, now houses the Nuclear Safety Pilot Plant (NSPP). The High Flux
Isotope Reactor is being constructed 1500 ft south-southeast of the MSRE
site. The main laboratory area of X-10, which is approximately 1 mile
to the northwest, 1s separated from the MSRE site by Haw Ridge. Melton
Hill bounds the valley on the southern side.

The MERE site is located within a well-established AEC-controlled

area. The ARC Patrol covers the roads and adjacent area to restrict
public access to certain designated routes through the controlled land.
A perimeter fence encloses the entire area of the MSRE. Approximately
40 people will be present in the exclusion area during the day shift and
6 or 7 on the evening and night shifts.

The location is shown in respectively increasing scale in Figs. 4.1,

4,2, and 4.3.

4,2 Population Density

 

The total population of the four counties (Anderson, Knox, Loudon,
and Roane) closest to the MSRE site is 370,145. Of this number, 177,255
people are located in cities with populations greater than 2500 persons.
The rural population density in these four counties is about 135 persons
per square mile. The average population density within a radius of 27.5
miles of the MSRE site, as determined from the data obtained in the 1960

census, is 147 persons per square mile. Table 4.1 lists the surrounding

 

¥This chapter was compiled by T. H. Row, Reactor Safety Group of the
Reactor Division, Osk Ridge National Laboratory.
UNCLASSIFIED
ORNL-LR-DWG 35219

 

 

  
 
  
  
  
 
  

   

   

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140

UNCLASSIFIED
ORNL-LR-DWG 4406R

 

      
   

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Fig. 4.2. Contour Map of Area Surrounding MSRE Site.
g v R

141

 

 

 

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BASIN AREA

 
 
  

  

  
    
 
  

      

 

MICAL WASTE

 
 

LY

. T

[ . ‘»\ + = PATROL -

™ > 2R OA D gu.
[ofe —— S
> MIOWAY ENTRANCE

I \
Ofo
Olo '

N

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=
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o= '
- _ w :
. )
. L H - ;
, . : o o UT-AEC. AGRICULTURAL
// MAIN RESERVOIR oW : _ ! RESEARCH L ABORATORY
5 o ‘ i : .
' , ROAD QEFICIAL . . 2
WOAD

 

 

VvALLEY i

d ()

’

 

ARC-TVA PROTECTIVR
FENCE

SOUTHERN REGID

 
    

A
~
&
x

1000 300 O 1000 2000 3000 4000 8000
==ttty

SCALE IN FEEY

    
  
    

= 3
KEAR MOLLOW ENTRANCE~ =

  

   

N KERR—s . HOLLOW—s—ROAY
Route 62

   

’ , . Lo
7 :
X
%{ :
X ‘

 

 

UNCLASSIFIED
ORNL DWG, 63-3741

 

2y
”"w, P e

& -

 

 

o Fig. 4.3. Map of Ok Ridge Area.

 

 

 

 

bttt e A b R
 

142

Teble 4.1. Population of the Surrounding Towns
‘Based on 1960 Census

 

 City ~ Distance

Time Downwind:

 

or from Site®  Direction  Population (%)
Town‘ (miles) - Night Day

Osk Ridge 7 NNE 27,124 5.6 5.5
Lenoir City 9 SSE 4,979 4,3 6.0
Oliver Springs 9 ‘N by W 1,163 2.3 2.7
Martel 10 SE 500P 1.4 2.8
Coalfield 10 W 650P 0.5 1.1
Windrock 10 N by W 5500 - 2.3 2.7
‘Kingston 12 WSW - 2,010 9.5 11.3
Harriman | 13 W o 5,931 2.2 3.7
South Harriman 13 W 2,884 2.2 3.7
Petros | 14 W by 790P 1.4 2.8
Fork Mountain 15 ' - NNW 4 700P 2.3 2.7
Emory Gap 15 W ' 500P° 2.2 3.7
Friendsville 15 SE - 606 1.4 ‘2.8
Clinton 16 NE 4,943 11.6 2.0
South Clinton ‘16 NE 1,356 11.6 9.0
Powell 17 ENE 500P 8.3 6.8
Briceville - 19 NNE 1,217 5.6 5.5
Wartburg 20 NW by W 800P 1.4 2.8
Alcoa 20 ESE 6,395 - 2.0 2.0
Maryville 21 ESE 10, 348 2.0 2.0
 Knoxville 18 to 25 E 111,827 1.5 2.7
- Greenback B 20 S by E 960° 5.5 4,9
Rockwood 21 Wby S 5,343 2.2 3.7
Rockford 22 SE 5,345 1.4 2.8
Fountain City - 22 ENE 10,365 8.3 6.8
Lake City 23 NNE 1,914 5.6 5.5
Norris - 23 NNE 1,389 5.6 5.5
. Sweetwater 23 SSW 4,145 8.4 12.7
Neubert 27  ENE 60QP 8.3 6.8
John Sevier 27 E 7520 1.5 2.7
Madisonville 27 s 1,812 5.5 11.9
Caryville 27 N by E o 1,234P 9.5 6.1
Sunbright , 30 W 6002 0.5 1.1
Jacksboro 30 ‘ N by E 5770 9.5 - 6.1
Niota | 30 SSW | 679 8.4 12,7

 

gBased on data obtained for HFIR site.
?Taken'from 1950 census.

“

()

 
143

communities with a population of over 500 and their approximate distance
and direction from the site. The rural population density in the four
surrounding counties is given in Table 4.2. A number of facilities are
located within the AEC-controlled area, and the approximate number of

employees at each plant is given in Table 4.3. The numbers of employees

indicate the total employment at each facility and do not attempt to show
the breakdown according to shifts., However, most of these employees work

the normal 40-hr week on the day shift.

 

 

 

Table 4.2. Rural Population in Surrounding Counties
. Estimated Population

Total Rurald ~ Topulation -
County  Area®  Popula- (NDGHSItyl Within Within Within
(sq. mile) tion A peo?le) 10-Mile 20-Mile 30-Mile
per Sd. M€/ padius  Radius Radius
Anderson 338 26,600 79 395 14,200 22,800
Blount 584 38,325 66 0 6,720 23,200
Knox 517 138, 700 238 13,100 46,400 96,000
Loudon 240 18,800 78 6,080 16,900 18,700
Morgan 539 13, 500 25 225 3,625 8,630
Roane 379 12, 50C 33 3,070 9,170 11,110

 

aDoes not include area within Oak Ridge reservation.

b1960 census; does not include communities with population of 500
or more,

An estimate was made of the distribution of the population in each
of the 16 adjacent 22 l/2° sectors of concentric circles originating at

the MSRE.
radii of 0 to 1, 1 to 2, 2 to 3, 3 to4, 4 to 5, 5 to 10, and 10 to 20

Seven different distances were considered from the MSRE site:

miles. The values obtained, given in Table 4.4, are representative of

the population in this area at all times. Very little change is experi-
enced due to either part-time occupancy or seasonal variation. The popu-
lation density in the area has been reasonably stable for a number of

years and is anticipated to remain so.
144

 

 

Table 4.3. Number of Employees in Specific Oak Ridge
Areas in June 1963
Distance Total
Ares from Site Direction  Number of
(miles) Employees

MSRE 35

NSPP Q.4 WNW 6

HFIR 0.25 SSE 40

ORNL 3830
X-10 area personnel 0.75=1.25 NW 3327
Construction personnel 0.75~1.25 NwW 194
7000 area personnel 1.0-1.4 NNE 309

HPRR 1.1 ESE 12

Tower Shielding Facility L.4 S 15

EGCR 2.0 NE 150

Melton Hill Dam® 2.25 S
Construction persomnnel, June 1964 25
Normal operation {remotely con- 2

trolled;

K-25 {Gaseous Diffusion Plant) 2751
K-25 area personnel 5.0 WINW 2678
Construction personnel 73

Y-12 (Electromagnetic Separations 5,75 NINE 6866

Plant )
Y~12 area personnel 5507
ORNL personnel 209
Construction personnel 450
University of Tennessee Agricultural 6.5 NE 160
Research Laboratory
Bull Run Steam Plant® 11.25 NE
Construction personnel
July 1964 1900
December 1964 1500
July 1965 1100
December 1965 700
Normal operation (one unit) 190

 

“Estimated from construction schedules, TVA, Knoxville, Tenn.,

June 5, 1963.
Table 4.4. Estimated Population Distribution®

 

Population in Given Sector

 

 

Radius

(miles) NNE NE: FNE E ESE SE SSE g SSW W WEW W W W NITW
0-0.5 0 0 0 Q 0 0 0 0 6 0 35
0.5-1 29 0 0 0 0 0 0 0 789 1,445
1-2 309 0 150 0 12 0 15 0 221 738 18
2-3 0 0 0 24 0 41 20 0 90P 20 90 45 0 0 0 0
34, 0 0 24 41 87 40 20 135 135 135 45 0 0 0 0
45 0 0 &7 g7 90 60 40 180 180 180 45 2,751 0 200
5-10 7,944 20,428 460 7,706 7,706 7,706 83 6,546 1,567 1,564 781 781 781 781 781 781
10-20 5,320 13,318 6,650 56,414 55,914 23,131 6,388 5,660 4,700 4,700 1,563 3,573 16,741 2,542 1,500 4,190

 

aIncludes
Does not

c
Does not

Oak Ridge Plants
include Melton Hill Dam, see Table 4.3,
include Bull Run Steam Plant, see Table 4.3.

ST
146

4.3 Geophysical Features

 

4.3.1 Meteorology

Oak Ridge is located in a broad valley between the Cumberland Moun-
talns, which lie to the northwest of the area, and the Great Swmcky Moun-
tains, to the southeast. These mountain ranges are oriented northeast-
southwest and the valley between is corrugated by broken ridges 300 to
500 ft high and oriented parallel tc the main valley. The local climate
is noticeably influenced by topography.

4.3.2 Temperaturer

The coldest month is normally January, but the differences between
the mean temperatures of the three winter months of December, January,
and February are comparatively small. dJuly is usually the hottest month,
but differences between the mean temperatures of the summer months of
June, July, and August are also comparatively small. Mean temperatures
of the spring and fall months progress orderly from cooler to warmer and
warmer to cooler, respectively, without a secondary maximum or minimum.
Temperatures of 100°F or higher are unusual, having occurred during less
than one-half of the years of the period of record, and temperatures of
zerc and below are rare.

The annual mean maximum and minimum temperatures are 69.4 and 47.6°F,
respectively, with an annual wrean temperaturce of 58.5°F. The extreme low
and high temperatures are -5 and 103°F, recorded in December 1962 and
Scptember 1954, respectively. Tegble 4.5 liste the average monthly tem-
perature range based on the period 1931 to 1960, adjusted to represent
obgervations taken at the present standard location of the weather station.

Information on the temperature gradient frequency and mean wind speed

for each month were presented in a recent report2 on the meteorclogy of

 

1U.5. Dept. of Commerce, Weather Bureau, Asheville, N.C., Local Cli-
matological Data, 1962, Oak Ridge, Tennessee, Area Station (X-10),
April 23, 1963.

“W. F. Hilsmeier, "Supplementary Meteorological Data for Osk Ridge,"
USAEC Report ORO-199, Oak Ridge Operations, March 1963,
147

Table 4.5. X-10 Climatological Standard Normals
(1931 to 1960)

 

 

Maximum Minimum Aversage
Temperature Temperature Temperature
January 48.9 31.2 40.1
February 51.6 31.8 41.7
Mzrch 58.9 37.0 48.0
April 70.0 46,3 58.2
May 79.0 54.8 66.9
June 86.1 63.3 T .7
July 88.0 66.7 77 .4
August 87.4 65.6 76.5
September 83.0 59.2 71.1
October 72.2 47.7 60.0
November 58.6 36.5 47,6
December 49 .4 31.3 40 .4
Annual 69.4 4'7.6 58.5

 

the Oak Ridge area. The seasonal and annual averages derived from this

information are presented in Fig. 4.4.

4.3.3 Precipltation?

Precipitation in the X-10 area is normally well distributed through-
out the year, with the drier part of the year occurring in the early fall.
Winter and early spring are the seasons of heaviest precipitation, with
the monthly maximum normally occurring January to March. A secondary
maximum occurs in the month of July that 1s due to afterncon and evening
thundershowers. BSeptember and Octobery are usually the driest months.

The average and maxinmum annual precipitation are 51.52 and 66.2 in.,
respectively. The maximum rainfall in the area in a 24-hr period was
7.75 in., recorded in September 1944. The recurrence interval of this
amount of precipitation in a 24-hr period has been estimated to be about
70 years. The maximum monthly precipitation occurs normally in March and
has a value of 5.44 in. The average monthly precipitation is given in

Table 4.6.
FREQUENCY (%)

Fig. 4.4.

148

UNCLASSIFIED
ORNL-DWG 63-2543

[] oar NIGHT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

20
15
10
5 e
0 ' .
20
15 [ SPRING .
10 e
5
/, v
20
15 ] SUMMER
10 l
5
0 //A
20
15 FALL
10 =] 7
5 _,!{Z}.l ’{; ’ 7 // s 47/?: - *
20
15 ANNUAL
10
5 7 7%
nr
. }7,/
o ‘

 

 

 

 

 

 

-2.4 -7 -0 —0.3+0.4 +14 +1.8 +2.5
TEMPERATURE GRADIENT PER 100 ft (°F)

Seasonsal Temperature Gradient Frequency.
149

Table 4.6. X-10 Average
Monthly Precipitation

 

 

Data
Morth Precipitation

(in.)
January 5.24
February 5.39
March 5.4%
April 4. 14
May 3.48
June 3.38
July 5.31
August 4.02
September 3.59
October 2.82
November 3.49
December 5.22

 

Light snow usually occurs in all months from November through March,
but the total monthly snowfall is often only a trace. The total snowfall
for some winters is less than 1 in. The averagge snowfall for the period
from 1948 to 1961 is 6.9 in. The maximum snowfall in a 24-hr period was
12 in. in March 1960. The maximum monthly snowfall, 21 in., also occurred
in March 1960. The heavy fogs that occasionally occur are almost always

in the early morning and are of short duration.

4.3.4 Wind?

The valleys in the viecinity of the MSRE site are oriented northeast-
southwest, and considerable channeling of the winds in the valley may be
expected, This is evident in Fig. 4.5, which shows the annual frequency
distribution of winds in the vicinity of CRNL. The flags on the wind-rose
diagrams point in the direction from which the wind comes. The prevailing
wind directions are upvalley from southwest and west-southwest approxi-

mately 40% of the time, with a secondary maximum of downvalley windsg from

 

3W. B. Cottrell, ed., "Aircraft Reactor Experiment Hazards Summary
Report,' USARC Report ORNL-1407, Oask Ridge National Laboratory, November

1952.
150

 

 

 

 

 

Feet

<

5

 

 

icinity

the V

in

lon of Winds

istribut

Annual Freaquency D

D,

J,

of X-10 Aresa.

Fig
151

northeast and east-northeast 30% of the time. The prevailling wind regimes
reflect the orientation of the broad valley between the Cumberland Plateau
and the Smoky Mountains, as well as the orientation of the local ridges
and valleys. The gradient wind in this latitude is usually southwest or
westerly, so the daytime winds tend to reflect a mixing down of the gradi-
ent winds. The night winds represent drainage of cold air down the local
slopes and the broader Tennessee Valley. The combination of these two
effects, as well as the daily changes in the pressure patterns over this
area, gives the elongated shape of the typical wind roses.

During inversions, the northeast and east-northeast winds occur most
frequently, usually at the expense of the southwest and west-southwest
winds. The predominance of light northeast and east-northeast winds under
stable conditions is particularly marked in the summer and fall when the
lower wind speeds aloft and the smaller amount of cloudiness allow the
nocturnal drainage patterns to develop.

Wind roses prepared from 5 years of data, 1956 to 1960, are shown
in Figs. 4.6 and 4.7 (ref. 2). These represent the wind direction, fre-
quency, and percentage of calm under inversion and lapse conditions for
the X-10 area and are applicable to the MSRE location.

Considerable varistion is observed both in wind speed and direction
within small distances in Bethel Valley and Melton Valley. It may be
sald that in nighttime or in sftable conditions, the winds tend to be gen-
erally northeast and east-northeast and rather light in the valley, re-
gardless of the gradient wind, except that strong winds aloft will con-
trol the velocity and direction of the valley winds, reversing them or
producing calms when opposing the local drainage. In the daytime, the
surface winds tend to follow the winds aloft, with increasing reliability,
a8 the upper wind speed increases. Only with strong winds aloft or winds
paraliel to the valleys would it be of value to attempt to extrapolate
alr movements for any number of miles by using valley winds. In the well-
developed stable situation, however, a very light air movement will fol-
low the valley as far downstream as the valley retains its structure,
“even though the prevailing winds a few hundred feet above the ground are

in an entirely different direction. In a valley location, the wind is
152

Uneclagsified
ORNL-DWG 63-2943

WINTER SPRING

   

(8198 obs)

 

LAPSE

SUMMER FALL

   

Ve ( 60il obs)
wind speed, mph
% Calm
-4 5-9 10-14 15-19 > 20
% Frequency
Q 3 10 15 20

1956 -1960 Data

Fig. 4.6. X-10 Area Seasonal Wind Roses.
153

Unclassified
CRNL-DWG 63-2944

WINTER

SPRING

   
   

   
   

(4515 obs) (3959 obs)

INVERSION

SUMMER FALL

 

{ 4792 obs)

/)/' (4434 obs)

Wind speed, mph
% Calm

—
@.4 5-9 10-14 15-19 2 20
M

% Freguency

Q D 10 15 20

M [ [ ]
1956 -1960 Data

Fig. 4.7. X-10 Area Seasonal Wind Roses.
154

governed by the local valley wind regime and the degree of coupling with
the upper winds.?

The wind flow between Melton Valley and the X-10 area was investi-
gated in conjunction with the Aircraft Reactor Experiment (ARE) hazards

3 Two patterns of wind flow were assumed to be of significance:

analysis.
(1) from the 7500 area northwest over Haw Ridge to the X-10 area, and

(2) from the 7500 area west to White Oak Creek, then northwest through

Haw Gap, and finally north to the X-10 area. The frequencies of these +
patterns during the period September to December 1950 were normalized to

the 1944-t0-1951 wind record at the X-10 area by a ratio method.? The )

normalized frequencies are listed in Table 4.7.

Table 4.7. Frequency of Wind Patterns Between
7500 and X-10 Area

 

Frequency of Wind
Pattern (%)

 

 

Over Ridge Through Gap
A1l observations 2.5 0.4
Day (9 am to 5 pm) 4.3 0.6
Night (9 pm to 5 anm) 0.0 0.4
Light wind (1 tc 4 mph) 2.6 0.4
Stronger wind (5 mph and over) 2.7 0.3

 

A comparison of the pibal* cbservations made throughout 1949 to 1950
at Knoxville and Oak Ridge shows that above about 2000 ft the wind roses
are almost identical at these twe stations. This identity in the data
makes possible the use of the longer period of record (1927 to 1950) from
Knoxville to eliminate the sbnormzalities introduced by the use of the
short record at QOsk Ridge. Annual wind roses are shown for Knoxville
(1927 to 1950) and Nashville (1937 to 1950) in Fig. 4.8. Pibal observa-
tions are only made when no clouds, dense fog, or precipitation is occur-

ring and so are not truly representative of the upper wind at all times.

 

*¥Pilot balloon visual observations. -
155

Three years of Rawin* data for Nashville (1947 to 1950) are available
that consist of observations taken regardless of the current weather at
the time of observation. A comparison of these wind roses for Knoxville
and Nashville shows that the mode for winds above 3000 meters mean sea
level should be shifted to westerly instead of west-northwest when obser-
vations with rain are included in the set. In summer and fall, the winds
aloft are lightest and the highest winds occur during the winter months
at all levels.

The northeast-southwest axis of the valley between the Cumberland
Plateau and the Smoky Mountains continues to influence the wind distri-
bution over the Tennessee Valley up to about 5000 ft, although the varia-
tions within the Valley do not extend above about 2000 ft. Above 500 ft,
the southwesterly mode gives way to the prevailing westerlies usually ob-

served at these latitudes.
Previous investigation of the relation of wind directiocn to precipi-
tation indicates that the directicon distribution is very little different

3 This is consistent with the experience

from that of normal observations.
of precipitation forecasters that there is little correlation between sur-
face wind direction and rain, particularly in rugged terrain. Figure 4.8
shows the upper winds measured at Nashville during the period 1948 to
1950, when precipitation was occurring at observatbtion time. In general,
the prevailing wind at any given level is shifted to the southwest or
south-gsouthwest from west or southwest, and the velocity is somewhat
higher during the occurrence of precipitation, with the shift being most
marked in the winter.

The percentage of inversion conditions of the total hours for the
seasons is given in Table 4.8 (ref. 2).

Tornadoes rarely occur in the valley between the Cumberland and the

Great Smokieg, It is highly improbable that winds greater than 100 mph
would ever be approached at the MSRE site.

 

¥Radio wind balloon observations.
156

UNCLASS|FIED

   

     

  

DWG. 16972
KNOXVILLE PIBAL NASHVILLE PIBAL NASHVILLE RAWIN NASHVILLE RAWIN
ALL OBS. PREGIPITATION
500m
25659 Obs. 20321 Obs 3341 Obs.
I500m
Q::\‘?‘ ~
.:33:;;. Ao
23449 Obs. {7940 Obs 3269 Obs 227 Obs.
3000m
14710 Obs 12276 Obs. 3193 Obs. 217 Obs.
6000m

  

3692 Obs 2964 Obs. 2684 Obs. 170 Obs.

10000m

  

 

 

536 Obs. TI7 Obs 1857 Obs &
% Calm Wind Speed M PH % Frequency
i-ig_1!-33 34-56 57-85 86-1i4 {15+ 10 0 10 20 30

- — I N

PIBAL —PILOT BALLOON VISUAL OBSERVATIONS
RAWIN—RAD!O WIND BALLOON OBSERVATIONS

Fig. 4.8. Wind Roses at Knoxville and Nashville for

Various Alti-
tudes.
157

Table 4.8. Summary of Seasonal Frequency of Inversions

 

Frequency of Inversion Average Duration@

 

Season (%) (hr)
Winter 3L..8 8
Spring 35.1 9
Summer 35.1 9
Fall 42 .5 10
Annual 35.9

 

. M. Culkowski, AEC, Oak Ridge, personal communi-
cation to T. H. Row, Oak Ridge National Laboratory,
July 8, 1963.

4,3.5 Atmospheric Diffusion Characteristics

 

Sutton's*s® methods of calculating the dispersion of airborne wastes
require a knowledge of several meteorological parameters. Values of these
parameters of the X-10 site were recommended by the Oak Ridge Office of
the U,S. Weather Bureau® and are listed in Table 4.9. They provide &

 

“0. G. Sutton, "A Theory of Eddy Diffusion in the Atmosphere,' Proc.
Roy. Soc. (London), 1932.

°0, G. Sutton, Micrometeorology, McGraw-Hill Co., Inc., New York,
1953.

®"pinal Hazards Summary Report, Experimental Gas-Cooled Reactor, "
Vol. 1, Book 1, pp. 3-6, USAEC Report ORO-586, Oct. 10, 1962.

 

 

Table 4.9. Meteorological Parameter Values for
Atmospheric Dispersion Calculations

 

 

Parameter Lapse Inversion
(weak)
Mean wind velocity, p (m/sec) 2.3 1.5
Stability parameter, n 0.23 0,35
Diffusion constant, Cy (m/ ?) 0.3 0.3

Diffusion constant, C_ (m/ 2) 0.3 0.033

 
158

reasonably conservative basis for the dispersion calculations in Section
8.7.2 and Appendix C; however, it is recognized that extreme meteorologi-
cal conditions will occasionally exist and could produce exposures greater

than those based on the values in the table.

4,3.6 Environmental Radioactivity

 

Atmospheric contamination by long-lived fission products and fallout
occurring in the general environment of the Oak Ridge area is monitored
by a number of stations surrounding the area. This system provides data
to aid in evaluating local conditions and to assist in determining the
spread or dispersal of contamination should a maJjor incident cccur. Data
cn the environmental levels of radiocactivity in the Oak Ridge area are
given in Tables 4.10, 4.11, 4.12, and 4.13 (ref. 7).

These levels of activity are, of course, strongly influenced by the
radiocactive wastes discharged by the various facilities at ORNL. The
MSRE will normally release only the gaseous wastes xenon and krypton at
rates of C.5 and 6.4 curies/day, respectively, and thus should not make

a significant contribution to the existing activity levels.

4.3.7 Geology and Hydrology

 

The bedrock beneath the MSRE site is a dark gray calcareous clay
shale with a bearing value cf © tons/ftz when unweathered. The overbur-
den on this fresh unweathered shale averages 20 ft in thickness and con-
sists of a thin vlanket of organic tepsoll, generally less than 1 T in
thickness, on top of weathered shale. The latter, if confined, has a
bearing value of 3 tons/ftz. The strata of the shale are highly folded
and faulted, and the dip, although it averages about 35° toward the south,
is dirregular and may vary from horizontal to vertical. The Melton Valley
in which the MSRE is located is underlain bty the Conasauga shale of the
Middle and Upper Cambrian Age. The more resistant rock layers of the
Rome formation, steeply inclined toward the southwest, are responsible

for Haw Ridge, which parallels the valley immediately to the northwest.

 

73. ¢. Hart (ed.), "Applied Health Physics Annual Report for 1962,"
USAEC Report ORNL-3490, Oak Ridge National Laboratory, September 1963.
 

159

 

!

 

Teble 4.10. Concentration of Radioactive Materials in Air in 19628

 

 

 

 

 

/ fl N Nunber of Particles (in dis- Number of
: o Long-Lived integrations per 24 hr) Total -
s | ;fii"n Location Activity by Activity RangesP Number of Iarvicles
er . Per
, -~ (pefec) . Particles . ;,0q pi3
‘ : L <10° 10°-10% 10%5-107 >107
a. '%aboratory7Area ' ,
x 10713 - | .
\ HP-1 S 3587 38 128 1.6 0.00 0.00 129 3.1
- HP-2 NE 3025 , 43 122 1.9 0.04 0.00 . 124 3.5
HP-3 SW 1000 - 37 129 2.1 0.10 0.02 131 2.1
HP-4 W Settling Basin 21 91 1.2 0.04 0.00 93 1.6
HP-5 E 2506 51 115 1.2 0.04 0.04 117 3.9
HP-6 SW 3027 . . _ 33 136 1.5 0.02 0.02 137 2.4
HP-7 W 7001 ‘ 40 115 1.8 0.00 0.00 117 2.3
HP-8 .Rock Quarry , 39 132 1.5 0.00 0.02 . 133 2.5
HP-9 N Bethel Valley Rd. 31 145 1.6 0.00 0.00 146 2.3
- HP-10 W 2075 38 126 1.3 0.00 0.00 128 3.1
Average 37 124 1.6 0.02 0.01 125 2.7
; - Perimeter Ares
HP-31  Kerr Hollow Gate 3% 135 1.6 0.04 0.04 137 2.7
HP-32  Midway Gate 37 132 2.1 0.02  0.00 134 2.6
HP-33  Gallaher Gate 32 113 1.4 0.00 0.02 114 2.2
‘ AP-34  White Wing Gate 3 153 )‘ 1.5 0.00 0.00 155 3.0
: HP-35  Blair Gate 39 168 1.6 0.00 0.02 169 3.3
f HP-36  Turnpike Gate 39. 158 2.2 0.02 0.04 161 3.2
I HP-37 Hickory Creek Bend T 34 © 114 1.6 0.02  0.00 115 2.3
; Average - 36 139 1.7 0.01 0.02 141 2.8
| | ' Remote Area
HP-51  Norris Dam . L A3 139 2.3 0.04 ~ 0.00 141 2.6
HP-52  Loudown Dam - .42 0 130 2.8 0,10 ~ 0.00 133 2.4
HP-53 ° Douglas Dam - .. 44 " " 150 2.6 0.02- 0.00 - 153 2.8
"HP-54 . Cherokee Dam L A0 164 2.4 0.04 - 0,02 © 167 3.0
HP-55  Watts Bar Dam =~ =~ 45 - 157 2.0 0.04 0.00° = 159 2.9
. “HP-56 - Great Falls Dam 46 71667 2.3 0.00 0.00 = 1p8 3.1
' HP-57  Dale Hollow Dam ' . . 38 .7 171 "~ 1.6 0.00 0.04 - 172 2.9
L Average - 43 . 154 2.3 0.03  0.0L 157 2.8

 

K 8pveraged weekly from filtérffiéperfidata;  , - )
Ppetermined by filtration techniques, .- . ' 5

 

 

 

«

v e . - 1 — - e o e e e e B i

 
 

160

Teble 4.,11. Radioparticulate Fallout in 19628

 

s

“Number of Particles (in dis-

 

 

Statl Long-Lived integrations per 24 hr) | . Total Total
,Nu:b on Location Activity by Activity Ranges . Number of Particles
er ' (pe/1t2) — Particles  per ft2,
s - <10® - 10°-10% - 10%-107  >107
Laboratory Area
x 10-12 | ) ,
HP-1 S 3587 ' 15 79 2.1 0.12 0.06 8L 42
HP-2 NE 3025 : 17 88 2.3 0.04 0.06 91 .49
HP-3 SW 1000 15 83 2.0 0.15. 0.06 86 42
HP-4 W Settling Basin 14 73 2.3 0.08 0.04 75 48
HP-5 E 2506 : 14 86 2.0 0.08 0.04 .91 50
HP-6 SW 3027 16 101 2.9 0.02 0.02 104 61
HP-7 W 7001 15 89 2.4 0.02 0.06 92 49
BP-8 Rock Quarry 17 89 2.6 0.00 0.08 9 46
HP-9 N Bethel Valley Rd. 16 88 2.9 0.06 0.12 91 41
HP-10 W 2075 15 100 2.3 0.04 0.00 103 55
' Average 15 88 2.4 0.06 0.05 91 48
i ‘ Perimeter Ares _
HP-31  Kerr Hollow Gate 17 103 2.13 0.13 0.10 105 &7
HP-33  Gallaher Gate 14 82 2.4 0.10 0.00 85 - 42
HP-34  White Wing Gate 18 104 2.2 0.19 0.08 106 47
HP-35 = Blair Gate 15 124 2.0 0.06 0.04 126 50
HP-36  Turnpike Gate 16 109 3.5 0.08 0.02 112 57
HP-37  Hickory Creek Bend 16 85 2.3 0.04  0.08 87 47
Average , 16 101 2.5 0.10 0.05 103 48
Remote Area .
HP-51  Norris Dam 14 86 2.2 0.12 0.04 89 T 36
HP-52  Loudoun Dam 13 70 2.7 . 0.06 0.06 7 29
HP-53  Douglas Dam 13 T 2.7 0.06 0.08 . 80 35
HP-54 Cherokee Dam 14 11 2.9 - 0.13 0.06 84 .35
HP-55  Watts Bar Dam 16 81 2.2 0.14 0.08 83 37
HP-56  Great Falls Dam 14 - 98 2.2 0.06 0.02 100 39
HP-57  Dale Hollow 14 96 2.0 '0.08 0.06 98 33
' Average ‘14 84 2.4 0.09  0.06 87 - 35

 

E"m'era.geél' weekly from gummed-paper data.

¢

iy

 
161

Table 4.12. Concentration of Radioactive
Materials in Rain Water in 19622

 

Activity in

 

Station Location Collected
Number Rain Water
(uc/cc)
Laboratory Ares
x 1077
HP-7 West 7001 10.3
Perimeter Ares
HP-31 Kerr Hollow Gate il
HP-32 Midway Gate 12
HpP-33 Gallaher Gate 10
HP-34 White Wing Gate 11
HP-35 Blair Gate 11
HP-36 Turnpike Gate 10
HP~-3"7 Hickory Creek Bend 11
Average 11
Remote Ares
HP-51 Norris Dam 14
HP-52 Loudoun Dam 1L
HP-53 Douglas Dam 13
HP-54 Cherokee Dam 11
HP-55 Watts Bar Dam 14
HP-56 Great Falls Dam 16
HP-57 Dale Hollow Dam 11
Average 13

 

aAveraged weekly by stations.
Table 4.13. Radloactive Content of Clinch River in 1962

 

 

 

‘ . . . Average
Concentration of N%clldes of Primary Concern Concentration Radiocactivity
_ ne/cc) (MPC ).,
Location of Total a as Percentage
: - (ne/ce)
qr20  geldd  (0gl37  Rul03-106 (60 .95 yn 95 Radiocactivity of (MPC),,
(nc/ce)
x 10-8 x 1078 x 108 x 107®% x 10-8 x 1078 x 1078 x 1070
Mile 41.5 0.16 0.14 0.02 0.78 (o) 0.42 1.5 0.90 1.7
Mile 20.8° 0.15 0.02 0.09 21 0.18 0.09 34 4.6 7.4
Mile 4.54  0.34  0.20  0.07 16 0.32 0. 54 17 3.5 4.9

 

“Weighted average maximum permissible concentration (MPC )y calculated
for the mixture by using values for specific radionuclides recommended in
NBS Handbook 69. ’

bNone detected.

cValues given for this location are calculated values based on the
levels of waste released and the dilution afforded by the river.

Center's Ferry (near Kingston, Tennessee, just above entry of the
Fmory River into Clinch River).

c9T
163

These layers dip beneath the shales of the Conasauga group in Melton
Valley. The shale layers in the area are in keeping with the general
structure of the surrounding area, as reported in a geological survey of
the area.® Conasauga shale, a dark red shale, contains thin layers and
lenses of limestone that are generally irregular in distribution. How-
ever, there are no persistent limestone beds in the upper strata of the
shale layers and, consequently, no underground solution channel or cavern
to permit rapid and free underground discharge of water.

The dominantly clay soils of the Oak Ridge area are generally of low
permeability, so the surface runoff of water is rapld. Cbservations in
test wells show that the Conasauga shale, although relatively impermeable,
is capable of transmitting small amounts of water through the soll a dis-
tance of a few feet per week. Furthermore, all the active isotopes, ex-
cept ruthenium, apparently become fixed in the immediate vieinity of the
point of entry into the soil. It is concluded that such ground water
flow as may exist below the surface in the soil surrounding the site will
be small and slow (a few feet per week) and that natural chemical fixation
will reduce the level of the activity of mixed, nonvolatile fission prod-
ucts more than 90%.

The Conasauga shale Tormation is extensive, quite heterogeneous in
structure, and relatively low in permea’bility.8 The depth and extensive-
ness of the formation provide a large capacity for decontamination of any
contaminated liguid reaching it, and the slow rate of percolation improves
the efficiency of the decontamination by ion exchange and radioactive de-
cay. However, the heterogeneity of the formation makes difficult the
prediction of pattern and rate of movement of water and hence of the fis-
sion products contained therein. Most of the seepage is along bedding
plains parallel to the strike.

Storm drains for the MSRE site will discharge into Melton Branch.

The area storm sewers consist of two individual trunk systems. One trunk

 

8p, B. Stockdale, "Geologic Conditions at the Ozk Ridge National
Taboratory (X-10 Area) Relevant to the Disposal of Radiocactive Waste,”
USAEC Report ORO-58, Oak Ridge Operations, August 1951.
164

system empties directly into Melton Branch (see Fig. 4.5) through a con-
necting ditch; the second trunk system combines with the cooling tower
blowdovn and the flow from the process waste pond before discharging into
Melton Branch through a connecting ditch. The storm drainage system is
based on a runoff coefficient of 30% for unpaved or seeded areas. The
high runoff coefficient applied to the unpaved areas is justified by the
highly impervious character of the solls present in the plant area.
Downstream from ORNL, numerous uses are made of the Clinch River
water by both municipalities and industries, as indicated in Table 4.14.
The first downriver water consumer is the K-25 plant, whose water intake
is located at river mile 14.4. Representative flows of the Clinch, Hmory,
and Tennessee Rivers are listed in Table 4.15. The local flow pattern in
the Watts Bar reservoir during the period May through September is sig-
nificantly affected by the differences in water temperature of the Clinch
River and the Watts Bar reservoir. When the Clinch River is considerably
cooler, stratified flow conditions caused by density differences may exist,
with the cooler water flowing on the bottom beneath the warmer water.
This phenomenon markedly affects the travel time of water through the
reservoir and complicates the analysis of flow. 1In addition, during the
period of stratified flow, some Clinch River water may flow up the Emory

River as far as the Harriman water plant intake.

4.3.8 Selsmology

Information cn the frequency and severity of earthquakes in the East
Tennessee area is reported in the ART Hazards Report.9 Barthquake forces
have not been considered in the design of facilities at CRNL or TVA struc-
tures in the Fast Tennessee areca. The Oak Ridge area is currently classi-
fied by the U.S. Coast and Geodetic Survey as subject to earthquakes of

intensity mm-6 measured on the modified Mercalli Intensity Scale.

 

W. B. Cottrell et al., "Aircraf: Reactor Test Hazards Summary Re-
port," USAEC Report ORNL-1835, Oak Ridge National Laboratory, January
1955.
Table 4.14.

Community Water Systems in Tennessee Downstream from ORNL Supplied by
Intakes on the Clinch and Tennessee Rivers or Tributaries®

 

 

Commmity Population Intake Source Approx?mate Remarks
Stream Location
ORGDP — K-25 Ares 2,678P Clinch River CR Mile 14 Industrial plant water system
Harriman 5,931° Fmory River FR Mile 12 Mouth of Emory River, CR Mile 4.4
Kingston Steam Plant 5004 Clinch River CR Mile 4.4
(TVA)
Kingston 2,0004 Tennessee River TR Mile 570 River used for supplementary sup-
ply
Watts Bar Dam (resort l,OOOd Tennessee River TR Mile 530
village and TVA steam
plant )
Dayton 3,500°  Richland Creek  RC Mile 3 Opposite TR Mile 505
Cleveland 16,196° Hiwassee River HR Mile 15 Mouth of Hiwassee River is at
TR Mile 500
Soddy 2,000d Tennessee River TR Mile 488
Chattanooga 130,009¢ Tennessee River TR Mile 465 Metropolitan area served by City
Water Company
South Pittsburg 4,130¢ Tennessee River TR Mile 435
Total 168,061

 

“HGCR Hazards Summary Report, USAEC Report ORO-586, Oak Ridge Opera-

tions, Oct. 10, 1962.

bBased on May 1963 data.
©1960 Report of U.S. Bureau of the Census.
dBased on published 1957 estimates.

GOt
Table 4.15. Flows in Clinch, Emory, and Tennessee Rivers, 1945-1951%

 

 

 

 

_ Flow (cfs)
Clinch River Emory River, . Tennessee River
Miles 20.8 and 13.2 Mile 4.4D Mile 12.8 Mile 529.9 Mile 465.3

 

Maximum  Mean Minimum Msximum Mean Minimm Maximum Mean Minimm Maximum Mean Minimm Maximum Mean Minim

 

January 22,900 8,960 1620 70,700 14,400 2120 50,000 4450 178 181,000 47,700 15,800 218,000 63,900 23,200
February 27,700 10,100 1230 88,800 15,800 2250 69,000 4550 468 204,000 50,800 17,900 195,000 67,900 19,700

March 12,700 5,850 690 26,700 9,450 1830 15,400 2910 507 100,000 32,400 13,300 148,000 44,200 19,500
April 8,540 3,400 306 13,300 5,620 752 6,600 1800 249 43,700 23,800 5,200 82,000 27,700 13,200

~ 6,600° 9, 730¢ 25200 34,900° .. " 45,900¢
Mey . 8,080 2,750 208 19,700 4,700 520 13,100 1610 58 38,300 20,000 3,000 95,900 28,800 17,900
June 7,420 2,820 224, 9,280 3,320 262 5,300 396 14 32,100 18,900 8,200 32,800 25,500 18,900
July 7,630 2,930 259 12,800 3,400 281 9,230 360 21 87,500 19,600 6,500 106,000 26,400 15,400 §_
August 8,390 4,520 374 8,760 4,800 378 3,060 177 4 37,600 21,400 9,900 43,300 28,100 17,800 ‘
September 8,450 4,620 341 13,000 4,940 462 5,500 224 2 39,900 22,200, 6,900 54,400 30,100 17,100
| 3, 5304 4,2304 5534 20,4008 | 27,8004
October 9,200 5,130 150 14,200 5,300 150 5,040 93 1 67,100 23,500 9,800 72,800 29,700 16,300
November 12,700 4,430 453 40,500 5,830 556 27,800 1100 2 128,000 25,400 10,300 167,000 34,000 13,600
December 27,000 8,360 569 - 60,300 11,700 593 33,300 2720 24 112,000 41,000 11,800 139,000 53,600 21,100

 

®EGCR Hazards Surmary Report, USAEC Report ORO-586, Oak Ridge Operations
Office, Oct. 10, 1962. TFilow data were collected prior to construction of
Melton Hill Dem on Clinch River. Although current data are not available,
flows are not expected to be significantly changed.

bFlows shown for Clinch River Mile 4.4 include Emory River flows.
“Nonstratified flow period.
Astratiried flow period.

®By egreement with TVA, a flow of not less than 150 cfs has been main-
tained in the Clinch River at Oask Ridge since August 28, 1943, ' -

 

 
167

Both LynchlO of the Fordham University Physics Department and
Moneymakerll of the Tennessee Valley Authority indicate that the earth-
quakes that occasionally coccur in the East Tennessee area are dquite common
in the rest of the world and are not indicative of undue seismic activity.

An average of one or two earthquakes a year occurs in the Appalachlan
Valley from Chattancoga to Virginia according to TVA records. The maxi-
mum intensity of any of these shocks recorded is 6 on the Woods-Neuman
Scale. A gquake of this magnitude was experienced in the Oak Ridge area
on September 7, 1956, and was barely noticeable by either ambulatory or
stationary individuals. Structures were completely unaffected. Distur-
bances of this type are to be expected only once every few years in the
Oak Ridge ares.

The Fordham University records indicate a quake frequency below that
estimated by TVA. However, the magnitude of the observed quakes is ap-
proximately the same. Lynch indicates that "it is highly improbable
that a major shock will be felt in the ares (Tennessee) for several thou-

sand years to come."

 

L0Letter from J. Lynch to M. Mann, November 3, 1948, quoted in a re-
port on "The Safety Aspects of the Homogeneous Reactor Experiment,"” USARC
Report ORNL-731, Oak Ridge National Laboratory, June 1950.

11B, C. Moneymaker, Tennessee Valley Authority, personal communica-
tion to W. B. Cottrell, Oak Ridge National Laborstory, October 1952.
168

5. CONSTRUCTION, STARTUP, AND COPERATION

5.1 Cecnstruction

Although no special constructicn practices were employed in assem-
bling the MSRE, meny special precauticns were taken to ensure a high-
quality, clean, and leak-tight assembly. A detailed specification re-
quiring better cuality contrcl than that of existing commercial cocdes was
prepared for each compenent of the system. Nondestructive inspection
techniques, such as ultrascnic testing, dye-penetrant inspection, X-ray
examination, and helium leak testing, were employed at each fabricator's
plant under the supervieion of ORNL inspectors. Assembly at the reactor
site is being examined similarly. After completicn, the separate systems
will be leak tested using rate-cf-pressure rise, mass-spectrometer, and
isotcpic~-tracer technicues.

During the last stages cf constructicn and for several weeks after
censtructicn 1s completed, the reactor ecuipment will be checked cut for
acceptance by the Operations Department. The individual systems, that
is, the cover gzs, offgas, pump cil, cooling water, component air-cocling
systems, etc., will be cperated fcr the first time. Remote mzintenance
will be practiced during this period also. The successive steps in the
startup of the reactor are listed in Table 5.1. The most important phases

are discussed below.

5.2 Flush-8al1lt Operation

 

It is planned to demonstrate the mechanical performance of the as-
sembled reactor by & several-month period of testing with & flush salt in
the fuel system. Simultanecusly, ccclant salt will be circulated in the
ccolant system. Each piece cf equipment will be examined tc determine
whether 1t perfcrms as designed, insofar as this can be determined with-
out nuclear heat generation. The flush salt will zlso serve to scavenge
oxygen and to remove other impurities. Another important benefit of the
flush-salt test will be the development of the operating skills necessary

Tor satisfactory contrcl of the system variables. Heat-balance methods
169

Table 5.1. MSRE Startup Plans

 

 

Startup Phase Month
Complete Reactor Installation 0
Operator and Supervisor Training 1-6
Equipment Checkout — Without Salt 1-3
1. Heaters, thermocouples, instruments, and logger
2. Cover-gas, offgas, pump-oil, cooling-water, component-air-
cooling, and radiator-cooling systems
3. Air-compressor, containment-ventilation, diesel-power, and
electrical systems
4. Remote maintenance
Flush Salt Operation 3=5
1. Coolant and flush-salt loading, inventory, and transfer
methods
2. Coolant-circulating system
3. Fuel-circulating system
4. Bampling and leakage detection
5. Heat-balance methods, heat-removal systems, and temperature
control
6. Graphite examination
Precritical Shutdown o=
1. Chemical plant performance (hydrofluorination of flush salt)
2. Loading of fuel salt
3. Completion of nuclear instrumentation and control-rod tests
4. Final maintenance
Critical Experiments 69
1., Check performance of nuclear instrumentation
2. Load fuel to critical
3. Measure pressure and temperature coefficients of reactivity,
control-rod worth, response to flow changes
4. Establish baselines for determination of effects of power
on chemistry, corrosion, and nuclear performance
Posteritical Shutdown 10
1. Seal-weld membranes on reactor and drain tank cells
2. Final check on contaimment leakage
3. Fill and check out vapor-condensing system
Approach to Full Power 11-14
1. Heat balances, heat-removal control
2. Shield and containment surveys
3. Sampling program for chemistry, etc.
4. Measurement of power coefficient, xenon polsoning, fuel

5

permeation of graphite, and offgas composition
Reactor control and logger performance

 
170

and temperature control will be practiced, and sampling operaticns will
be tried.

After the flush-salt cperating periocd of one to three months, the
flush salt will be transferred tc the chemical plant, where it will be
treated tc remove accumulated cxides in a trial performance of the chemi-
cal plant. While the reactor system is shut down, the fuel salt will be
lcaded into cne cf the drain tanks, with about two-thirds of the U?35 es-
timated for criticality, and the multiplication of the fuel will be meas-
ured at several liquid levels. Alsc during this shutdown, nuclear in-
strumentaticn and contrcl reds will be given a final checkout, and remote-

maintenance practice will be completed.

5.2.1 Critical Experiments

 

After the reactor system has been heated to 1200°F and neutron count-
ing rates have been measured with the reactor empty, the fuel mixture,
with about two-thirds of the final U??° concentration, will be slowly
transferred to the reactcr. Ccunting rates will be determined with the
control rods in varicus positions and a2t several temperatures. If criti-
cality is not attained, the fuel salt will be drained and more U2?3° will
be added and mixed with the original salt before refilling the core, This
procedure will be repeated after every fuel addition. When the counting
rates indicate that criticality 1s near, further fuel additions will be
made in approximately 100-g increments at the fuel-circulating pump through
the sampling mechanisn.

When the critical mass has been determined, the temperature and pres-
sure ccefficients will be measured, and the ccntrol-rod calibrations will
be completed. Next, the effective fraction of delayed neutrons will be
studied to learn the fraction lost by circulation. The reactor must be
cperated at zerc power for a pericd of one to three months to establish
baselines for determinaticn of the effects ¢f power operation on salt

chemistry, corrcsion, and nuclear performance.

5.2.2 Power Operation

 

When the baseline data at zero power are judged to be satisfactory,

the power will be increased in increments of several hundred kilowatts
171

over a period of six to ten weeks. At each successively higher power,
samples will be collected and measurements will be made of heat output,
power coefficient, radiation levels inside and outside shielding, xenon
poisoning, offgas composition, and possible permeation ¢f the graphite by

Tuel.

5.3 Operations Perscnnel

 

The operation of the MSRE facility will be the responsibility of the
Reactor Division Operaticns Department. The previcus assignments of per-
sonnel in this group Include the ccnstruction, startup, and operation of
four other experimental reactors: the Low-Intensity Test Reactor, the
Homogeneous Reactor Experiments 1 and 2, and the Alrcrarft Reactor Experi-
ment.

The experiment will be conducted con a three-shift basis, emplcying
four coperating shifts and a day staff for reactor analysis and planning.
The organization chart 1s shown in Fig. 5.1. EFach of the four shifts will
be headed by a senior-level supervisor for the first few months. A Junior
engineer and three or four nontechnical operators, many with the previously
mentioned reactor experience, will complete the shift organization.

The Reactor Analysis Group will be compesed of four to six engineers
with broad experience in fluid-fuel reactors. Its function will be prin-
cipally to plan, supervise, and analyze the experimental progran.

Over the years this organization has developed training, operating,l
emergency,2 and maintenance practices3 that especially contribute to ex-
rerimental reactor safety. The same methods and policies will be applied

to the Molten-Salt Reactor Experiment.

 

1R. H. Guymon, "MSRE Design and Operations Report, Part VIII, Operat-
ing Procedures,' USAEC Report ORNL-TM-908, Oak Ridge National Laboratory,
to be issued.

°R. H. Guymon, "MSRE Design and Operations Report, Part IX, Safety
Procedures and Emergency Plans," USAEC Report ORNL-TM-909, Osk Ridge Na-
tional Laboratory, to be issued.

3B, (. Hise and R. Blumberg, "MSRE Design and Operations Report, Part
X, Maintenance Fquipment and Procedures,' USAEC Report ORNL-TM-910, Oak
Ridge National Laboratory, to be issued.
 

DEPARTMENT HEAD

SECRETARY
CLERK

 

 

 

 

UNCLASSIFIED
ORNL DWG. 64-5660

 

 

 

 

OPERATIONS CHIEF
ASSISTANT CHIEF

TECHNICIANS (2)

 

 

L

 

A-SHIFT SUPERVISCR
ASSISTANT SUPERVISOR
TECHNICIANS (3)

 

 

B- SHIFT SUPERVISOR
ASSISTANT SUPERVISOR
TECHNICIANS (3)

 

 

C-SHIFT SUPERVISOR
ASSISTANT SUPERVISOR
TECHNICIANS (3)

 

 

 

D-SHIFT SUPERVISOR
ASSISTANT SUPERVISOR

TECHNICIANS (3)

 

 

 

Fig. 5.1,

 

 

ANALYSIS CHIEF

 

 

 

 

 

 

MAINTENANCE AND
DESIGN CHIEF

 

 

 

-————I COMPUTER PROGRAMMER

 

 

 

NUCLEAR ENGINEERS (4)
TECHNICIAN

 

 

 

CHEMIST
TECHNICIAN

 

 

 

CHEMICAL ENGINEER
TECHNICIAN

 

 

 

 

 

METALLURGIST

 

 

 

 

 

 

 

INSTRUMENT ENGINEER

 

 

 

MAINTENANCE ENGINEER

 

 

 

DESIGN ENGINEER

 

 

 

 

 

PROCUREMENT AND
PLANNING ENGINEER

TECHNICIANS (2)

 

 

Reactor Division, Operations Department Organization for the MSRE.

LT
173

In addition to the capabilities of the Operations Department person-
nel, the Reactor Division can provide agsistance as needed from its larger
Design, Development, Analysis, Engineering Science, and Irradiation Engi-
neering Departments. Also, other ORNL research and service divisions are
participating in many aspects of the MSRE program.

The engineers and technicians assigned to the MSRE operating group
are prepared for their duties as reactor supervisors and operators by a
combination of formal classroom work and inservice training. The first
undertaking is a two-week period devoted exclusively to lectures and in-
struction on reactor physics, chemistry of salts, MSRE design, nuclear
instruments, etc.

Next, the candidates are assigned to work on circulating-salt loops,
a research reactor, or engineering developments, depending on the candi-
dates' previous experience. The next agsignment is at the MSRE to assist
in calibrations, checkout, and startup of equipment.

Following these tasks the personnel will be organized into shift
teams to begin operation with molten salt in the reactor systems. At the
end of the several months period of circulating hot salt, both the reactor
and the operators should be ready for the uranium loading and the initial
critical experiment. However, during the maintenance period (approximately
one month) prior to criticality, all the operating teams will be given a
refresher series of lectures and an examination to conclude the eight-month

training program.

5.4 Msintenance

The maintenance practices to be employed on the MSRE are the outgrowth
of several years of experience in repairing and replacing parts of the
egrlier fluid-fuel reactors, the ARE, HRE-1, and HRE-2. Basically the
system of maintenance makes use of long-handled tools manipulated by hand
through special shielding.

ATter a fault has been discovered, the reactor will be shut down,
drained of salt, and coocled. Plans for the repair will be made and
previously prepared procedures medified, 1f necessary, to cover not only

the repair details but safety precauticns also. When the system 1s cocl,
174

a hole will be cut in the top sealing membranes at a location directly
above the faulty equipment. The maintenance shield will be set in place
and manipulated to permit remcval of the shielding blocks below the mem-
brane. Next, the faulty part will be disconnected, using tools through
the special shield. If the faulty part is highly radicactive and must be
removed from the reactor cell, the operators will retire to the shielded
maintenance control room from which the building crane can be operated
remotely tc remove the part and store it in a safe place. The operators
can then return to the area above the cell and install the new part, using
the special shield. Viewing windows and remote television are available

to assist the operatcors in all these procedures.
PART 2. SAFETY ANALYSES
177

6. CONTAINMENT

6.1 General Design Considerations

 

It is required that the containment be adequate to prevent escape
of large amounts of radiocactivity to the surrcunding area during opera-
tion and maintenance and in the event of any'credible accident. The
containment must also prevent the release of.dangerous amounts of other
hazardous materials and, in general, serve to protect personnel and ex-
ternal equipment from damage.

Any equipment that contains or cculd contain multicurie amounts of
redicactive material must be surrounded by a minimum of two barriers to
prevent escape of the radiocactivity. During coperation of the MSRE, the
first barrier, the primary container, consists of the walls of the vari-
cug components in the system and the connecting piping. The reactor and
drain tank cell enclosures provide the secondary containment. These
cells are normally operated at subatmospheric pressure to assure inleak-
age.

The controlled ventilation areas in the high bay and in the varicus
cells constitute a third barrier tc the escape of activity during normal
operaticn of the reactor. Alir is drawn from the enclosed areas that are
at subatmospheric pressure, passed through absclute filters, and moni-
tored for radicactivity before discharge from the 100-ft-high ventila-
tion-system stack located south of Building 7503.

When the reactor or drain tank cells are opened for maintenance, air
flows through the openings at velocities in excess of 100 fpm and sub-
stitutes as the secondary barrier, The controlled ventilation area of
the high bay above the cell is still the third barrier. However, if the
primary piping in the cell 1s opened for maintenance, the air flow through
the cell opening becomes the primary barrier and the contrclled ventila-
tion area becomes the secondary containment.

In the hypothesis of the maximum credible accident (see sec. &.35),
in which it is assumed that hot fuel salt mixes with the water in the

cells to generate steam, the total pressure in the containment vessels
178

cculd exceed the design value of 40 psig, i1f not controlled. A vapor-
condensing system is therefore provided that can rapidly condense the
steam and also retain the ncnccondensable gases.

During normel operation, the atmosphere in the reactor and drailn
tank cells is maintained as an inert mixture (95% N, 5% O,) to eliminate
hazards from combustion of flammable materisls in the cell, such as the
cil in the lubrication system of the fuel circulating pump.

The overall leak rate of the entire primary system is expected to be

less than 1 cm”

of salt per day under normal cperating conditions. The
gystem is ¢f all-welded ccnstruction, and all flanged Jjoints are monitored
for leakage; a few Jjoints at the least vulnerable lccations have autoclave-
type fittings. Pipe lines that pass thrcough the cell walls (i.e., the
secondary containment) hzve check valves or air-operated block valves or
both which are controlled by radiation monitors or pressure switches that
can sense a rise in cell pressure. The portion of thig piping outside

the cell, the portion between the cell wall penetration and the check or
blcck valve, and the valves are enclosed to provide the required second-
ary containment. These auxiliary enclosures are designed to withstand

the same moaximum pressure (40 psig) =s the reactor and drain tank cells.

6.1.1 Resctecr Cell Design
[

 

The reacter cell, shown in Fig. 6.1, is a cylindrical carbcn steel
vessel 24 Tt in diemeter and 23 £t in overall height that has a hemi-
spherical bottom and a flzt top. The lower 24 1/2 £t was built for the
ART in 1956. It was originally designed according to Section VIIT of the
ASME Boiler and Pressure Vessel Code for 195 psig at 565°F and was tested
hydrostatically at 2300 psig.l The hemispherical bottom is 1 to 1 1/4 in.
thick. The cylindrical portion is 2 in. thick, except for the section
that contains the large penetrations, which is 4 in. thick.

This vessel was modified for the MSRE in 1962 by lengthening the

cylindrical section & 1/2 ft. Severzl new penetrations were installed,

 

1W. F. Ferguson et al., 'Termination Report for Cernstruction of the
ART Facility, USAEC Repcrt ORNL-2465, Oak Ridge Naticnal Laboratory,
Nov. 21, 1958,
 

 

SHIELDING
ANNULUS

 

 

 

 

) o

TYPICAL
FREEZE
FLANGE

TYPICAL
PIPE HEATER

'SECONDARY
CONTAINER
WALL

Fig. 6.1. Reactor Cell Model.

.

‘l w§

UNCLASSIFIED
PHOTO 38793

SHIELDING
ANNULUS

 

 

 

6LT

 

e ot sl bt

*
180

and a 12-in. section of 8-in. sched-80 pipe closed by a pipe cap was
welded into the bottom of the vessel to form a sump. The extension to
the vessel is 2 in. thick, except for the top section, which was made as
a 7 1/4- by l4-in. flange for bolting the top shield beams in place. The
flange and top shield structure are designed to withstand a pressure of
40 psig, and the completed vessel was tested hydrostatically to 48 psig,
measured at the top of the cell. Both the original vessel and the exten-
sion are made of ASTM A201, grade B or better, fire-box-quality steel.

A1l the welds on the reactor cell vessel were inspected by magnetic
particle methods, 1f they were in carbon steel, or by liquid penetrant
methods, if they were in stainless steel. All butt welds and penetration
welds were radiographed. After all the welding was completed, the vessel
was stress relieved by heating to between 1150 and 1200°F for 7 1/2 hr.

The top of the cell is constructed of two layers of 3 1/2-ft-thick
reinforced-concrete blocks, with a stainless steel membrane between. The
top layer is ordinary concrete, with a density of 150 lb/ftB, and the
bottom layer is magnetite concrete, with a density of 220 lb/ft3. The
blocks in both layers run east and west. To ald in remote maintenance,
the bottom layer is divided into three rows of blocks. Blocks in the
outer rows are supported on one end by a 13- by 4-in. channel-iron ring
welded to the inside of the cell wall. Two 36-in. I beams provide the
rest of the support fcr the bottom layer of blocks. These beams have
angle-iron and steel-plate stiffeners. The cavities are filled with con-
crete for shielding. The beams rest on a built-up support plate assembly,
which is welded to the side of the cell at the 847-ft 7-in. elevation.
Offsets 6 1/4 by 26 in. are provided in the ends of the bottom blocks to
fit over the support beams. Guides formed by angle iron assure proper
alignment. Several of the bottom blocks have stepped plugs for access
to selected parts of the cell for remote maintenance,

The sides of the blocks are recessed 1/2 in. for 14 in. down from
the top. With blocks set side by side and with a l/2-in. gap between, a
1 1/2-in. slot is formed at the top. COne-inch-thick steel plates 12 in.
high are placed in the slots for shielding. The 1ll-gage, ASTM-A240, type
304 stainless steel membrane 1s placed on top of the bottom layer of blocks
181

and 1is seal welded to the sides of the cell. Cover plates are provided
over each access plug. These are bolted to the menbrane and are sealed
by necprene O-rings. A l/8—in. layer of Masonite is placed on top of the
membrane to protect it from damage by the top layer of blocks.

The top blocks are beams that reach from one side of the cell to the
other. The ends of these blocks are bolted to the top ring of the cell
by fifty 2 1/2-in. studs, 57 1/4 in. long, made from ASTM-A320, grade L7,
bolting steel. These studs pass through holes that were formed by casting
3-in. sched-40 pipes in the ends of the blocks.

The reactor cell vessel is installed in another cylindrical steel
tank that is referred to here as the shield tank. This tank is 30 ft in
diameter by 35 1/2 ft high. The flat bottom is 3/4 in. thick, and the
cylindrical section is 3/8 in. thick. The shield tank sits on a rein-
forced concrete foundation., The reactor vessel cell is centered in the
shield tank and is supported by a 15-ft-diam, 5-ft-high cylindrical skirt
made of l-in,-thick steel plate reinforced by appropriate rings and stiff-
eners. The skirt is Joined to the hemispherical bottom of the reactor
cell in a manner that provides for some flexibility and differential ex-
pansion,

The annulus between the shield tank and the reactor cell vessel and
skirt ig filled with magnetite sand and water for shielding. The water
contains about 200 ppm of a chromate-type rust inhibitor, Nalco-360. A
4=in.-diam overflow line to the cooclant cell controls the water level in
the annulus.

The region beneath the reactor cell vessel inside the skirt contains
only water, and steam will be produced there if a large quantity of salt
is spilled into the bottom of the reactor cell (see sec. 8.6). An 8-in.-
diam vent pipe, which has a capacity for three times the estimated steam
production rate, extends from the water-filled skirt area under the vessel
to the coolant cell, where the steam would be vented.

Numerous large sleeves are required through the walls of the reactor
cell and shield tank to provide for process and service piping, electrical
and instrument leads, and for other access. The 4- to 36-in.-diam pipe

gsleeves are welded into the walls of the reactor cell and the shield tank.
182

Since the temperature of the reactor cell will be near 15C°F when the
reactor is operating and the temperature of the shield tank may at times
be as low as 60°F, bellcows were incorporated in most of the sleeves to
permit radial and axial movement of one tank relative to the other with-
out producing excessive stress. The bellows are covered to prevent the
sand in the shield tank from packing tightly arcund them.

Several other lines are installed directly in the penetrations, with
welded seals at cne or both ends, or they are grouped in plugs filled
with concrete and inserted in the penetrations. The major openings are
the 36-in.-diam neutrcn instrument tube and drain tank interconnection
and the 30-in.-diam duct for ventilating the cell when maintenance is in
progress. The coriginal tank contained several other 8- and 24-in.-diam
sleeves, and they were either removed or closed and filled with shielding.
The smaller penetrations and methods of sealing them are described in

Sec. 6.1.3.

¢.1.2 Drain Tank Cell Design

 

The rectangular drain tank cell, shewn in Fig. 6.2, is 17 It 7 in.
by 21 ft 2 1/2 in., with the corners beveled at 45° angles for 2 1/2 ft.
The flat floor is at the 8l4-ft elevation. The open pit extends to the
852-t elevation.

The cell was designed to withstand a pressure of 40 psig, and when
completed in 1962 it was hydrostatically tested at 48 psig measured at
the elevation of the menbrane at 838 1/2 ft. The bottom and sides have
a 3/16—in.-thick stainless steel liner backed up by heavily reinforced
concrete. Magnetite concrete is used where reguired for biological
shielding.

Vertical cclumns in the north and south walls are welded to hori-
zontal beams embedded in the concrete of the cell floor. A 4-in. slot
is provided at the top of each horizontal beam so that a steel key may
be wedged inte it to hold down the top blocks.

The top blocks are arranged in two layers. Both layers of blocks
are ordinary concrete (density 150 1b/ft>). The bottom layer is 4 £t
thick and the top layer is 3 1/2 ft thick. An ll-gage type 304 stainless
|
]

 

 

 

 

s

 

e

w

"

»

 

183

 

SECONDARY
CONTAINMENT

TYPICAL
FREEZE §
VALVE

FUEL SALT
DRAIN TANK

NO. 1

 

ig. 6.2. Drain Tank Cell

UNCLASSIFIED
PHOTO 38768

* TYPICAL SUPPORT
.AND WEIGHT CELL

ARRANGEMENT

FLUSH SALT
DRAIN TANK

 

 

 

e b At

Lt s e

 

 
184

steel membrane is placed between the twc layers and seal welded to the

cell liner at elevation 838 1/2 ft.

6.1.3 Penetrations and Methods of Sealing

 

Piping and wiring penetrations through the cell walls were designed
to eliminate possible sources of gas leakage.

All electrical leads passing through the cell walls are in magnesia-
filled copper sheaths (Fig. 6.3). The sheaths were sealed to the 3/4-in.
pipe penetrations by two compression-type fittings, one inside and one
outside the cell. The ends of the sheaths that terminate inside the cells
were sealed at the disconnect by glass-to-metal welds. The ends that ter-
minate outside were sealed by standard mineral-insulated cable-end seals,
such as those manufactured by the General Cable Company. (The seal was
formed by compressing a plastic insulating material arocund the wires. )

All thermocouples have Fiberglas-insulated leads in multiconductor
sheathed cables. The sheaths were sealed tc the 3/4—in. pipe penetra-
tions inside and outside the cell by using soft solder. The ends of the
sheaths terminating inside the cells were sealed at the disconnect with
glass-to-metal welds. The ends cof the cables ocutside the cells terminate
in epoxy-sealed headers. The headers can be pressurized to test for
leaks.

A1l instrument pneumatic signal lines and instrument air lines are
sealed to the 3/4~in. pipe penetrations by two compressiocon type fittings,
one inside and one outside the cell. Fach of these lines contains a
block valve located near the cell wall; the valves close automatically
if the cell pressure becomes greater than atmospheric.

Methods of sealing certain lines require special mention, as follows:

1. The 30-in.-diam cell ventilation duct contains two 30-in. motor-
operated butterfly valves in series. These valves are under strict ad-
ministrative contrcl to assure that they remain closed during reactor
operation.

2. The component cooling system blowers are sealed in containment

tanks to guard against loss of gas at the shaft seals.
7 i

 

    

|
}
|
|

THREADED PIPE ~
COUPLING ™ |
COMPRESSION GLAND, ™

CCMPRESSION NUT A ‘

!

SLEEVE~

Fig. 6.3.

QUTER VES3SEL WALL

rREACTOR

N CONTAINMENT
VESSE L
WA L L

o SAKD-WATER ANNULUSD e

- T EXPANSION
S BELLOWS

FENETRATION PLUG SLEEVE —_

.L”
T 25 MIiN OFFSET

% NPs (TYPicAaL)
M\ CABLE SLEEVE-—

CONCRETE ¥
BETWEEN CABLE SLEEVES

Typical Electric Lead Penetration of Reactor Cell Wall.

\ .
2oz -

 

FPENETRATION
SLEEVE

FITTINGS TYPICAL
[ BECTH ENDS

MULTI-CONDUCTOR
MINERAL- INSULATELD
COPPER-SEHEATHED
ELECTRIEC CABLF

G8T
186

3. The cell evacuation line contains a block valve, HCV565, which
automatically closes in the event radiocactivity is detected in the line
by the radiation monitor.

4. The air supply lines for the cell sumps contain soft—seatgd
check wvalves.

5. Jet discharge lines from the sumps each contain two block valves
in series that automatically close 1f the cell pressure becomes greater
than atmospheric. A 1/2-in. connection is provided between the valves to
test them fcor leak tightness.

6. The fuel sampling and enriching system is interlocked to prevent
a direct opening to the atmosphere.

7. The steam-condensing system used in conjunction with the drain
tank heat-removal system is a closed loop, except for the water supply
lines, which contain soft-seated check valves, and the vent, which relieves
to the cell vapor-condensing system (see sec. 6.2).

8, All cooling water lines entering the cell have soft-seated check
valves or blcck valves contrelled by radiation monitors. All lines leaving
the cells are provided with blcek valves controlled by radiation monitors.

9. The lubricating-oil system for the fuel circulating pump 1is a
closed circulating lccp. Strict supervision is provided during additions
of 0il or oil sampling tc assure that the containment is not violated.

10. The leak-detector system tubing operates at a higher pressure than
the reactor process systems, sc no cutleakage is anticipated.

11. Several differential pressure cells and pressure transmitters
are lccated cutside the cells but are connected to process piping inside
through instrument tubing. The instrument lines are doubly contained.
The case of the differential pressure cell provides primary containment.
The instrument cases are located inside a chamber designed for an internal
pressure of 50 psig to provide the secondary containment.

12. A1l helium supply lines connected to process equipment inside
the cells contain one or more scoft-seated check valves.

13. Offgas lines that carry gaseocus fission products toc the charcoal
beds are doubly contained for their entire length. The offgas lines from

the charcoal beds have a common block valve, which closes on detection of
187

radiocactivity in the line. (For details, see Dwg. D-AA-A-40883, Appendix
B. )

The gtresses which exist at various penetrations in the containment
vessel wall were studied and were found to be within allowable values,

with the maximum stresses occurring in the nozzles, @

6.1.4 Leak Testing

After construction of the reactor and drain tank cells was completed,
leak tests were performed after the cells had been hydrostatically tested
to 48 psig. The cells were held at 20 psig for 20 hr, and a leakage rate
of O.25%/day was measured. When the reactor is operating, the cells will
be kept at 12.7 psia, and the rate of inleakage will be monitored continu-
ously. After each maintenance period, the cell leakage rate will be mea-
sured and determined to be less than l%/day (assumed in dispersion calcu-
lations, sec. 8.7.2) before the reactor is started.

The meny service penetrations carrying air, water, etc., into the
cells are equipped with various closing devices, as described above. The
cell leakage test does not, of course, test these devices. It is intended
that such closures be tested at intervals not greater than 6 months as

verification of their integrity.

6.2 Vapor-Condensing System

 

The MSRE vapor-condensing system 1s similar in principle to those
of the SM-1A plant at Fort Greeley, Alaska, and the Humbolt Bay power
plant. In the event of the maximum credible accident, vapcor will be
discharged from the reactor cell into tanks, where the steam will be
condensed in water and the noncondensable gases will be retained. The
MSRE system differs from the SM-1A and Humbolt Bay systems in important
details, however, because the energy release rates are low and the system
is adapted tc an existing containment wvessel. Design of the vapor-con-
densing system is based on the following:3s %

1. The volumes of the reactor and drain tank cells are, respec-
tively, 11,300 and 6,700 ft2. These cells will operate at —2 psig and
150°F, Calculations have shown that only 7000 ft2 of noncondensable
188

gas will be discharged to a vapor-condensing system during the maximum
credible accident. The free volume of the system was therefore specified
as 4500 ft3, which is enough to contain 10,000 ft? of gas from the reactor
cell at 30 psig and 140°T or 13,500 £t3 at 40 psig and 140°F.

2. The vapor-condensing system will contain 1200 ft2 of water at
70°F or less. The full 5 x 10 Btu calculated to be released by the salt
can be abscrbed in the water without exceeding 140°F.

3. By the time the reactor cell pressure can reach 40 psig, the
steam generation rate will be 16 lb/sec or less. The rate will decrease
as the rate cf release of salt into the cell decreases. The relief line
from the reactor cell to the vapor-condensing tank was specified to pass
16 lb/sec of steam at 40 psig with a pregsure drop of 10 psi or less.

The discharge end of the line is tc be located 6 ft below the surface of
the water in the tank to ensure complete condensation of the steam.

. t 1s desirable to keep the vapor-condensing system isolated
from the reactor cell and to limit its use to accidents in which the
pressure rises above 15 psig. Bursting disks are to be installed in the
relief line tc the vapor-condensing tank. The disks will burst at about
20-psi pressure difference. Rupture of the disk in the second (smaller)
line will not affect the pressure at which the other disk breaks.?

5. Vacuum-relief wvalves will be installed in the reactor cell relief
line at the vapcr-condensing tank, This will permit vapors to return to
the reactor cell and prevent the pressure in that cell from going below
-5 psig when the steam condenses at the end of the accident.

6. The vapcr-condensing system may contain as much as 400,000 curies
of gaseous fisgion precducts and daughter prcducts 5 min after the beginning

of a maximum credible accident and 40,000 curies after 1 hr. In addition,

 

°L. F. Parsly, "MSRE Containment Vessel Stress Studies,” ORNL inter-
nal document MSR-62-15, Feb. 2, 1962.

*R. B. Briggs, "MSRE Pressure-Suppression System,” ORNL internal
document MSR-£1-135, Nov. 15, 19&1.

“L. F. Parsly, "Design of Pressure-Suppressicn System for the MSRE,”
ORNL internal correspondence tc R. B. Briggs, Oct. 17, 1961.

SLetter from T. E. Northup to R. G. Affel, May 28, 1964, Subject:
MSRE Vapor Condensing System — Parallel Rupture Discs.
189

the water will contain some radiocactive golids that will be transferred

as entrainment in the vapor. The radiation level at 10 meters from an
unshielded tank would be about 2000 rad/hr at 5 min and 200 rad/hr at 1 hr;
however, the tanks will be shielded sufficiently to reduce the initial
radiation level to below 00 rad/hr. Provision has been made for the con-
trolled discharge of the gases to the atmosphere through the filter system
and stack.

The vapor-condensing system will be installed about 40 £t east of
Building 7503, as shown in Fig. 6.4. It will consist of two tanks, the
vapor-condensing tank and the gas-retention tank. The vapor-condensing
tank, sometimes referred to as the water tank, i1s a vertical cylinder
that is normally maintained about two-thirds full of water through which
gases forced from the reactor cell in a major accident would be bubbled
to condense the steam. Tre tank containg about 1200 ft° of water plus a
corrosion inhibitor. The estimated maximum of 5 X 10° Btu of heat that
could be released from the fuel salt in the reactor and drain tank cells
would raise the water temperature to about 140°F,3 The noncondensable
gases are vented to a large gas storage tank.

The vertical vapor-ccndensing tank is 10 ft in outside diameter and
23 ft 4 in. high, including the flanged and dished, l/2-in.—thick, top
and bottom heads. The shell is 3/8 in. thick and constructed of SA-300
class I, A-201, grade B firebox steel. There are stiffening rings, 1 in.
thick and 3 in. high, located on the exterior abcocut 2 ft 6 in. apart.

The 12-in. gas inlet pipe in the top head extends 13 ft 8 in. intc the
tank to about 6 ft below the normal water level. The tank is installed
with the top about € ft below the normal grade level of 850 ft and the
bottom at an elevation of about 819 £t. Additional earth is mounded
above the tank to provide a total of 11 to 13 ft of shielding.

The gas inlet line in the interior of the tank has a pipe cross
(12 x 12 X 12 X 12 in.) about 3 ft above the water level to which is
connected two 1l2-in. cast-steel-body check valves. These check valves
close when the gas flow 1s into the tank but open and act as vacuum-relief
valves to prevent water from being drawn in reverse flow through the inlet

line during cooldown of the cell after the accident.
SPECIAL EQUIPMENT ROOM
IN REACTOR BLDG.

FLOOR ELEV 852 ft

40 ft FROM REACTOR BLDG.

UNCLASSIFIED
ORNL-LR-DWG. 87162R2

 

 

 

 

 

 

 

ELEV 848 ft

 

 

 

BURSTING DISK

4-in. LINE1

 

 

ELEV 836 ft

 

12-in-DiAM. )
RELIEF LINE

t————3  +— ~ {00 ft ——————

 

 

XX
N

    

30-in.-DIAM. VENT DUCT
FROM REACTOR CELL

Fig. 6.4.

«f TO STACK

VAPOR
CONDENSING
1C-ft DIAM. x 23-ft HIGH TANK

\)
{800 #3 - 1200 1> WATE R

BUTTERFLY P o
VALVES ‘

 

 

Vapor-Condensing System.

 

  
 

ELEV. 858 ft
2-in-DIAM. VENT LINE X o —
TO FILTERS AND STACK~y

GROUND ELEV. FILL GAS
851 ft- 6 in. RE11'_E\I\|I\|T‘<ION
~ o

e
10-ft-DIAM. x 66 ft-LONG
3900-t3

N _VACUUM

RELIEF VALVE

 

ELEV. 830 ft.

 

 

   

ELEY. 8249 ft

06T
191

The noncondensable gases leave the top of the vapor-condensing tank
through the 12-in. outlet pipe and flow through the expansion joint in
the line to a side nozzle on the gag-retention tank. The gas-retention
tank is 10 ft in outside diameter snd 66 ft 3 1/2 in. long, including
the two 1/2-in.-thick flanged and dished head. The shell is 3/8 in,
thick and is reinforced with l-in.-thick, 3-in.-high rings located about
4 £t 7 in. apart. The tank is fabricated of SA-300 class I, SA-201, grade
B, firebox steel. Gases can accumulate in this tank, initially at atmos-
pheric pressure, but should not exceed a pressure of 20 psig. A normally
closed hand valve is provided to make possible the venting of gas to the
absolute filters and the stack.

The entire vapor concensation system is designed to meet Section VIII

of the ASME Boiler and Pressure Vessel Code.

6.3 Containment Ventilation System

 

The third containment barrier, controlled ventilation, is provided
for all areas that surround the secondary containment barrier. These
areas include the high-bay area above the cells, the auxiliary egquipment
rooms on the southeast side of the building, the coolant equipment room
on the south side, and the instrument room on the north side of the reactor
cell (see Fig. 3.1). The high-bay area is lined with steel sheets sealed
at the joints to limit the leakage to approximately 1000 c¢fm at a pressure
of 0.3 in. H»0. The other rooms have concrete or steel-lined walls, and
each is designed for low leakage. The doors are gasketed; all service
penetrations are sealed; and internal heating and ccoling are provided.

The general pattern of air flow is from the less hazardous to the
potentially more hazardous zones. EFach zone 1is kept at a negative pres-
sure by the ventilating fan and filter system described below. This
guarantees that, when the fans are running, activity leaks will be dis-
charged to the atmosphere after filtration. It also keeps outleakage to
a minimum if the fans are off; the outleakage is first intc the standard
building and then to the outdoors.

The driving force for the ventilation system is located outside the

reactor building and consists of two 20,000-cfm fans arranged in parallel
192

so that in case of failure ¢f one fan the other fan starts autcmatically
to provide uninterrupted ventilation. Both fans pull thrcough a bank of
CWS filters that are capable of removing particles down to at least 1
micron in size. In a fission-product spill the noble gases are the prin-
cipal activities which escape from the filters. This type of filter sys-
tem is employed at other reactor installaticns at the Oak Ridge National
Izboratory. The fans discharge the filtered air to a 100-ft-high stack.

The ventilaticn system is designed for several purposes:

1. During normal cperaticn when the reactor and drain tank cells
are sealed off, the ventilation system exhausts air from the building at
2 rate of 18,000 cfm and maintains a slight negative pressure (~0.1 in.
H,0) on the limited leakage zones outside the containment vessel.

2. During an accident in which fission gases leak from the contain-
ment vessel intc the high bay, the building inlet supply is stopped and
the ventilation fans exhaust only the building inleakage, estimated to be
about 1000 cfm at C.3 in. HyO.

3. During maintenance when the containment cells are opened, the
ventilation system maintains an air flow of approximately 100 ft/min
through the opening (in the centainment cells) through which maintenance
work is perfcrmed. This 100-fpm flow of air is intended to prevent the
transfer of fission-product gases and radiocactive particles from inside
the containment vessel ocut into the working area of the building.

%. During norrmal cperation the ventilation system serves to dilute
the small gquantity of fissicn-prcoduct gz=ses which are removed from the
reactor charcoal beds. Since the stack filow is 20,000 cfm, and only a
few cubic centimeters of fissicn gas are discarded each day, the dilution
factor is approximately 10%1. The stack, as it discharges into the at-
mosphere, provides ancther factcr of 100 to 1000 dilution, depending on
the weather conditicns at the time.

The containment ventilation system utilizes either of two 20,000-cfm
(nominal capacity) centrifugal fans located at the base of the 100-ft-high
discharge stack to induce air flow through the various containment areas
in and adjacent tc Building 7503. The estimated rates of air removal from

these areas during normal reactor operation are shown in Teble 6.1.
193

Table 6.1. Estimated Containment Ventilation System
Air Flow Rates During Normal Reactor Operation

 

Alr Flow Rate

 

(cfm)
Through supply air filter house 14, 000-17, 000
Through change room 750-1, 000
Leaving high-bay at sampler enricher 12,000-15,000
Through liquid waste cell 200—400
Teaving liquid waste tank 0-100
Through remote-maintenance pump cell 200-1, 500
Through decontamination cell 200~1, 500
Through equipment storage cell 200—400
Through fuel-processing cell 200—4C0
Through spare cell 200—400
Leaving electrical service areasg® 400~600
Teaving reactor and drain tank cellsP O
Leaving coolant cell 800-1, 200
Leaving special equipment room 200~400
Through vent house 200—400
Leaving charcoal beds cell 0—-100
Leaving service tunnel 400~600

 

a . . . i

Electrical service area includes transmitter
room, north service area, and south service area
(which includes west tummel).

b . . . .
During maintenance periods the air leaves the
high bay through the open cell at a rate of 12,000
to 15,000 cfm.

The bulk of the air, 14,000 to 17,000 cfm, enters the main building
at the northwest corner through an intake air filter house. In saddition
to dust filters, this house contains steam heated coils for warming the
intake air during the winter months. A bypass damper in the house wall
assures an air supply even though the filters become excessively clogged.
Another counter-balanced bypass damper prevents the negative pressure in
the high-bay area from becoming low enough to collapse the sheet metal
liner.

Alr from the intake filter house enters the high-bay area at the
northwest corner of the bay through an opening about 22 ft above the

operating floor level, About 1000 cfm of air is also drawn into the
194

high-bay area through the change (or locker) room located at the same
level.

Exhaust ducts leading to the main intake ducts for the stack fans
withdraw air from each of the six small cells located beneath the oper-
ating floor level* and cause alr to be drawn from the high-bay area down
through the openings between the concrete roof blocks covering these
cells. The ventilation rates for the cells vary from 200 to 1500 cfm.
Another exhaust duct pulls 400 to 600 cfm of air from the electric ser-
vice area transmitter room, west tunnel, etc.; the air reaches these
rooms from the high-bay area through various stairways, passages, etc.
The bulk of the air from the high-bay area, 12,000 to 15,000 cfm, is
normally exhausted at the cperating floor level through two openings
that flank the sampler-enricher station area.

Air is drawn from the coclant cell by the same induced circulating
system, and it is exhausted from the cell through an opening in the south
wall at about the 861-ft elevaticn. Air also enters the cooclant cell
through general leakage, etc., frcm cutside the building and 1s only
partially drawn from the high-bay arez.

During maintenance operaticns in the reactor or drain tank cells,
air can be exhausted from the cells to cause downflow of air through
the cpenings where concrete rcof plugs have been removed. This positive
dcwvnward movement of 100 fpm, or more, aids in preventing escape cof air-
borne contaminants into cccupied building areas. To accomplish this
circulaticn, air is withdrawn, =2t up to 15,000 c¢fm flow rates, through a
30-in.-diam opening at the bottom of the scuthwestern sector of the con-
tainment vessel. The a‘r then flows thrcugh a 30-in.-diam carbon steel
duct across the cecclant drain cell and thrcugh the special equipment
room to a valve pit containing two 30-in.-diam, motor-operated butterfly
valves in seriesg, The valves are normally closed during reactor operation
and are opened only when roof plugs are removed for maintenance and cell
ventilation is required. (A 12-in., sched-40 pipe takes off from the

30-in.-diam duct just upstream of the butterfly valves and leads tc the

 

¥Liquid waste cell, remote-maintenance pump cell, decontamination
cell, equipment storage cell, fuel-processing cell, and the spare cell.
195

vapor-condensing system described in Section 6.2.) The 30-in.-diam duct
continues underground to the vicinity of the filter pit, where it turns
upward to join the main 36-in.-diam duct leading to the pit.

The two motor-driven stack fans and the 100-ft steel stack for dis-
charge of the air are located about 110 ft south of Building 7503. (0off-
gas from the charcoal adsorber beds is alsc vented through the filter pit
and stack,)

The containment ventilation system is provided with numerous manuaslly
positioned dampers to adjust the air flow. Differential pressure cells

indicate direction of air flow, filter resistances, etec.
196

7. DAMAGE TO PRIMARY CONTAINMENT SYSTEM

7.1 Nuclear Incidents

 

There are several conceivable incidents in which nuclear heating
could produce undesirably high temperatures and pressures inside the pri-
mary system. Some of these incidents are such as might occur in any power
reactor, while others are peculiar to the MSRE. In the former category
are uncontrolled rod withdrawal* and 'cold-coolant" accidents that cause
excursions by reactivity increases larger or faster than those normally
encountered. Also conceivable are '"loss-of-flow"” and "loss-of-coolant”
accidents in which the temperature increases because the heat removal is
drastically reduced. Conceivable incidents associated with particular
features of the MSRE involve either premature criticality while the core
is being filled with fuel salt or increases in the amount of uranium in
the core during operation. Fach of these could occur in more than one

way.

7.1.1 General Considerations in Reactivity Incidents

The severity of the power, temperature, and pressure transients as-
sociated with a reactivity incident depends upon the amount of excess re-
activity involved, the rate at which it is added, the initial power level,
the effectiveness of the inherent shutdown mechanisms, and the efficacy
of the reactor safety system. All these factors depend to some extent on
the fuel composition, because, as shown in Table 7.1, the composition de-
termines the magnitude of the various reactivity coefficients, the prompt
neutron lifetime, and the control rod worth.

In general, equivalent physical situations lead to larger amounts of
reactivity and greater rates of addition with fuel B than with either A

or C. This is a consequence of the larger values of the reactivity

 

*This accident is commonly called "the startup accident” because it
could occur while the reactor is being taken to criticality. This termi-
nology is not used in the MSRE analysis to avoid confusion with the filling
accident, which could occur at an earlier stage of a plant startup.
197

Table 7.1. Nuclear Characteristics of MSRE with Various Fuels

Temperature: 1200°F
Graphite density: 1.86 g/cm’

 

 

Fuel A Fuel B Fuel C
Uranium concentration, mole %
Clean, noncirculating
235 0.29 0.18 0.29
Total U 0.31 0.19 0.83
Operating®
U2’ 0.3 0.20 0.35
Total U 0.36 0.21 0.89
Uranium inventory,b kg
Initial criticality
U233 79 48 77
Total U 85 52 220
Operating®
Ue35 91 55 92
Total U 98 59 230
Reactivity coefficientsC
Fuel temperature, °F-1 ~3.0 X 107° =5,0 x 10~% =3.3 x 10~?
Graphite temperature, °F~1 =3.4 X 1077 .9 x 107% =3.7 x 1077
Uranium concentration 0.25 0. 30 0.18d
0.21F
Fuel-salt density 0.19 (.35 0.18
Graphite density 0.76 Q.73 Q.77
Prompt-neutron lifetime, sec 2.3 x 1074 3.5 X 107% 2.4 % 1074
Control rod worth, % ok/k 5.6 7.6 5.7

 

aFuel loaded to compensate for 4% 5k/k in poisons.
bBased on 73 ft3 of fuel salt at 1200°F.

at initial critical concentration. Where units are shown, coef-
ficients for variable x are of the form: (1/k)(dk/dx). Other coeffi-
cients are of form: (x/k)(dk/dx).

dBased on uranium isotopic composition of clean critical reactor.

®Based on highly enriched uranium (93% U239).
198

coefficients and control rod worth, the absence of the poisoning effect
of Th?3? or U238, and the lower inventcory of U227 in the core.

The power and temperature transients assocliated with a given reac-
tivity incident increase in severity as the initial power level is reduced.
The reason for this is that, when the reactor becomes critical at very
low power, the power must increase through several orders of magnitude
before the reactivity feedback from increasing system temperatures becomes
effective. Thus, even slow reactivity additions can introduce substantial
excess reactivity if the reactor power is very low when kgpp = 1.

The power level when the reactor is just critical depends on the
strength of the neutron source, the shutdown margin prior to the approach
to criticality, and the rate at which reactivity is added to take the re-
actor through the critical point. The minimum neutron source strength to
be considered is 4 X 10° neutrons/sec, which is the rate of neutron pro-
duction in the core by (G,n) reactions in the fuel salt. (Ordinarily the
effective source will be much stronger, because an external antimony-
beryllium source will normally supply about 107 neutrons/sec to the core
and, after the reactor has operated at high power, fission~-product gamma
rays will generate up to 1019 photoneutrons per second in the core.) The
physical design of the MSRE limits the meximum credible rate of reactivity
addition to about 0.1% BK/k per second. The ratio of the nuclear power
at criticality to the source strength varies only *10% for reactivity ad-
dition rates between 0.05 and 0.1% 8k/k per second and the maximum shut-
down margins attainable in the MSEE., For these conditions, the power
level at criticality is about 2 Mw if only the inherent (,n) source is
present; it is proportionately higher with stronger sources. The power
level at criticality 1s also higher for lower rates of reacctivity addi-
tion.

The principal factor in the inherent shutdown mechanism for the MSRE
is the negative temperature coefficient of reactivity of the fuel salt.
Since most of the fission heat is produced directly in the fuel, there is
no delay between a power excursion and the action of this coefficient.

The graphite moderator also has a negative temperature coefficient of re-

activity, but this temperature rises slowly during a rapid power transient
199

because only a small fraction of the energy of fission is absorbed in the
graphite. As a result, the action of the graphite temperature coefficient
is delayed by the time required for heat transfer from the fuel to the
graphite. ©Since fuel salt B has the largest negative temperature coeffi-
cient of reactivity, a given change in reactivity produces smaller excur-
sions with this fuel sslt than with either fuel salt A or C.

The MSRE safety system causes the three control rods to drop by grav-
ity when the nuclear power reaches 15 Mw or when the reactor cutlet tem-
perature reaches 1300°F., In the analysis of reactivity incidents, conser-
vative values were assumed for delay time and rod acceleration, namely,
0.1 sec delay time and 5 ft/sec® acceleration. (As stated in Section 2,
tests of the rod system showed that the acceleration was about 12 ft/sec?.)
It was also assumed that one of the three rods failed to drop.

In the sections which follow it is shown that for all credible reac-
tivity incidents, the combination of the inherent neutron source, the in-
herent kinetics of the reactor, and the action of the safety system lead

to tolerable transients in temperature and pressure.

7.1.2 Uncontrolled Rod Withdrawal

 

It can be postulated that through some combination of misoperation
or control system malfunction or both, all the control rods are withdrawn
simultaneously, beginning with the reactor subcritical, so that consider-
able excess reactivity is introduced before inherent or external shutdown
mechanisms begin to take effect. This accident is most severe when criti-
cality is achieved with all three control rods moving in unison at the
position of maximum differential worth. Since this condition is within
the range of combinations of shutdown margin and rod worth for all three
fuels, it was used as a basis for analyzing this accident. The maximum
rates of reactivity addition by control-rod withdrawal are 0.08, 0.10,
and 0.08% 5k/k per sec when the system contains fuels A, B, and C, re-
spectively. The initial transients associated with these ramps were cal-
culated for fuels B and C, starting with the reactor just critical at
0.002 watt and 1200°F. (The transients for fuel A would be practically
the same as for fuel C.)
200

The first 15 sec of the transients in some of the variables are
shown in Figs. 7.1 and 7.2 for fuels B and C, respectively. The curves
illustrate the behavior of the power, the nuclear average temperatures of
the fuel and graphite, T% and Tg, the temperature of the fuel leaving the
hottest channel, (To)MaX’ and the highest fuel temperature in the core,
(T ) Max.

Although the rate of reactivity addition is smaller for fuel C than

T

for fuel B, the excursions are more severe for fuel C because of the
smaller fuel temperature coefficient of reactivity and the shorter prompt
neutron lifetime associated with this mixture. The power excursion oc-
curred somewhat later with fuel C because of the greater time required
to reach prompt criticality at the lower ramp rate, Peak core pressure
rises were 18 and 21 psi for fuels B and C, respectively.

During steady-state operation the maximum fuel temperature occurs
at the outlet of the hottest channel. However, during severe power ex-
cursions that are short compared with the transit time of fuel through
the core, the maximum fuel temperature at a given time may be at a lower
elevation in the hottest channel where the power density is relatively
higher. This is illustrated by the difference between the maximum fuel
temperature and the hot channel outlet temperature during and immediately
after the initial power excursion. These two temperatures converge as
the fuel is swept from the region of maximum power density toward the
core outlet while the power 1s relatively steady. The rise in the mixed-
mean temperature of the fuel leaving the core is about one-half that
shown for the hottest fuel channel.

The transient calculations were stopped before the fuel that was
affected by the initial power excursion had traversed the external loop
and returned to the core. The trends shown in Figs. 7.1 and 7.2 would
continue until the core inlet temperature began to rise, about 16 sec
after the initial excursion in the outlet temperature. At that time, the
power and the outlet temperatures would begin to decrease; the nuclear
average temperatures would continue to rise as long as rod withdrawal was
continued. However, the rise in graphite temperature resulting from heat
transfer from the fuel would reduce the rate of rise of the fuel nuclear

average temperature.
201

Unclasgified
CRNL-DWG 64-601

Temperature (°F)

 

400 :

300

Power (Mw)

 

6 8 10 12 14 16
Time (sec)

Fig. 7.1. Power and Tempersture Transients Produced by Uncontrolled
Rod Withdrawal in Reactor Operating with Fuel B.
202

Unclassified -
ORNL-DWG 64-602

1700

v (TO )max

[
o
o
Q

Temperature (°F)
}_J
n
O
S

1400 pEELS
1300

12C0

400
300 Eigs

200

Power (Mw)

100

 

6 g 10 12 14 16
Time ({sec)

Fig. 7.2. Power and Temperature Transients Produced by Uncontrolled
Rod Withdrawal in Reactor Operating with Fuel C. ~
203

It 1s clear from Figs. 7.1 and 7.2 that intolerably high fuel tem-
peratures would be reached in this accident if complete withdrawal of the
control rods were possible. This 1s prevented by the safety system, which
disengages the clutches on all three rod drives when the power reaches
15 Mw and allows the rods to fall into the core.

The adequacy of the high-power safety action was examined by calcu-
lations for fuel C. The transients described in Fig. 7.2 were allowed
to develop until the power reached 15 Mw., The period at this time was
98 msec. At that point it was assumed that two rods were dropped (be-
ginning 0.1 sec after the flux reached 15 Mw, with an acceleration of
only 5 ft/secz), while the third rod continued to withdraw. The power
and temperature excursions for this case are shown in Fig. 7.3. The pres-
sure excursion amounted to orly & psi. It is evident that dropping two
control rods in this accident limited the initial excursions to toler-
able proportions. The reactor cannot be made critical by the withdrawal
of only one control rod. Therefore the safety action is adequate for
the postulated accident, and failure of the rod-drop mechanism on one rod

does not impair the safety.

7.1.3 '"Cold-Slug" Accident

 

The "cold-slug' accident is one in which the mean temperature of the
core decreases rapidly because of the injection of fuel at an abnormally
low temperature. The reactivity increases in this case because of the
fuel's negative temperature ccefficient of reactivity.

The design of the MSRE fuel system, with the core at the lowest
point of the circulating loop, is conducive to thermal convection, so it
is unlikely that the temperature of the fuel outside the core can be re-
duced much below that in the core., Furthermore, if the fuel pump stops,
control system interlocks automatically prevent the pump from being re-
started unless the control rods are inserted. The shutdown margin of the
rods is enough so that the reactor cannot be made critical by any cold
slug. (With the rods in, and the graphite at 1200°F, the critical tem-
perature of the fuel is below the freezing point.)

Although the occurrence of a cold slug accident is highly improbable,

the consequences of such an accident were calculated. In the calculations
204

UNCLASSIFIED
ORNL DWG. 63.8178

TEMPERATURE (°F)

POWER {(Mw )

 

6 8 10 12 1L 16

TIME (sec)

Fig. 7.3. Effect of Dropping Two Control Rods at 15 Mw During Un-
controlled Rod Withdrawal in Reactor Operating with Fuel C.
205

i1t was assumed that one core volume of fuel galt at 900°F entered the core
at 1200 gpm with the core operating at 1200°F and 1 kw. No corrective
action was assumed. Calculations for fuels B and C gave pressure excur-
sions of about 6 psi in either case. Power and temperature excursions
were larger for fuel B and are shown in Fig. 7.4. The progress of the
transients showed that as the front of the cooler salt entered the core,
the inlet temperature dropped to 900°F., As the volume of the core oc-
cupied by the cold slug became larger, the fuel's nuclear average tem-
perature decreased slowly and added reactivity. When the reactivity ap-
proached prompt critical, the power rose sharply (on a 110-msec period)
and substantial nuclear heating occurred. This caused the fuel's nuclear
average temperature to rise sharply and limit the power excursion. The
outlet temperature rose at first, reflecting the increased heat genera-
tion, and then, as the leading edge of the cold slug reached the top of
the core, it dropped sharply. Simultaneocusly, the reactor inlet salt tem-
perature returned to 1200°F as the available amount of cold fluid was ex-
hausted. The channel outlet temperature passed through & maximum upon
arrival of the fluid heated at the center of the core by the initial power
transient and then decreased until, finally, the rise in the salt's inlet
temperature was again reflected (about 9.4 sec later) as a rise in the

channel outlet temperature.

7.1.4 Tilling Accidents

 

In any shutdown during which the reactor must be cooled, the fuel is
drained from the core. Refilling the core with fuel salt is therefore =
routine operation. It 1s conceivable that criticality could be attained
before the core was full and that damaging temperatures might be produced
in the partially filled core. This could cccur in several ways if no
safety system were available for protection,

In a normal startup, the reactor and the fuel salt in the drain tank
are heated to near the operating temperature before the salt is transferred
by gas pressure into the core. The control rods are positioned so that
the reactor remains subcritical after the salt fills the core. They are

not fully inserted, however, so that they may be dropped if criticality
206

UNCLASSIFIED
ORNL DWG. 63-8179

.

-
£

(do) HUNLVEAIAT

 

(M) MEMOd

25.0

15.0

10.0

TIME (sec)

Power and Temperature Transients Following Injection of

Fig. 7.4.
Fuel B at S00°F

to Core at 1200°F; No Corrective Action Taken.

in
207

is unexpectedly attained during s fill. Criticality could be reached pre-
maturely while the core is being filled if: (1) the control rods were
withdrawn too far, (2) the core temperature were abnormally low, or (3) the
Tuel were abnormally concentrated in uranium. Interlocks and procedures
are designed for effective prevention of premsture criticality for any of
these reasons. The consequences were examined, however, to determine the
adequacy of the safety system for preventing damage should such an acci-
dent occur.

The amount of reactivity available in a filling accident and the rate
of addition depend on the conditions causing the accident, the character-
istics of the fuel salt, and the rate of filling. The filling rate is
physically limited by the gas-addition system, which contains a rupture
disk to limit the supply pressure and a permanently installed restrictor
that controls the gas flow rate into the drain tank. In the case of'fill-
ing the reactor with the control rods fully withdrawn, the excess reactiv-
ity is limited to the amount in the narmal fuel loading. Only about 3%
6k/k will be required for normal operation (see sec. 1.3), and the amount
of uranium in the fuel will be restricted by administrative control to
provide no more than required. Filling at the normal rate (about 0.5
ft2/sec) with all rods fully withdrawn results in a reactivity ramp of
0.010% dk/k per sec when k = 1. In filling the core at an abnormally low
temperature, excess reactivity is added by means of the negative tempera-
ture coefficient of the fuel. For fuel B (the mixture with the largest
negative temperature coefficient of reactivity), filling the core with
salt at the liguidus temperature (840°F) produces a reactivity addition
rate at k = 1 of only 0.006%/sec. The greatest excess reactivity and the
highest rates of addition could be encountered if the core were filled
with fuel abnormally concentrated in uranium.

Selective freezing of the fuel salt is a mechanism whereby an ab-
normélly high concentration of uranium could be produced. The crystal-
lization paths of the fuel salt mixtures are such that under equilibrium
conditions, a fraction of the salt can be solidified before uranium (or
thorium) appears in the solid phase. By such selective freezing the ura-

nium concentration can be increased by a factor of 3. This phenomenon can
208

only be produced by extremely slow cooling; normal freezing does not re-
sult in any separation. t ig very unlikely that the salt in the drain
tanks would be cooled slowly enough to produce uranium separation. It is
even more unlikely that the concentrated ligquid would be transferred to
the core, since the solid phase would remelt if the salt were heated to
the normal fill temperature. Nevertheless, the consequences of a severe
cagse of selective freezing were examined.

The reactor vessel is the first part of the fuel loop to be filled,
and only 61% of the salt inventory is required to raise the level to the
top of the graphite in the core. Therefore it was assumed that all the
uranium was concentrated in this amount of salt. (It was assumed that
the solid was composed entirely of 6LiF-BeF2-ZrF4.) If the core could be
filled with this concentrated salt at 1200°F, the excess reactivity would
be sbout 4% 8k/k for fuels A and C and 15% 8k/k for fuel B. (The differ-
ence is caused by the abvsence in fuel B of thorium or much U238, which act
as poison in the concentrates of fueis A and C.) The reactivity addition
rate at k¥ = 1 also is muck the highest for fuel B: 0.025% 8k/k per sec
compared with 0.0l%/sec for fuels A and C.

The rates of reactivity incresse through criticality are high enough
that a considerable power excursion will result and cause the safety sys-
tem to drop the rods at 15 Mw. In the severe filling accident which has
been postulated, the excess reactivity available is greater than the shut-
down capacity of the contrcl rods. Therefore it is necessary to stop the
filling to prevent the rezctor from going critical again after the rods
are dropped. This is done by manipulation of valves in the reactor gas
systemn.

A simplified flowsheet of the reactor fill, drain, and vent systems,
showing only those features that pertain directly to the filling accident,
is presented in Fig. 7.5. All valves are shown in the normal positions
for pressurizing fuel drsin tank No. 1 to force fuel up intc the core.
Three independent actions are available to stop the addition of fuel to
the primary loop:

1. Opening HCV-544 equalizes the loop and drain tank pressures.
209

UNCLASSIFIED
ORNL-DWG 63-7320

 

£y
] <}
EP PCv-522
_ FHX /___I" HCV-v 5 .
= 533 4
> TO FILTER,

(><]

FAN, AND STACK

<]
1>

    

REACTOR

>
2
FFT
FD-2 AUXILIARY
CHARCOAL

 

FFT BED
FO-2
X D

Hov-$

544 HCV-573

P>
73, 50-psig
DISK
FD-2

FFT

 

 

40-psig
HELIUM
SUPPLY

Fig. 7.5. Bystem Used in Filling Fuel Loop.

2. Opening HCV-573 relieves the pressure in the drain tank by vent-
ing gas through the auxiliary charcoal bed to the stack.

3. Closing HCV-572 stops the addition of helium to the drain tank.

During a filling accident all three actions would be initiated auto-
matically (at 15 Mw) to ensure stopping the fill. Either of the first two
actions stops the fill almost immediately and alsoc allows the fuel in the
primary loop to run back to the drain tank. If the only action is to stop
the gas addition, the fuel does not drain back to the tank, and the level
"coasts" up for some distance after the gaé is shut off. This '"coastup"
is a consequence of the pressure drops in the fill line and the offgas
line during filling. Because of this, when gas addition is stopped, the

fuel level in the primary loop continues to rise until the dynamic head
210

losses have been replaced by an increase in the static head difference
between loop and drain tank.

In the case of a filling accident, the inherent shutdown mechanism
for power excursions is less effective than normally because the tempera-
ture coefficient of reactivity of the fuel in the partially filled core
is substantially less than that in the full system. This 1s so because,
in the full system, the size of the core remains constant, and expansion
causes some fuel ©o be expelled. In the partially full core, on the other
hand, fuel expansion increases the effective height of the core, tending
to offset the decrease in reactivity due to increased radial neutron leak-
age. The effective fuel temperature coefficient of reactivity of fuel B
with the core 60% full is approximately —0.4 X 10”5/°F compared with
-5.0 X 107%/°F for the full core. The temperature coefficient of the
graphite is not significantly affected by the fuel level.

The most severe of the postulated filling accidents was analyzed in
detail. It was assumed that the urarium ir fuel B was concentrated to 1.6
times the normal value by selective freezing of 39% of the salt in the
drain tank, and several other abnormal situations were postulated to occur
during the course of the accident, as follows:

1. Only two of the three control rods dropped on request during the
initial power excursion.

2. Two of the three actions available for stopping the fill failled
to function. Only the least effective action, stopping the gas addition,
was used in the snelysis. This allowed the fuel level to coast up and make
the reactor critical after the two control rods had been dropped to check
the initial power excursion.

3. The helium supply pressure was 50 psig, the limit imposed by the
rupture disk in the supply system, rather than the normal 40 psig. This

—

pressure gave a fill rate of 0.5 ftB/min when criticality was achieved and
produced a level coastup of 0.2 ft after gas addition was stopped.

The results of calculations of the power and temperature transients
associated with the accident described above are shown in Figs. 7.6 and
7.7. Figure 7.6 shows the externally imposed reactivity transient exclu-
sive of temperature compensatiocon effects. The essential features are the

initial, almost-linear rise which produced the first power excursion as
211

Unclassified
ORNL-DWG 64-6

ret
4

1.0

 

C.5F

Net Reactivity {%)
s
h

ety

.....

,,,,,,

0 50 100 150 200 250 300 350
Time (sec)

Fig. 7.6. Net Reactivity Addition During Most Severe Filling Acci-
dent.

fuel flowed into the core, the sharp decrease as the rods were dropped,

and the final rige as the fuel coasted up to its equilibrium level. Figure
7.7 shows the power transient and some pertinent temperatures. The fuel
and graphite fiuclear average temperatures are the quantities which ulti-

mately compensated for the excess reactivity introduced by the fuel coastup.
212

Unclassified
ORNL-DWG 64-605

1360

1320

1280

Temperature (°F)

1240

 

1200

)

Power (

 

Fig. 7.7. Power and Temperazure Transients Following Most oevere
Filling Accident.
213

The maximum fuel temperature refers to the temperature at the center of
the hottest porticn of the hottest fuel channel., The power at initisl
criticality was assumed to be 1 watt. The initial power excursion was
limited to 24 Mw by the dropping control rods, which were tripped at 15 Mw,
This excursion is not particularly important, since it did not result in
much of a fuel temperature rise. After the initial excursion, the power
dropped to about 10 kw, and some of the heat that had been produced in the
fuel was transferred to the graphite. The resultant increase in the graph-
ite nuclear average temperature helped to limit the severity of the second
power excursion. Reactivity was added slowly enough by the Tfuel coastup
that the rising graphite temperature was able to 1imit the second power
excursion to only 2.5 Mw., The maximum temperature attalned, 1354°F, is

well within the range that can be tolerated.

7.1.5 Tuel Additions

Small amounts of uranium will be added to the circulating fuel during
operation to compensate for U232 burnup and the buildup of long-lived poi-
sons. The design of the fuel-addition system 1s such that only a limited
amount of uranium can be added in any one batch, and it is added in such
a way that it mixes gradually into the stream circulating through the core.
These limitations ensure that the reactivity transients caused by a normal
fuel addition are inconsequential.

Fuel is added during operation through the sampling and enriching
mechanism. Frozen salt (73% LiF-27% UF,) in a perforated container hold-

U?3° is lowered into the pump bowl where it melts

ing at most 120 g of
into the 2.7 ft° of salt in the bowl. The 65-gpm bypass stream through
the pump bowl gradually introduces the added uranium into the main circu-
lating stream. The net increase of reactivity, once the 120 g of U233 is
uniformly dispersed in the 70.5 ft? of salt in the fuel loop, is 0.07%
8k/k for fuel B and 0.03% 8k/x for fuels A or C. This increase is auto-
matically compensated for by the servoe-driven control rod.

An upper limit on the reactivity transient caused by a fuel addition
was estimated by postulating that all 120 g of U??° was instantly dispersed

throughout the pump bowl, from where it was mixed into the circulating
214

stream. This produced & step increase in the fuel concentration leaving
the pump, and when this fuel began filling the core there was a reactivity
increase. The largest reactivity increase would occur in the case of fuel
B. (Reactivity effects would be less by a factor of 2 for fuels A and C
because of their higher uranium concentrations and smaller concentration
coefficients of reactivity., The maximum rate of reactivity 1ncrease for
fuel B was only 0.017% 8k/k per sec, which is slower than the rate at which
the regulating rod moves. Therefore, with the reactor under servo controel,
there would be no disturbance in power or temperature. If there were no
rod action the maximum reactivity increase would be 0.09% Bk/k, during the
first passage of the more concentrated fuel through the core. Only moder-
ate transients in power or temperature could be produced by this much ex-

cess reactivity.

7.1.¢ UO, Precipitation

 

The chemical stability of the fuel poses safety problems in a fluid
fuel reactor; for if pheses which are unusuelly concentrated in uranium
can appear, deposits may cocllect and cause local overheating or, if they
shift position, reactivity excursions. The only known way in which ura-
nium could become concentrated in the MSRE fuel-circulating system is by
gross contamination of the salt with oxygen, with consequent precipitation
of UOp. As discussed in Section 1.1, the fuel salt for the MSRE is pro-
tected zgeinst oxidation and subsequent precipitation of uranium by the
inclusion of Zrr, in the fuvel saglt. The ratio of ZrF, to UF,; in the fuel
will be over Z, and mcre than 2 f£2 of water or 9000 ft2 of air would have
to react with the 70 ©t° of fuel salt before so much of the ZrF, was con-
verted to oxide that U0, would begin to appear. The only way tlLst such
quantities of water or air could enter the system in a short period of
time would be some failure or accident during maintenance with the system
open. During operation, only gradual contamination is at all credible.
After any maintenance period and at frequent intervals during operation,
fuel salt samples will be examined and analyzed for zirconium and other
oxides. Thus oxide contamination should be detected long before the ura-
nium began to precipitate, and appropriate action could be taken to pre-

vent such precipitation.
215

Although it is extremely unlikely that the precautions against oxygen
contamination, the protection of the ZrF,, and the surveillance of samples
will fail to prevent UO, precipitation, the consequences of such precipi-
tation were considered. The most likely regions for precipitated UO, to
collect are in the heads of the reactor vessel, where velocities are
lowest. Collection in the lower head would create the greatest potential
for a large, positive reactivity disturbance because the lower head is
near the inlet to the core. There is no known way in which a substantial
amount of deposited UOp could be quickly resuspended, either by physical
disturbance or dissolution in the fuel salt. Nevertheless, if some UO;
were resuspended in the salt below the core, it could pass rapidly through
the center of the core and cause a reactivity excursion. The greatest
effect would result if it passed up through the high-velocity (2 ft/sec)
channels around the center of the core,

Excess uranium introduced at the lower end of a central channel would
cause the reactivity to increase to a peak in about 1.5 sec and then de-
crease over a like period. Calculations were made using the reactivity
coefficients and neutron lifetime of fuel C to determine the consequences
of such excursions. The temperature and pressure transients caused by
this type of reactivity excursion with peaks up to 0.7% Bk/k are tolerable
without any control rod action.* Rod drop at a power level of 15 Mw does
not raise the tolerable reactivity addition very much in excursions of
this kind, because the period at 15 Mw is so short (about &80 msec) that
self-shutdown occurs before the rods have time to contribute very much.
Calculations for fuel C indicated that the 15-Mw rod drop raised the toler-
able reactivity peak from 0.7% to about 0.9% 8k/k.

Rather large amounts of excess uranium would be required to produce
the l1imiting excursions. Assuming that all the uranium passed through a
central channel in a single blob, over 700 g of U (0.8 kg of UO,) would
be required to give a peak of 0.7% 8k/k when the fuel is of composition C.

 

*Tolerable here means that the pressure excursion is less than 50
psig and the peak temperature is less than 1800°F for the most unfavor-

able initial power.
216

Equivalent excursions would be produced by about 200 g of U in fuel A or
100 g of U in fuel B.

The question arises: how do these quantities compare with the amount
of U0, whose separation could be detected by its effect on reactivity? A
reactivity effect would be produced 1if the average nuclear importance of
the separated uranium differed from that of the circulating uranium. The
low-velocity regions where precipitated UO,; may collect are at lower-than-
average importance, so the reactivity would probably decrease if UO; were
lost. During critical operation a reactivity balance will normally be
made at 5-min intervals. (The on-line digital computer connected to the
MSRE is programmed to do this automatically, but operation of the computer
is not regarded as a necessary condition for nuclear operation.) In this
balance, the calculated effects of temperature, power, xenon, burnup, and
control rod positions are included. The minirum change in reactivity,
occurring over a period of a few hours, that could be recognized as an
anomaly is expected to be about 0.1% Sk/k. The reactivity change caused
by deposition of UO, where it contributes nothing to the chain reaction
in the core is 0.23, 0.45, or 0.066% 8k/k per kg of UO, for fuels A, B,
or C, respectively. Therefore the minimum amounts of UO,; separation which
could be detected by this means are about C.4, 0.2, and 1.5 kg of UO;
for fuels A, B, and C, respectively.

It is evident that the reasctivity balance cannot be relied on to de-
tect U0, loss below the level at which the concelvable effects of complete
and sudden recovery becorme important. Any UO, deposits would probably be
quite stable, however, so the probability of sudden resuspension of a large
fraction of a deposit is very small. (In the HRE-2, where deposits could
be dispersed by the movement of the loose core inlet diffusers or by steam
formation, and the dispersed material was soluble, the largest "instanta-
neous ' recovery was less than 0.1 of the existing power-dependent deposits
of uranium.) Therefore it is probable that if UO,; precipitation were to
develop, it would produce a detectable reactivity decrease before any
serious reactivity excursion would have a reasonable chance of occurring.
This is only added assuranéé, however, for the real protection against
damaging excursions is provided by the measures which prevent UO, forma-

tion in the first place.
217

Abnormal, localized heating would result if a UO, deposit (or other
solids containing uranium) were located where the neutron flux caused fis-
sion heating in the deposit. One place where the reactor vessel would
most likely be subject to such heating is the lower head. For this reason
thermocouples are attached to the outside of the lower head and the dif-
ferences between these tenperatures and the temperature of the fuel enter-
ing the reactor vessel are monitored continuously to detect abnormal heat-
ing. This is primarily a means of detecting a deposit of uranium solids
rather than a guard against overheating, since, from the standpoint of
damage to the vessel, the heating in the lower head is not expected to be
serious, even 1if fairly heavy deposits of uranium collect. The reason is
that the specific heat source in the uranium-bearing solids on the lower
head would not be intense (3 to 8 w per g of U0, when the power is 10 Mw)
because the neutron flux would be greatly attenuated by the salt and the
INOR grid below the core. Calculated temperature differences at 10 Mw are
shown in Fig. 7.8. (The differences are proportional to power.) The
curves are different for fuels A, B, and C because of differences in flux
and uranium enrichment.

Another region susceptible to local heating due to UO, deposition
1s the pocket between the top of the core shell and the reactor vessel,
above the core shell support flange. There are passages at 10-in. inter-
vals around the periphery of the support flange which bypass some salt
from the vessel inlet for cooling and for providing some rotational flow
in this region. Although the vertical component of the velocity in the
pocket may not be sufficient to eliminate settling of UO;, the swirling
motion should result in an even distribution if deposition should occur.
The area for deposition is large enough (180 in.?) that the loss of ura-
nium from the circulating salt would be detectable before a UO, deposit
could become deep enough to cause intolerable heating in this region. The
maximum vessel temperature in this vicinity was calculated by assuming
that enough uranium separsted to cause loss of 0.2% 6k/k in reactivity
and that all this uranium accumulated as UOs on top of the core shell sup-

port flange. The calculated maximum temperature in the vessel wall was
218

Unclassified
ORNL-DWG 64-609

 

U0, Deposit (kg/ft?)

Effects of Deposited Uranium on Afterheat, Graphite Tem-

Fig. 7.8.
perature, and Core React

ivity.
219

less than 1600°F with the reactor at 10 Mw and the core inlet temperature
at 1175°F.

The possibility of settling on the upper support flange was examined
experimentally by examining the flange area of the full scale hydraulic
mockup after it had been operated with iron filings in the circulating
water, but no preferential settling in this area was observed. In summary,
this examination indicates that the conseguences of uranium precipitation

do not appear to be of a serlous nature from the hazards standpoint.

7.1.7 Graphite Loss or Permeation

 

If a small part of the graphite in the core were replaced by an equal
volume of fuel salt, the reactivity would increase. The effect would not
be large, amounting to less than 0.003% 8k/k per in.? of graphite replaced
by fuel or only 0.13% Sk/k 1f the entire central stringer were replaced
with fuel. The reactivity transients caused by loss of graphite due to
credible mechanisms presert no hazard.

Minor loss of graphite from the core will occur if chips of graphite
break loose and float cut of the core. The vertical graphite pieces are
fastened at their lower erds and are secured by wire through their upper
tips. Thus it 1s very unlikely that a piece of graphite large enough to
cause a noticeable change in reactivity could fioat out of the core.

Another mechanism whereby fuel replaces graphite in the core is the
bowing of the graphite stringers, which causes the volume fraction of the
fuel near the center of the core to increase. The bowing is caused by the
longitudinal shrinkage of the graphite under fast neutron irradiation.

The radial gradient of the neutron flux causes unequal shrinkage on op=-
posite sides of a stringer and produces bowing that is convex in the di-
rection of the gradient. The shrinkage is very siow, and the increase in
reactivity due to unrestrained bowing has been estimated at only 0.6% 6k/k
per full-power year.

Graphite shrinkage also causes small increases in reactivity because
the moderator density is increased and because the lateral shrinkage
gradually opens the channels and allows more fuel in the core. These ef-
fects are of no concern because they cause less than 0,2% ak/k increase

per full-power year and there is no possibility of rapid changes.
220

A reactivity increase would also result from fuel salt permeating the
pores in the graphite and thus increasing the amount of uranium in the
core. There is no known way, however, in which increased permeation could
cause a sudden or dangerous increase in reactivity. Tests invariably have
shown that the fuel salt does not wet the graphite and that penetration
of the fuel into the graphite pores is limited by surface tension, the
size of the pores, and the external pressure. Neither of the first two
variables 1s subject to rapid or drastic change. As indicated in Table 1.5
(sec. 1.1.3), the measured permeability of salt in the graphite is only
0.2 vol % at 150 psig. The pressure effect on permeation produces little
reactivity effect, being less than 5 X 1076 Ek/k per psi and is therefore
unimportant either as a means of externally increasing reactivity or as a
feedback during transients involving pressure excursions.

Another means for increasing the reactivity of the core is sorption
of uranium on the graphite surface, None of the many out-of-pile tests
has shown evidence for such scorption. However, uranium in amounts as great
as 1 mg of uranium vper em® of graphite surface has been observed in a
relatively thin surface layer on specimens from capsules irradiated for
1500 hr to fission power densities considerably above those expected for
the MSRE. TUranium sorption Las been observed so far only on graphite from
experiments in which copious quantities of F, and CF, were evolved by
radiolysis of the frozen fuel mixture. (Such radiolytic evolution prob-
ably occurred several times during the irradiation experiment since the
assemblies were ccoled through reactor shutdowns many times during their
irradiation history.) It is possible that the uranium was laid down on

the graphite by reactions such as
4UF3 + 2C — UC, + 3UF,,

during sudden reheat of the capsule in which the frozen fuel had thus be-
come deficient in Fp. According to this explanation the sorbed uranium
would not be expected in MSRE operation. It is possible, on the other
hand, that the sorbed uranium is due to some unknown (and thermodynami-
cally unlikely) effect of irradiation and fission at elevated temperatures
and that it can occur under conditions of operation of the MSRE. Uranium

in or on the graphite, through whatever mechanism, would increase the
221

reactivity of the assembly, the graphite temperature during operation, and
the quantity of afterheat in the graphite. Figure 7.9 shows each of these
effects as calculated for each of the three fuel compositions previously
discussed, All effects are considerably smaller when salt C (the first
salt proposed for MSRE operation) is used, but the reactivity increase,

in any case, 1s sufficiently large to be hazardous only if it can occur

in a short time.

7.1.8 Loss of Flow

Interruption of fuel circulation can produce power and temperature
excursions through two mechanisms: a reduction in heat removal from the
core and an increase in the effective delayed-neutron fraction. When the
fuel is circulating normally, delayed-neutron precursors produced in the
core are distributed throughout the entire volume of circulating fuel, and
nearly half decay outside the core. If the circulation is interrupted at
any time, the delayed-neutron effect produces a reactivity effect of 0.3%
Sk/k over a period of many seconds. (The increase in the effective de-
layed-neutron fraction would be slow, even if the flow could be stopped
instantaneously, because cof the time required for the precursor concen-
trations to build up to equilibrium.) A core temperature rise of less
than 50°F is enough to compensate for the increased reactivity and, be-
cause of the nature of the reactivity increase, no hazardous power excur-
sion is produced in any case. If the reactor is at high power when cir-
culation is interrupted, the decreases in heat removal from the core and
heat transfer to the coolant will cause fuel temperatures to rise and the
coolant temperature to fall.

The only likely cause of loss of fuel flow is failure of the power
to the fuel pump. The results of a simulated fuel pump power failure at

10 Mw, with no corrective action, are shown in Fig. 7.10. The coastdown

 

in flow after the failure of the pump power was simulated by reducing the
circulation rate exponentially with a time constant of 2 sec until it
reached the thermal-convection circulation rate determined by the tempera-

ture difference across the core. (The flow deceleration was based on pump
222

Uneclassified
ORNL

mperature rise
I core at 30 days

.4

Temperature (°F)

IS

Graphite temperatu
s {at 10 Mw

srature (°F)

Tem

5
6y
L,

k/k)

inereasge

Reactivity (%

 

-

o fLi

0 0.2 0.4 0.6 0. 1.0
(

Uranium Deposited on Graphite (mg/cm?®)

Fig. 7.9. Power and Temperatures Following Fuel Pump Failure, with
no Corrective Action.
223

UNCLASSIFIED
CRNL-LR-DWG 67578

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1
1400 o ——
ORE OUTLET
/
———— CORE FUEL MEAN
1300 .////,, B -t
///i;// hfififii
/ """  GRAPHITE MEAN
1200 dn
¥
g’ iy -
- ——————__| CORE INLET
o —
L \
5 100 e
% RADIATOR INLET
w
= \\
1000 \\\%\\\\\M
\\\
\ \
900 .
\ —\
SALT LEAVING RADIATOR e
h—____- -
-—..‘:
800
12
/\
10 AN
\\QSWER
—_ 8
=
=
- N
m 6 T
)
K I \\
e -
\\
2 ._-.__-__*
0
0 20 40 60 80 100 120
TIME (sec)
Fig. 7.10. Power and Temperatures Following Fuel Pump Power Failure.

Radiator doors closed and control rods driven in after failure.
224

loop measurements. !

The heat transfer and thermal convection were pre-
dicted from basic data.) The initial increase in nuclear power was caused
by the delayed-neutron effects. The decrease in heat removal from the
core, coupled with the high production, caused the core outlet temperature
to rise. As the fuel flow and the heat transfer in the heat exchahger
fell, the continued heat extraction at the radiator caused the cocolant
salt temperature to decrease and reach the freezing point in less than

2 min. The behavior in simulator tests at lower power was similar, but
the coolant temperature remained above the freezing point 1f the air flow
through the radiator was such that the initial power was less than 7 Mw.

The effects of & fuel pump power Ffailure are minimized by automatic
action of the safety and control systems. At nuclear powers above 1.5 Mw,
control interlocks start a rod reverse (insert rods at 0.5 in./sec) as
soon as the pump begins to slow down. A rod reverse is called for in any
event if the core outlet temperature reaches 1275°F. At an outlet tem-
perature of 1300°F the safety system drops all rods. These actions hold
down the heat generation and core temperatures. The coolant salt is kept
from freezing by other szction. When the pump stops, an automatic "load
reverse' lowers the radiator air flow to about 0.2 of the normal 10-Mw
rate. The safety system drops the doors to shut off the air completely
if the temperature of the salt leaving the radiator reaches 900°F. These
actions prevent freezing of the coolant salt in the radiator.

Simulator results that illustrate the effectiveness of actions simi-
lar to those designed into the control and safety systems are given in
Fig. 7.11. Rod insertion was simulated by a negstive reactivity ramp of
~0.075% 6k/k per sec beginning 1 sec after the pump power was cut. The
radiator door action in the present control system was not simulated.
Instead the simulated heat removal from the radiator tubes was reduced
to zero over a period of 30 sec beginning 3 sec after the pump power

failure.

 

10, W. Burke, "MSRE — Analog Computer Simulation of the System for
Various Conditions," internal ORNL document CF-61-3-42, March 8, 1961.
225

UNCLASSIFIED
ORNL-LR-DWG 67579

1300
CORE QUTLET

GRAPHITE MEAN

 

w
° 1200
E FUEL MEA
kel
x
> CORE INLET
g
o
wi
; 1100
= RADIATOR INLET
—
1000 RADIATOR OUTLET
FISSION POWER
>
=
o
wi
=
O
o

 

O 20 40 60 80 100 120
TIME (sec)

Fig. 7.11. Effects of Afterheat in Reactor Vessel Filled with Fuel
Salt After Operation for 1000 hr at 10 Mw,

7.1.9 Loss of Load

The type of incident usually referred to as the loss-of-load accident
takes the form, in the MSRE, of interruption of heat removal from the ra-
diator while the reactor is operating at high power. The most likely way
for this to occur would be for the radiator doors to drop. Failure of
the cooling air blowers would cause less of a load drop because natural
draft through the radiator can remove up to 3 Mw of heat.

The thermal and nuclear characteristics of the MSRE are such that
core temperatures do not rise excessively in any loss-of-load accident,

even with no external control action. Temperature transients are mild
226

and there is no core pressure surge. This behavior was shown in simulator
tests in which the heat removal from the radiator tubes was instantaneously
cut off with the reactor initially at 10 Mw. The temperature coefficients
of reactivity for fuel C were used and no control rod action was assumed.
Tn these tests the core outlet temperature rose less than 40°F, over a
period of about 4 min. The average temperature of the coolant salt rose
196°F, with 100°F of this rise occurring in the first 60 sec. The coolant
salt volume increased by 1.0 ft3, which, if there were no gas vented from
the coolant pump bowl, would cause a pressure rise of 20 psi. The pressure
control system would be capable of limiting the pressure rise to less

than 11 psi.

7.1.10 Afterheat

The problems asgsociated with the decay heat from the fission prcducts
(afterheat) are quite moderate in the MSRE because of the relatively low
specific power in the reactor. Thus temperatures change slowly and no
rapid emergency action is required to prevent overheating. This may be
illustrated by considering a hypothetical situation in which the reactor
vessel remains full of fuel, with the circulation stopped, after long-
term, high-power operation. The vessel contains three-fourths of the fuel
salt in the system and gbout three-~fourths of the fission products. The
total heat capacity of the fuel, the graphite, and the metal vessel is
8030 Btu/°F. The estimated heat loss from the vessel at 1200°F is about
30 kw. The effects of fission-product heating in the vessel after 1000
hr at 10 Mw are shown in Fig. 7.12. It appears that no serious increase
in temperature will result from afterheat if the fuel is trapped in the
core.

After the fuel is drained from the reactor vessel there will still
be some afterheat in the core from fission products which remain in the
graphite. These products are there because of diffusion of gaseous prod-
ucts into the graphite and because of direct production by fissions in the
fuel which had soaked into the graphite. The heating rate in the drained
core 1 hr after shutdown from extended operation at 10 Mw is 4.5 kw, most

of which is produced by the gaseous fission products and thelir daughters.
22

Unclassgified

CRNL-DWG 64-607
600

+

ANo’heat

- Temperature
I Televa bl o
+ AA‘l,

T — To (°F)
w
O
o

100§

 

100

&
&
Q
5
a 40
O
"
g
o
<
i
a
9
o 20
©
2
2
as]

 

10

0 20 40 60 80 100 120 140
Time (hr)

Fig. 7.12. Temperature Rise of Fuel in Drain Tank Beginning 15 min
After Reactor Operation for 1000 hr at 10 Mw.
228

After 24 hr the heating rate is down to only 0.44 kw, so heat losses from
the graphite will probably keep the temperatures from rising significantly.
But even if there were no heat loss from the core graphite (which has a
heat capacity of 3350 Btu/°F), the temperature rise would not be hazard-
ous: 27°F in one day, 144°F in 30 days. These values were calculated by
assuming that fuel occupiles 0.1% of the graphite volume. For this condi-
tion, the fission products produced in situ produce only one-third or less
of the total heating, so the results are not very sensitive to the exact
amount of permeation.

If the fuel were transferred to a drain tank shortly after high-power
operation, the rate of temperature rise in the absence of cooling would
be higher than in the core because the total heat capacity connected with
the fuel would be less in the drain tank than in the core, However, the
drain tanks are provided with cooling tubes designed to remove 100 kw of
heat from the salt at 1200°F. This heat-removal capacity is adequate to
prevent & significant rise in temperature. The predicted heat loss from
a drain tank at 1200°F is 20 kw. Figure 7.13 shows the expected tempera-
ture rise of the fuel charge in a drain tank beginning 15 min after the
end of 1000 hr of operation at 10 Mw. (It will take approximately this
long to get the salt into the drain tank. The temperature at this time
will depend on the manner of shutting down the nuclear power and the ra-
diator heat removal, but it will be near 1200°F.) The heat removal rates
of 20 and 120 kw correspond, respectively, to heat losses and the heat
removal with the cooling system working as designed. The heat capacity
of the INOR tank and structure was ignored in computing the temperature
rise.

As was described in Section 1.2.5, the cooling system for the drain
tanks is based on the evaporation of water, but the water and salt are
not in contact with a common wall., The evaporation-condensation circuit
is completely closed and requires no pumps because the condensate is re-
turned by gravity. An emergency reservoir of 500 gal is installed to
provide cooling for 6 to & hr. During this time, hoses could be connected

to provide a temporary supply for an indefinite time.
229

33
O
-~
4y
ol
©v
0
@
—
&
5

]
:

5

arnyexadus],

100

 

~100

Time (hr)

Heating of Reactor Vessel Lower Head by U0, Deposits

7.13,
with Power Level at 10 Mw.

Fig.
230

7.1.11 Criticality in the Drain Tanks

 

Fuel salt having the uranium concentration appropriate for power
operation of the reactor cannot form a critical mass in a drain tank or
in a storage tank under any conditions. Only if there were a large in-
crease in uranium concentration could criticality occur in a tank. The
only credible way for this to occur would bhe for the salt to freeze gradu-
ally, leave the UF, in the remaining melt, and thereby concentrate the
uranium intc a fraction of the salt volume. Calculations indicate that
concentration by more than a factor of 4 would be necessary for criticality
in any such case.?

Concentration of the uranium is not impossible; gram-scale studies of
equilibrium cooling of fuel salt mixtures indicated that the last phase to
Treeze was 7LiF-6(Th,U)F4 containing about three times the uranium concern-
tration of the original mixture.? However, 1in a vessel as large as the
frel drain tank, it 1s unlikely thal such gross segregation would occur.
Convection currents within the lurge mass and initial freezing on the many
cooling thimbles should keep the segregated salt distributed through the
mess. Even so, special measures have been taken to prevent freezing.

Heater feilures are not lixzely to cause accidental freezing, since a
temperature of 1200°F can be maintained with only about 0.6 of the installed
heaters in cperation. Even 1f the heater vower should be cut off, more
than 24 hr would be recuired for the heat losses (about 20 kw) to ccol the
salt to the liguidus temperature of &40°F. TFission-product afterhesat
would extend this tire by many hours. A complete power failure for this
length of time ig extremely unlikely because the reactor ares is supplied
by two separste feeder lines and there are three emergency dissel-clectric
generating units in the area, any of which could supply vower to the
heaters. In the event =11 these sources failed, or for any reason it
should become necessary to allow the fuel salt to sclidify and cool to

ambient temperature, the risk of criticality due ic selective freezing

 

?J. R. Engel and B. E. Prince, "Criticality Factors in MSRE Fuel
Storage and Drain Tanks, " USAEC Report ORNL-TM-752 (to be published).

3 0ak Ridge National Laboratory Status and Progress Report November
1963, p. 15, USAEC Report ORNL-3543.
231

will be eliminated by dividing the fuel between two tanks, neither of which
will contain enough uranium for criticality. This can be done because one
drain tank will be kept empty for such an emergency.

Although criticality in the drain tanks is undesirable and can be
avoided as described above, the consequences would not be severe. Because
criticality could occur only after at least three-fourths of the salt had
frozen, the initial temperatures would be moderate. When the concentrat-
ing process slowly reached the point of criticality, the power would rise
to match the heat losses without any serious power excursion. (The salt
contains a strong (a,n) source; the reactivity would be increasing very
slowly through the critical point, and there is a negative temperature
coefficient of reactivity.) The establishment of a self-sustaining fis-
sion chain reaction in a tank should therefore present no threat to the
containment of the fuel. TIFurthermore, the location of the tanks and the
shielding around them is such that radiation from a critical tank would
not prevent occupancy of the building and emergency work toward restoring

heater power.

7.2 Nonnuclear Incidents

 

7.2.1 Freeze-Valve Failure

 

The freezing and thawing of the freeze valves could conceivably re-
sult in a rupture of the piping. This is specifically minimized in the
design by meking the freeze-valve gsection as short as possible so that
there 1s small danger of bursting as a result of entrapment of liquid be-
tween the ends of a plug. Because the salt expands as it thaws, special
precautions are taken to apply the heating so that expansion space is
always available. A single freeze valve has been frozen and thawed more

than 100 times without apparent damage.

7.2.2 Freeze-Flange Failure

 

The flanged Jjoints in the fuel-circulating loop might also fail. The
freeze flange was selected as the simplest and most reliable joint avail-

able, and the strengths of the bolting and the flange compression members
232

are considerably in excess of those necessary to maintain a tight joint.
Typical flanges have been tested by thermal cycling over a period of more
than two years, and there has been no indication of unsatisfactory per-
formance., The helium-pressurized leak-detection system will monitor con-

stantly for leakage and will protect against leakage to the outside.

7.2.3 Excessive Wall Temperatures and Stresses

 

v

Overheating of pipe and vessel walls might occur because of external
electric heating or internal gamma heating. The heater elements have a
melting point several hundred degrees above that of the INOR-8, but it is
not considered likely that the THOR-E could be melted by external heating
sC long as salt was present ilnside the pipes. 1If salt were not present,
the pipe wall might melt before the heater element melted, but in this
case only a small fraction of the activity would be released. Wall tem-
peratures will be measured by hundreds of thermocouples and will be con-
stantly scanned to prevent excessive heating. Also during prestartup
testing, the controls of sll external heaters and the resistance-heated
drain line will be provided with mechanical stops set to limit the power
input of each circuit to values that meet the heat input requirements
with a small excess capacity.

During reactor operation, various components will be exposed to high
gamma fluxes that will cause garmma heating of the components. If this
heating produces large teuwperature gradients, excessive thermal stresses
may arise. The effects of germa heating on the core vessel and the grid
structure were investigated because these structures will be in regions
of the highest gamma flux.

The gamma heating of the core vessel will result in a temperature
difference of only 1.3°F ascross the vessel walls and will produce a cal-
culated thermal stress of 300 psi, which is not serious. The temperature
difference across each grid of the support grid structure will be 3.8°F,
and the resulting thermal stress will be 850 psi, which 1s also not seri-
ous,

Camma and beta heating of the top of the fuel-pump bowl will result

in a maximum thermal stress of about 8000 psi at the junction of the bowl
233

with the volute support cylinder, with cooling air flowing at 200 cfm
across the bowl. The Junction temperature under these conditions will be
about 1000°F. Considerably higher stresses will occur at cooling air flow
rates other than 200 cfm during power changes. The highest stress, about
19,000 psi, will occur in changing from O tc 10 Mw, with the cooling air
flowing at 50 efm. With a gas pressure of 5 psig in the bowl, the maxi-
mum combined circumferential and meridional pressure stress will be about
300 pei in the pump bowl and about 400 psi in the volute support cylinder
at the Jjunction. The resulting maximum combined stress with cooling air
flowing at 200 cfm will be 8400 psi, and, with ailr flowing at 50 cim,
19,400 psi.

In a normal thermal cycle after a complete shutdown, the temperature
of the reactor will vary between 70 and 1300°F. With such a range, there
are possibilities for excessive stresses as the piping expands and con-
tracts. Particular attention has therefore been given to providing a
flexible layout. Analyses of the extreme conditions have indicated that
the maximum stress caused by expansion or contraction is only 7050 psi.
Instrumentation is provided to observe the normal rate of heatup or cool-
down, which will not exceed 120°F/hr. Thus it does not appear reasonable
that primary contaimment fallures will occur as a result of excessive tem-

peratures or thermal stresses.

7.2.4 Corrosion

Another possible cause for failure of the primary container is cor-
rosion. As reported in Section 1.1.2, the corrosion rates experienced
with the INOR-& alloy have been very low (less than 1 mil/yr) for periods
as long as 15,000 hr. All available evidence indicates that it is ex-
tremely unlikely that corrosion will be a cause of piping failures.

TNumerous corrosion tests have been completed with fuel mixtures of
the type to be utilized in the MSRE. Results from 37 INOR-8 thermal-
convection loops, 17 of which operated in excess of one year, have shown
complete compatibility between INOR-8 and the beryllium-based fluoride
systems. Experiments conducted in INOR-8 forced-convection loops for one

year or more similarly have shown low corrosion rates in fluoride mixtures
234

of this type. The operating conditions of these experiments are shown

below:
Forced- Thermal -
Convection Convection
Variable Loops Loops
Fluid-metal interface 1300°F 1350°F
temperature
Fluid temperature 200°F 170°F
gradient
Flow rate ~2 gpm ~7 fpm

Metallographic examinations of INOR-8 surfaces following salt exposure
in these loop experiments reveal no significant corrosion effects in time
periods up to 5000 hr. At times longer than 5000 hr, =a thin (less than
1/2 mil) continuous surface layer develops at the salt-metal interface.

Some typical weight-loss data are presented below:

Weight Loss

 

 

Time

(hr) mg/cm2 mg/cmz/mo
5,000 1.8 0.26
10,000 2.1 0.15
15,000 1.7 0.08

Sixteen in-pile capsules exposed for 1500 to 1700 hr each* and five
in-pile fuel-circulating lOOps5 which ran a total of 3000 hr have been
examined Tor evidence that corrosion under irradiation is different from
that out of pile. Particular attention was paid to the possible effects
of free fluorine. Mo evidence was found which indicated that high ir-
radiation levels altered the normal corrosion pattern.

It has been suggested that the high radiation levels in the control-

rod thimbles and in the reactor cell might cause formation of sufficient

 

40gk Ridge National Laboratory, "MSRP Semiann. Prog. Rep. Jan. 31,
1963, " USAEC Report ORNL-3419, pp. 80-107, and "MSRP Semiann. Prog. Rep.
July 31, 1963," USAEC Report ORNL-3529, Chapter 4.

5D. B. Trauger and J. A. Conlin, Jr., "Circulating Fused-Salt Fuel
Trradiation Test Loop," Nucl. Sci. Eng., 9(1): 346~356 (March 1961 ),
235

nitric acid (from the Ns, 05, and II,0 in the cell) to create corrosion
problems. A calculation for the most pessimistic conditions indicates
that the concentration of nitrogen oxides in the cell atmosphere might
reach 1% in 4000 hr, the longest anticipated period of continuous opera-
tion. The actual concentration should not be this high. The HRE-2 was
operated under similar conditions without observable damage from HNO3, so
no real problilem is expected.

Corrosion in the MSRE will be followed continuously by means of the
salt chemistry, which will indicate nickel removal. Salt samples for this
purpose will be removed and analyzed daily. At approximately é-month in-

tervals, the surveilllance specimens referred to below will be tested.

7.2.5 Material Surveillance Testing

 

Surveillance specimens of INOR-8 and type CGB graphite will be placed
in a central pogsition of the reactor core (as shown in Fig. 1.6) to survey
the effects of reactor operations on these materials from which the re-
actor and moderator are constructed. The specimens will be made from the
material stock used to fabricate the reactor primary system and moderator.
Separate but identical control specimens will be exposed to a duplicate
thermal history while submerged in unirradiated fuel salt in an INOR-8
container to differentiate between the effects of temperature and the
effects of drradiation. ©Specimens will be removed from the reactor and
examined after six months of operation and then again after periocds of
operation that should result in meaningful data based on the results of
the first set of specimens. New specimens will replace the material re-
moved.

The analysis of INOR-8 specimens will include metallographic examina-
tion for structural changes and corrosion effects, mechanical properties,
and a general check for material integrity and dimensional changes. The
graphite specimens analysis will include metallographic and radiographic
examination for salt permeation and possible wetting effects, mechanical
properties tests, dimensional checks for shrinkage effects, chemical analy-
sis for deposits on the graphite, electrical resistivity measurements, and

a general inspection for material integrity.
236
The results from surveillance specimens will be available in plenty
of time to guide decisions on the allowable safe exposure of the graphite

and INOR at varicus locations in the reactor.

7.3 Detection of Salt Spillage

 

The escape of activity from the primary container would be detected
by radiation monitors. If the spillage were into the secondary container,
the activity would be indicated by monitors on a system which will con-
tinuously sample the cell atmosphere. Leakage into a service line (e.g.,
the cooling water) would be detected by monitors attached to the line just
outside the cell. In either case, the action of the monitors would be to
stop power removal and to insert the control rods. The salt would be
drained unless the leak were 1n a drain tank, in which case it would be

transferred to another tank.

7.4 Most Probable Accident

 

Although the chances of any failure are extremely small, one of the
above described nonnuclear incidents can be considered as the most likely
cause of leakage Trom the primary containment system. Based on previcus
experience with fluid-fuel reactors, it is believed that the most probable
type of leakage would be a slow drip or spray at a rate of a few cubic
centimeters per minute. It is further believed that the leakage would be
detected and the reactor shut down before more than 3 or 4 liters had es-
caped. By this time, about 20,000 curies would have been released into
the secondary container. If 10% of the solid fission products (approxi-
mately 1400 curies), 10% of the I, (approximately 480 curies), and all the
xenon and krypton (approximately 3200 curies) were dispersed in the cell
atmosphere, the concentration of activity inside the container would be
107 curies/cm3, which would alarm the cell air radiation monitors and
shut down the reactor. This activity (neglecting decay, for simplicity)
would be pumped from the cell by the container vacuum pumps as they main-

tained the cell pressure below atmospheric by compensating for the 10“6%/min
237

normal inleakage (at 13 psia). The activity would be discharged through
line 917 to charcoal bed 3 and then to the absolute filter-stack system.

Assuming that the carbon bed and the absolute filters are reasonably
effective, the disposal of the activity would not be a large problem, and
concentration could easily be kept below the MPCa. Even without the char-
coal bed the concentration of iodine would probably be tolersble (see
Appendix D).

After an accident such as that described above, it would be necessary
to open the secondary container to repair the fallure, and the activity
would be released at a rate considerably greater than the normal rate of
cell inleakage. In this circumstance, the container inleakage would be
increased manually, and the concentration of activity in the stack would
be increased to levels which would permit meximum discharge without ex-
ceeding permissible exposures downwind. In any case, the container would
not be opened for repairs until the activity concentration inside was low

enough to prevent overexposure in case the cell ventilation system failed.
238
2. DAMAGE TO THE S=ECONDARY CONTAINER

The possibilities for demage to the secondary container were con-
sidered. As indicated below, damege from missiles is unlikely, and pro-
tection is provided against the puildup of excessive internal pressure.
Site studies have indicated that no problems exist with respect to earth-
quakes and floods, and protecticn hzs been provided against the conse-
quences of arson. The analysis cof tThe maxinum credible accident, which -
involves simultanecus release of the meolien fuel and water intc the sec-
ondary container, led to the incorporaticn of a vapor-condensing system .
fer limiting the container pressure to 40 psig. The conseguences of
release of radicactivity frcom the secondary container atfter the maximum
credible accident were analyzed Ior two situations: with the stack fan
off and with the stack fan still running and passing the effluent thrcugh
ebeclute filters. ALl doses were found 1o be low encugh to allow evacua-

ticn of the bhuilding or arez in reascnable Times without overexposure.

£.1 Mgsile Damage

Damage by missiles does not appear to be likely. The maximum pres-
sure expected in the reactor system is lese thar 100 psig, and the INOR-8
strucTural material 1is very ductile at the ncrmal cperating temperature.
no very large pressure excursicn can be envisicored without assuming that

the core inlet and exit liznes are both frozen. Although missiles with
significant velccities are not thought to be credible, the reactor vessel
ig protected from missiles by the stainless steel thermal shield that

completely surrcunds 1t.

2.2 Excegsive Pressure

 

£,2.1 Salt Spillage

The spillage of the salt at 3 high temperature does have possibilities
for raising the cell pressure to high values. A rapid spill intoc the cell

of all the salt in both the fuel and coclant systems would heat the cell
239

atmosphere sufficiently to produce a 2.4-psig final pressure.l If the fuel
were released as a fine spray, the maximum pressure would be 16.4 psig.

The worst situation would be the simultaneous release of the salt
and the inleakage of the correct amount of water to allow the generation
of steam without subsequent cooling from additional inleakage of water.
This accident is considered tc be the maximum credible accident and is

discussed in detail in Section 8. 6.

8.2.2 0il ILine Rupture

 

The fuel-pump lubrication system contains a maximum of 28 gal of oil,
which, in the event of an oil-line rupture, could come into contact with
the hot pump bowl and the reactor vessel. With an atmosphere containing
the normal 21% of oxygen in the cell, the oil would burn and produce an
excessive pressure in the cell, cr it would form an explosive mixture
that might later be ignited.

To ensure against containment damage by these possibilities, the
oxygen content of the cell will be kept below 5% by dilution with nitro-
gen. Nitrogen will be fed into the cell continuocusly to maintain the low

oxygen ccntent.

8.3 Acts of Nature

8.3.1 Farthguake

As reported in the site description, Section 4.3.8, earthquakes are
not a problem at this site.
g8.3.2 TFlood

The topography of the MSERE site indicates that flooding is highly
unlikely.

 

1S. E. Beall, W. L. Breazeale, and B. W. Kinyon, "Molten-Salt Reactor
Experiment Preliminary Hazards Report, " USAEC Report ORNL CF-61-2-46, Oak
Ridge National Laboratory, February 1961.
240

8.4 Sabotage

Severe damage to the reactor by sabotage would be difficult; arson
is probably the best possibility. The consequences of fire are minimized
by fireproof structures, a sprinkler system, and an alarm system that auto-
matically transmits to the fire-protection headquarters at X-10. Water
for this system is supplied to the building sprinklers by the main line
from X-10 and two 1.5 million-gallon reservoirs near the site. Only a
person with intimate knowledge of the reactor would be capable of inflict-
ing damage that might result in reactor hazards. This possibility is
minimized by adequate personnel policiles and security regulations, particu-

larly with respect to visitors.

8.5 Corrcsion from Spilled Salt

 

Tn an accident invelving contact ¢f water and flucoride salts in the
containment vessel, fairly rapid generation of hydrofluoric acid is ex-
pected. Part of the I will be dispersed as vapor in the cell and part
of it will dissclve in the water. Corrosion damage resuliing from dis-
solved IF is gre=tly deperden® upcn the temperature and the concentra-
<fon. Cerrosion penetraticns ranging from 0.009 in. per year tc 0.79 in.
per vear with temperature ranges from 70 to 300°F and HI' concentrations
from 10 to 100% by weight have been reporzed. Sirce the bottom hemi-
spherical section of the containment vessel is l-in.-thick plate with a
protective paint ccating, there is 1ittle irmediate danger of ccrrosion-
caused weakening of the cell. Provision has been made for the addition
of & neutralizing scluticn in case experiments indicate that the corro-

sion might damage the vessel within several months after a spill.

2.6 Maximum Credible Accident

The maximum credible accident for the MSRE involves release of the

molter fuel from the reactor vessel and fuel-circulation system into the

 

2Handling of Flucrine and Fluorine Compounds with Inco Nickel
Aloys, The Internaticnal Nickel Co., Inc., New York, N. Y.

 
241,

reactor cell.” This can occur as a result of a break in the 1 1/2-in. -
diam drain line or in a 5-in.-diam fuel-circulaticn line. The fuel cir-
culation lines enter and leave the reactor vessel near the top; so 4000 1b
or less of salt would be discharged through a severed line; this could
occur in about 15 sec. All 10,000 1b of salt could be discharged through
a severed drain line in about 370 sec or through a combination of severed
drain and circulation lines in about 280 sec. The calculated quantity of
fuel released as a function of time is shown in Fig. 8.1 for the latter
two conditions.

The pressure produced by the release of salt into the secondary
container depends on the reactions with the environment. The oxygen
content of the cell atmosphere will be kept low enough (by purging with
nitrogen) to prevent fires and explcsions. No significant exothermal
chemical reactions can occur between the salt and the materials in the
reactor cell. Any pressure rise must result from heating of the cell
atmosphere and from genération of steam by contact of the salt with
water. The actual pressure rise will depend on the rate of release of
heat from the salt to the atmosphere and to water and on the rate of
dissipation of heat from the atmosphere by space coolers and by transfer
through the walls of the reactor and drain-tank cells. Because of the
many uncertainties in these rates, the reactor and drain-tank cell pres-
sures were conservatively estimated on the basis of (1) equilibration of
the fuel salt with the cell atmosphere and (2) equilibration with an amount
of water that would evaporate completely to form saturated steam.

Equilibration of the salt with the cell atmosphere alone would re-
sult in a rise in cell pressure to about 18 psig and a gas and salt tem-
perature near 1200°F, Because the total heat capacity of the gas is
small in comparison with that of the salt, the final temperature and
pressure calculated in this manner depend only slightly on the amount of
salt spilled if it is in excess of 1000 1b.

Water will be used as the coolant for the cell azir coolers and the

pump motor, and there will be water in the reactor thermal shield. Thus

 

’r. B. Briggs, "MSRE Pressure-Suppression System,” ORNL internal
document MSR-61-135, Nov. 15, 196l1.
10,000

8,000

6,000

4,000

Weight of Salt Spilled (1b)

2,000

8.1.

100

Relesgse

242

Uneclassified
ORNL-IR-TWG 66848

Case 1 — Pirst 4000 1b of salt relessed
from broken 5-in.-diam cirecu-
lating line; remaining 6000 1b

from 1 1/2-in. drain line.

S Case 2 — A11 10,000 1b of salt released
through 1 1/2-in. drain line.

 

200 300 400
Time (sec)

of Fuel Through Severed Lines.
243

there is a possibllity of simultaneous spillage of water and salt into
the cell. Equilibration of all the fuel salt with the cell atmosphere
and just enough water to form the maximum amount of saturated steam would
result in the maximum pressure in the secondary container. The relation-
ship between the amount of salt equilibrated with water and the cell at-
mosphere and the resulting pressure is shown in Fig. 8.2. With no relief
device, pressures as high as 110 psig could result.

The vapor-condensing system described in Section 6.2 is designed with

a rupture disk to relieve at 20 psig, so that after an estimated 10 sec,

UNCLASSIFIED
ORNL DWG. 64-5654

/

100
SALT EQUILIBRATED WITH WATER
(NO PRESSURE RELIEF) fifi‘\\

80

 

120

 

 

 

40 /

 

CELL PRESSURE (psig)

RUPTURE DISK RATED AT 20 psig

 

 

 

 

 

 

 

 

 

20 | —_—
L~ - Vg
7 SALT WITHOUT WATER
opf
-20
0 2000 4000 6000 8000 10,000

WEIGHTS OF SALT AND WATER EQUILIBRATED (Ib)

Fig. 8.2. Relationship Between Cell Pressure and Weights of Fluids
Equilibrated.
244

the steam would flow to the vapor-condensing tank at a rate of about 16
1b/sec. After 5 min® this worst accident should be ended and the pressure
approaching normal. The fission product afterheat would be dissipated
easily to the vessel and its annulus which has a heat capacity of about
4oo,doo Btu/°F.

Two relatively large-scale experiments4 were conducted to study some
of the consequences of a salt spill. The first test was a l/20—scale ex -
periment with 500 1b of salt at 1500°F. The MSRE containment vessel was
represented by a 1l0-ft-diam spherical vessel with a water-filled annulus.
The test demonstrated that large amounts of heat can be transferred at
relatively high fluxes (40,000 to 100,000 Btu/hr-ft?) from the pool {or
cake) of salt to the steel vessel and its annulus without damage to the
tank. The average heat transfer rate for this experiment was 3 X 106
Btu/hr. The temperature of the tank bottom rose to 940°F in 75 sec and
fell to less than 250°F in 6 min.

In the second test, the same gquantity of salt (500 1b) was quickly
drained (50 sec) into the 10-ft-diam tank filled to a 5-ft depth with
water. The salt contained 25 tc 20 me of mixed fission-product activity
(6 days o0ld). Although the physical shock to the water and the tank were
violent, nothing approaching an explosion was observed., Only & small
amount of steam reached the surface of the water. The radicactivity of
the water was measured afterwards and was found to account for less than
5% of the original activity.

These relatively crude but very convincing tests indicate that the
assumptions made in the maxirum credible accident are reasonably conserva-
tive, that the contaimment vessel should withstand the thermal shock of
hot salt, and that most of the fission products remain in the salt. As
mentioned in Section 6.1.1, and &-in. vent pipe is installed to relieve
the steam which might be formed under the bottom of the MSRE containment

vessel in the event of a large szlt spill.

 

“L. A. Mann, "ART Reactor Accident Hazards Tests, " USAEC Report ORNL
CF-55-2-100, Osk Ridge National Laboratory, February 1955.
245

8.7 Release of Radiocactivity from
Secondary Container

 

 

8.7.1 Rupture of Secondary Container

 

Tt is considered incredible that any single accident (sabotage ex-

cluded) could rupture both the primary and secondary containment.

8.7.2 Release of Activity After Maximum Credible Accident

 

As discussed abcve, the most severe accident considered to be cred-
ible is the simultaneous spillage of salt and water into the secondary
container. The release of fission products and the ccnsequences of re-
lease have been studied, fcllowing the general procedure outlined in
Part 100 of the Code of Federal Regulations, Title 10, as published in

the Federal Register. The specific assumptions cn which the caleculatlions

 

were based were the following:

1. The secondary contalner pressure is vented to the vapor-condensing
system at 20 psig. The pressure rises to a maximum of 39 psig and decays
to atmospheric pressure as the steam condenses over a 4-hr periocd (see Ap-
pendix E).

2. The secondary container initial outleskage is 1% per day (0.04%
per hr) at 39 psig.

3. The building outleakage is 10% per day {(0.4% per hr).

4. The iodine release from the salt is 10% or 2.50 x 10° curies,
with a 50% plateout on the secondary container surfaces. (Based on ex-
periments5 in which the solubility of the fuel salt in water was mea-
sured, much less iodine is expected to be released.)

5. The noble gas release is 100% or 3.75 x 10° curies.

6. The solid fission product released is 10% or 6.8 X 10° curies.

The methods of caleculation are given in Appendix A. Two cases were
considered: one with the fan off and the other in which the stack fan

continued to operate to discharge the contents of the building from the

 

5Ruth Slusher, H. F. McDuffie, and W. L. Marshall, "Some Chemical
Aspects of Molten Salt Reactor Safety,  USAEC Report ORNL-TM-458, Oak
Ridge National Iaboratory, Dec. 14, 1962.
246

100-f4-high stack a2t 15,000 ft°/min. With the fan running, the effluent
passes through absolute filters, which have an assumed efficiency of
99.9% for 0.5-u particles.

Activities Inside Building 7503. The calculated dose rates with

 

the fan running are plotted in Tig. &€.3, which shows that the dose rate
reaches a maximum of about 12 r/hr after 1.1 hr. With the stack fan off,
the peak activity levels inside the building are approximately a factor
of 5 higher. The concentraticns of tThe variocus fission-product constitu-
ents inside the building are shown in Fig. &.4.

Personnel within the bullding coculd escape from the bullding high
bey without receivirg the maxinum permissible doses, as indicated in
Table 8.1. Appendix A giveg details of the methed used for these calcu-
lations. The escape times are nct =2 fected appreciably by the operaticn
of the fans; the minimum of 6.5 min (for = bone seeker dose of 25 rem)

is more than sufficient time Tor evacusticn from any lecation within the

 

 

 

building.
Table 5.2. Bullding Escape Time Based on
Maximum Fermissible Dose
Type of Maximum - L Lscape Time
T - A rhalaticn
Figeicm Permissible Tose
Product Doce - Tan On Fan Off
Todine 300 rem 960 ue 40 min 33 min
Heble 25 r ext. o0 2.2 hr
gases
Selids 25 renm 221 ue 7.0 min 6.5 min

 

Dose Rates Outside Building 7503 Following the Maximum Credible Ac-
cident. The meximum activity ir the building is calculated to be 183
curies at 1.1 hr with the fan on and 932 curies at 4 hr with the fan
off {see Appendix A, Tables A.1 and A.2). Figure 2.5 shows the dose rate
at various distances from the building due to the ectivity in the build-
ing. The radiation level from the radicactive cloud (normal atmospheric

conditions) and the total radiation level are plotted on the same graph.
247

Unclassified
ORNL-DWG 64-611

 

50

Noble Gasés
20

10

Todine -

Radiation Level {r/hr)

0 1 2 3 4 5 6 i
Time After MCA (hr)

Fig. 8.3. Radiation Level in Building Following Maximum Credible
Accident (r/hr = 935 X uc/em® X Mev).
248

Unclasgified
ORNL-DWG 64-612

;Iodiné

Activity Concentration (pe/em’)

 

Time After MCA (nr)

Fig. 8.4. Activity Concentrations in Building Air Following Maxi-
mum Credible Accident.
249

UNCLASSIFIED
ORNL DWG. 64-5655

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    
 

 

 

 

 

 

 

 

 

 

100.0 T
80.0—!—j
| \ FAN OFF
60.0 _l._ L — e FAN ON — |
‘ \
40.0 —1 — -
|
|
20@-——r—— ] e
\
10.0 \ FROM BUILDING (ALSO TOTAL, FAN ON)
8.0 \ - . o
£ 60 \ S ——
: \
- \
~ N
o N 40-hr MPC
4 50 VN |
a ’ \ ~\ |
3 \ N TOTAL, FAN ON
- ~N
XTI N
o / \ ~o NN ]
' \ ST FROM RADIOACTIVE
0.8}—— —\, ~ N2 CLOUD
~ YN
\! SO D™
0.6 A —
\\ \\\\
\ \\.
0.4 <
Y FROM BUILDING
N
| \
FROM RADIOACTIVE s
o2f——— CLOUD TN
| N\
“
| ~N
~N
0.10 I N
0 200 400 600 800 1000 1200 1400

DISTANCE DOWNWIND OF MSRE (m)

Fig. 8.5. Noble Gas Activity After Maximum Credible Accident as
Function of Distance Downwingd.

With the fan off, essentially all the radiation is due to the activity
in the building. Beyond 450 meters from the building, which is the ap-
proximate distance to the HFIR, the peak radiation level is less than

2.5 mr/hr (see Appendix A for calculations).
250

Release of Activity from the Building. The release of activity from
the high-bay area of the buillding was considered with the fan off and with
the fan on and under normal and inversion atmospheric conditions. Since
there is an installed spare fan and the fan can be operated from diesel
power 1f necessary, the probability is very low that the fan would be off.
Figure 8.6 shows that the activity release rate is an order of magnitude
lower 7 hr after the maximum credible accident with the fan running. The
meteorological conditions and method of calculation are given in Appendix A.

Figure 8.7 shows the total integrated thyroid and bone dose resulting
from the maximum credible accident. It can be seen that the maximum emer-
gency doses of 300 rem to the thyroid and 25 rem to the bone are exceeded
only at distances within 300 meters downwind of the MSRE with the fan off.
With the fan running the maximum doses are less than 5 rem. The maximum
off-site exposure (beyond 3000 meters) is 6 rem from iodine under inver-
sion conditions with the fan off (see Appendix A for calculations).

The peak activity concentrations are plotted in Fig. 8.8 for iodine
and solids. These peak levels occur at 1.1 hr after the maximum credible
accident with the fan on and 4 hr after the maximum credible accident with
the fan off. The change in release rate with time is plotted in Fig. &.6.
It was assumed that 99.9% of the solids would be removed by the filters
with the fan on, but there would be no removal of iodine. There 1s some
evidence that iodine in large quantities would be removed by absolute fil-
ters by agglomeration of the very fine particles, ® but no credit was taken
for this effect.

In the case of the maximum credible accident coineciding with an un-
favorable wind in the direction of one of the nearby sites, such as HFIR
(460 meters southeast) or CRNL (800 meters northeast), the possibility was
considered that evacuation might be required. A quarterly dose of &€ rem
to the thyroid and 1.5 rem to the bone was considered the maximum allow-

able dose in this case. From Fig. &.7 it can be seen that the doses at

 

6M. H. Fontana and W. E. Browning, Jr., "Effect of Particle Agglomera-
tion on the Penetration of Filters Utilized with Double Containment Sys-
tems, " USAEC Report ORNL-NSIC-l, Oak Ridge National Laboratory, Sept. 25,
1963,
251

Uneclassified

ORNL-DWG 64-615
1072

v -Fan off

- Fan on

- r

-¥Based on activity }eleased‘during
‘maximunm credible accident .

¥

1073

|

Fractional Releasc Per Day*

Q
&
=
0
Ik
- -
=
Ay
O
-
o
Q
S
S e
w’
O
g
P
<
4
cfiA
r_{
-t
Q
-

10™4

 

1073
0 1 2 3 4 5 6 7
Time After Accident {(hr)

Fig. 8.6. Change in Rate of Activity Release from Building with
Time.
252

UNCLASSIFIED
ORNL DWG, 64-5585

 

 

 

 

10Q0 \ T T % 1 X
| *DOSES FROM BUILDING SEEPAGE
N\ N N\
NORMAL INVERSION
CONDITIONS CONDITIONS
N\ \

 

 

 

\

N
AN\ .

MAXIMUM EMERGENCY
THYROID DOSE

 

 

 

d
A
prd

 

 

 

 

 

 

0‘5

AN

|
/e»\*‘

 

 

1 $\00\
// /A

—_MAXIMUM EMERGENCY
BONE DOSE
\\ \
MAXIMUM \
IODINE DOSES
Y, FROM STACK "\
X(NORMAL ®
\ INVERSIONA \
¢ \

 

\

TOTAL INHALED DOSE (rem)

 

 

 

 

 

 

 

 

d

 

 

 

 

 

 

 

 

NOTE: MAXIMUM DOSE FROM STACK \ \
| RELEASE OF SOLIDS IS 1.3mrem FOR _N\

NORMAL CONDITIONS AND 2.0 mrem \ N
FOR INVERSION CONDITIONS. \ \
I N \

o] 100 1000 10,000
DISTANCE DOWNWIND OF MSRE {m)

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 8.7. Total Integrated Doses Following Maximum Credible Acci-
dent.

the above two distances do not exceed the maximum doses for continuous
occupancy except with the fan off under inversion conditions. Under in-
version conditions the total integrated iodine dose at 460 meters would
be 13 rem. However, the dose for the first & hr would be only 0.75 renm
(for a meximum air concentration of C.16 pc/m?), and the maximm dose
would never be reached with an 8-hr working day and the decreasing re-
lease rate. In the case of bone seekers for the same conditions, the

first &-hr integrated dose would approach the 1.5 rem, but unless the
253

UNCLASSIFIED
ORNL DWG. 64-814R

 

l

FAN OFF, 4.0 hr AFTER MCA

NORMAL _ | == e=———FAN ON, 1.1 hr AFTER MCA —]
ATMOSPHERIC
CONDITIONS

  
   
    
  

 

 

 

INVERSICN CONDITIONS

 

 

 

IODINE

_'—'——._-—._

 

—— ]

 

 

 

 

 

 

 

 

 

 

ACTIVITY CONCENTRATION (,uc/m3)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

[~
el g reoM\PC40 = \\
|

 

 

 

 

 

 

 

 

 

\_\
l NORMAL ATMOSPHERIC CONDITIONS S—
04| | L f ’ | LT
O 400 800 {200 1600 2000 2400 2800

DISTANCE DOWNWIND OF MSRE (m)

Fig. 8.8. Peak Iodine and Solids Activities after Maximum Credible
Accident as Function of Distance Downwind.
254

accident occurred at the start of a work period, the dose would be less
than 1.5 rem.
Rainout. The hazards associated with the deposition of all the
fission products in the radicactive cloud due to rainout were considered
for the case of the far cn under ncrmal atmospheric conditicns. Ry in-
tegration of the formulas for the amcunts of lodine, noble gases, and
solids in the building times the release rate with respect tc time, a
total of approximately 400 curies of zctivity was found to be released,
essentially all in the first 4 hr following the maximum credible acciden<.

J
The depcsiticn, in curies/mg, was determined by using the formula,7

Q

¥ ’
welfzc Xz—n/z

v =
R max

 

vhere
¢ = emissicn rate, curies/sez,
. oo cs / /
0 = crosswind diffusion ccefficient, (meters)®/ 2,
y
X = Qovwnwind distance, meters.

The dose rates in repfhr, given in Table £.2, were obltained by multiplyling

 

Y + . " T T - ft a »
777, &. Depurtment of Cormerce, Weather Bureau, Meteorology and Atomic

y - Tt AT T g ATIMNTY TS Taq10 =7 [
“nergy, USALRC Zevort AECU-3066, July 1925,

Table §.2. Dose Rates from Rainout

 

. Maximum Deposition
Distance from b b

 

 

Building 7503 (curies/m?) Dose Rate
(m) : (rep/hr)
Todine Total
50 4 X 107% 0.168 1.68

100 1 x 10772 0.048 0.48

200 3 x 1073 0.013 0.13

400 g8 x 1074 3.5 x 1073 0.035
1,000 1.4 x 10~ 6 X 104 0.006
4,000 1 x 10-° 4.5 X 10-° 0.00045
10,000 1.8 x 10~ 8 x 106 0.00008

 
255
the surface contamination in curies/m® by 10 (ref. 8). For inversion

conditions the deposition is about 30% greater because of the higher

stablility parameter.

8.8 Release of Beryllium from Secondary Ccntainer

 

The volume of fuel salt in the primery loop is 73 ft°, and the weight
of beryllium in the primary system is 220 kg (salt A). Although it is
doubtful that as much as 1% or 2.2 kg of the beryllium could be suspended
in the secondary container atmosphere, even in the event of the maximum
credible accident, this amount was assumed for the hazard calculation.

At a 1% per day initial leak rate from the container (with a pressure of
39 psig), the beryllium concentration in the building would reach the
maximum permissible concentration for workers without respiratory protec-
tion (25 pyg of Be per w?) in 150 sec. This is sufficient time to put on
masks and to escape from the building. TIrom Fig. 8.9, it can be seen that
the beryllium concentrations downwind of Building 7503 would be consider-
ably less than 2.5 ug/m3, the maximum continuous occupational level. The
retention of 99.9% of the beryllium on the filters was assumed.

A second possibility is that beryllium might escape from the coolant
system radiator and be released up the coolant stack. In this case, be-
cause of the high air velocity through the radiator, 10% of the cocolant
charge of 202 kg of beryllium was assumed to be dispersed. Although the
relessed beryllium would be highly dispersed, the maximum downwind con-
tamination levels would be well below the tolerance level, as shown in
Fig. 8.8. At a stack velocity of 32 ft/sec, it was calculated that 3-mm-
diam particles could be carried up the stack. These and smaller particles,
amounting to perhaps half the total weight, would fall out of the stack
plume within the first 100 ft of travel. The average ground concentra-

tion in this area would be about 5 x 1073 g/cm?. The contaminated area

 

83. E. Beall and S. Visner, ”Homogeneous Reactor Test Summary Report
for the Advisory Committee on Reactor Safeguards,” p. 183, USAEC Report
ORNL-1834, Oak Ridge National Iaboratory, January 1955,
256

Unclassified
ORNL-DWG 64-616

 

oncentratlon

Fuel and Flush Salt
- eman o= Coolant Salt

-1 ; A A O
10 ) T T L s = Note: Concentrations due to release
7 = : T : - of fuel and flush salt with
fan running are all <107% ug/m’

Beryilium Concentration (ng/m’)

« Fuel salt, normal -
- atmogspheric con- Co:
iitions, Tan off o

a 500 1000 1500 2000 2500 3000 3500
Zstance Downwind of MSRE (m)

Fig. 8.9. Beryllium Contamination After Maximum Credible Accident
as Function of Distance Downwind.
257

or a larger area, if necessary, could be roped off and washed down with
hoses to reduce the hazard.

Protection against dangerous concentrations of beryllium in the
coolant air stream will be provided by a continuous alr monitor at the
stack; the air inside the building will be monitcred by taking lrequent
air samples from 12 sampling stations. Discharge from the stack can be
stopped within a few seconds by stopping the fans and closing the radia-
tor docrs. Beryllium concentrations will be followed iIn a manner similar
to that used to monitor radicactivity in the air. The frequent samples

will be analyzed and will be reported each day to reactor supervision.
APPENDICES
261

Appendix A

CALCULATIONS OF ACTIVITY LEVELS

Calculations of Building Activity Levels

 

The calculations were based on the following definitions:

a = activity released into cell, in curies,
t = time after maximum credible accident, in hours,
¢ = activity in bullding, in curies,
k = fraction of activity in bullding released per hour,
0.0004 = initial fraction of activity in cell leaking into building

per hour.

It was assumed that the cell pressure would reach atmospheric pressure
in 4 hr, at which time leakage into the building would stop. Then
(0.0004 — 0.0001t) = fraction of activity in cell leaking into building
per hour at t hours, and the curies of activity in the building from O
to 4 hr after the accident would be

t
¢ = j; [(0.0004 — 0.0001t)a — ke] dt ;

thus

Ke'kt + 0.0004ax + 0.0001la — 0.000lakt

k2

 

Cc =

When t = 0, ¢ = O and K/ = K/%?,
K/ = —0.0004ak — 0.0001a .

For t = 0 to 4 hr,

0.0004  0.0001 0.0001t
-kt
a (L —e ) — —,

+
k K k

C =

 

For t > 4 hr,
262
With the fan on, k¥ = 1.875 hr=!, and for t = 0 to 4 hr,

c = a [0.0002405(1 — "1.8750) —-o.oooo534§]

For iodine,

¢ = 30.2(L — e"1-8750 _ g 2205t) .
For noble gases,

¢ = 90.6(1 — e 1-875% _ g 2205t) .
For solids,

¢ = 164.2(1 — e"1-875V _ 5 2205¢) |

With the fan off, k = 0.004 hr™*, and for t = O to 4 hr,

1l

c = a [5.35(1 ~ 70.004Ty _ 0.025€]
For iodine,
c = 7.95 % 10° (1L — e70-004% _ 3 937 x 1073¢)
For noble gases,

2.385 x 10° (1 — e70:004t _ 3 937 « 1073¢)

0
1l

For solids,

0o
lh

4.32 x 10% (1 — e70.004% _ 3 937 x 1073¢) .

Data on the activity in the building are presented in Tables A.1 and

A.2.

Calculation of Exposures and Building Escape Times

 

The internal exposure was determined from the following expression:

 

108 [
E == j; ¢ at
Table A.1l.

Cell leak rate:
Fan discharge rate:
lodine released in cell:

Noble gases released in cell:

Solids released in cell:

Activity in Building with Fan On

1% per day or 0.0004 per hr
15,000 £t?/min or 1.875 building volumes per hr

2.5 % 10% x 10% x 50% = 1.25 x 10° curies
3.75 % 10° curies

Solids reledsed through filters:

6.8 X 10° x 10% = 6.8 x 10° curies

0.1% of solids to filters

 

 

 

 

. Fraction of Todine Noble Gases Solids
Time . .

After  AcbIvity Re- . . .
Accident leased from Release Activity in Release Activity in Release Activity in
(hr) Building Rate Building Rate Building Rate Building

per Day (curies/day) (curies) (curies/day) (curies) (curies/day) (curies)
0.25 3.5 x 103 436 9.7 1308 29.1 2.4 53
0.50 5.4 x 1073 675 15 2025 45 3.7 82
0.75 6.4 x 1073 800 17.8 2400 53.4 e & 97
1.0 6.8 x 1073 855 19.0 2565 57 4.6 103
1.1 7.0 x 1073 874 19.4 2622 58.2 4.8 105
1.2 6.8 x 1073 855 19.0 2565 57 .6 103
1.5 6.6 x 1073 828 18.4 2484 55,2 4.5 100
2.0 5.8 x 1072 729 16.2 2187 48.6 3.9 88
2.5 4,75 x 1073 594 13.2 1782 39.6 3.2 72
3.0 3.6 x 1073 455 10.1 1365 30.3 2.5 55
3.5 1.4 x 1073 176 6.8 528 20.4 1.0 37
4.0 1.3 x 10-3 160 3.55 480 10.65 0.9 19.3
5.0 1.9 x 1074 24,1 0.535 72 1.60 0.13 2.9
6.0 3.0 x 1072 3.8 0.0845 11.4 0.24 0.02 0.46
7.0 4.7 x 107 0.6 0.0130 1.8 0.039 0.003 0.071
8.0 7.2 x 1077 0.1 0.0020 0.3 0.006 0.0005 0.011

 

£9¢
Table A.2. Activity in Building with Fan Off

Cell leak rate: 1% per day or 0.0004 per hr

Building leak rate: 10% per day or 0.004 per hr

Todine released in cell: 2.5 x 10% X 10% X 50% = 1.25 x 10° curies
Noble gases released in cell: 3.75 x 10° curies

Solids released in cell: 6.8 x 10% x 10% = 6.8 x 10° curies

 

 

 

) Fraction of Todine Noble Gases Solids
Time e
After Activity Re- . . .o . R
Accident leased from Relcasc Activity in Release Activity in Release Activity in
(hr) Building Rate Bullding Rate Building Rate Building
i per Day (curies/day) (curies) (curies/day)  (curies) (curies/day)  (curies)
0.25 9.7 % 107° 1.21 12.1 3.63 36.3 6.58 65.8
0.50 1.86 x 1072 2.33 23.3 7.0 70 12,7 127
0.75 2.72 x 1072 3.39 33.9 10.2 102 18.4 184
1.0 3.5 x 1077 4 .37 43,7 13.1 131 23.8 238
1.5 4.9 x 10-? 6.07 60.7 18,2 182 33.0 330
2.0 6.0 x 107° 7. 48 . & 22 .4 224 40.6 406
2.5 6.8 x 1077 8. 54 85.4 25.6 256 46 . 4 46l
3.0 7.4 x 1072 9.30 93.0 27.9 279 50.5 505
3.5 7.6 x 1077 9.49 94.9 28.5 285 51.6 516
4.0 7.9 X 1077 9.88 98.8 29.6 296 53,7 537
5.0 7.86 x 1072 9. 8% 98 .4 29.5 295 53.5 535
8.0 7.8 x 1072 9,71 97.1 29.1 291 52.8 528
10.0 7.7 x 1073 9.63 96,3 28.9 289 52.% 524
16.0 7.5 x 1072 9.40 9%, 0 28.2 282 51.1 511
24,0 7.25 x 1077 9.08 90.8 27.2 272 49.4 494,

 

79¢
265

where
E = internal exposure in uc,
¢ = activity in building in curies,
V = building volume in m° (13.6 x 10°),
B = breathing rate in r’/hr (1.8),
t = time in hr.

The exposures and escape times are given below for the noble gases, iodine,

and fission-product solids.

Noble Gases

The noble gas escape time 1s based on an external exposure of 25 r.
Using the formula r/hr = 935 X pc/em® X Mev (assume = 1), the radiation
level is calculated and the exposure 1s obtained by integration. With
the fan on, k¥ = 1.875, and the concentration of noble gases in the build-
ing in pc/cm? is

20.6

—————— (1 - 728750 g 2018)
13.6 x 10°

The radiation level in r/hr is
6.24(1 — ¢™1-8750 _ g,221t) .

Integrating the radiation level with respect to time, the exposure in

roentgens is

e~1.875% t
6.24 |t + Té,.?T— - O.llO5t2 0 s

and the exposure at t hours after the accident in roentgens is

e-1.875% _ 4

6.24 [% + TR

- 0. llO5t2:|

With the fan off, k = 0,004, and the concentration of noble gases
in the building in pc/cm’ is
260

2.385 X 10°
13.6 x 10°

(L — ¢-0.004% . 0,003937%)

The radiation level in r/hr is
1.64 x 10° (1 — ¢70-004% — 0,003937t)

Integrating the radiation level with respect to time, the exXposure in
roentgens 1s
t

- 0.0019685t5] ,
0

e~0. 004t

. 5 1.
1.64 x 10 [J + 5 00%

and the exposure at t hcurs after the accident in roentgens is

5 . e-0.00é'fiZ -1 5
1.64 » 10 T+ ————-,T-fi'éz——— — 0.0019685t

Todine

The exposure per inhaled mwicrccurie of lodine is given in Teble

A.3.Y Based on Table A.3, the permissible tolal iodine exposure, that

ig, the exposure for a 300-rem dose, is

300,000
312.¢6

= 960 pe

With the fan on,

t
132 x 30.2 j; (1 — e~1-875% _ §.2205t) dt

: e-—-l.875t -1 2
= 3980 |t + ——1——5-75——— - 0.11%

il

Exposure

 

7, J. Burnctt, "Reactors, Hazard vs Power Level," Nucl. Sci. Fng.,
2: 382 {August 1956).

 
267

Table A.3. Iodine Exposures

 

 

 

Ratioc of Activity Exposure Dose per
Todine of Isotope to per Unit of
Tsotope Total Iodine Unit of Total Iodine
Activity Activity Activity
(L-hr decay) (mrem/uc) (mrem/ue )
pisl 0.115 1484 170.8
I3 0.170 53 9.0
Iti23 0.249 399 99.4
i34 0.244 25 6.1
7133 0.222 123 27.3
Total 312.6

 

With the fan off,

t
Exposure = 132 X 7.95 x 10° J; (1 — e~0.004T _ 3 937 x 10734%) at
-0,004t
= 8 e ~ _—31_ 2
= 1.05 X 10 (t + 500 0.0019685t ) .
Sclids

The maximum permissible inhalation intake (MPI;) of solids for a
25-rem bone exposure is 221 uc based on an exposure of 113 mrem per in-
haled microcurie of fuel irradiated for one full-power year.1:2 With the

fan on,

t
Pxposure = 132 x 164.2 [ (1 —e™>-875% _ 0,2205¢) at
0

e-la 875t — l
1.875

|

2.17 x 10% (t + ~ O.llt2>

 

2Third Draft, "American Standard for Radiation Protection at Reactor
Facilities," by ASA Subcommittee N-7.5 (June 1963).
268
With the fan off,

t

Exposure = 132 X 4.32 x 10° j; (1 — e70-004t _ 3 937 x 1072%) dt

i

6“0.004—t -1

5.71 X 108 ("G + W — 1,97 X lO—Btz)

Calculation of Effect of Meteorological Conditions
on Diffusion of Activity

Calculations of the diffusion of activity from the stack were made

using Sutton's formula for diffusion from a continuous point source:>s %

2
x:%em(_m)}

#C_C pxt” cex®7H
Yz

Q = emission rate, curies/sec,

= crosswind diffusion coefficient, (meters)?/2,
C, = vertical diffusion coefficient, (meters)nlz,
U = wind speed, meters/sec,

¥ = downwind distance, meters,

y = crosswind distance, meters,

n = height of plume, meters,

n = stability parameter,

= activity concentration, pc/cm’, for unit emission.

The following parameter values were used as recommended by the U.S.
Weather Buresau, Oak Ridge office:

 

37U.S. Department of Commerce, Weather Bureau, 'Meteorology and Atomic
Fnergy," USAEC Report ARCU-3066, July 1955.

7. J. Di Nunno et al., 'Calculation of Distance Factors for Power
and Test Reactor Sites,” p. 5, USARC Report TID-14844, March 23, 1962.
269

 

 

Value Under Lapse Value Under Inversion
Parameter (Weak) Conditions Conditions
o 2.3 1.5
n 0.23 0.35
Cy 0.3 0.3
Cy 0.3 0.033

All calculations were made assuming y = O; that is, the dispersion
was directly downwind of the source. Activity levels from building seep-
age were made assuming h = 0. The values obtained for X at various dis-

tances for both inversion and normal conditions are listed in Table A.4.

Table A.4. Diffusion Pactor, X, Versus Distance Downwind

 

%, Diffusion Factor®

 

 

 

Distance
Downwind Inversion Conditions Normal Conditions
(m)
Ground Release Stack Relesgse Ground Relesase Stack Release
100 2.1 x 1072 <10~10 8.0 x 107 2.9 x 1077
160 9.8 x 10~3 <10710 3.5 x 107% 1.1 x 1077
200 6.3 x 1073 <10710 2.3 x 1074 2.3 x 1077
325 3.0 x 1073 <10-19 1.0 x 1074 3.7 x 107°
500 1.5 x 1073 2.1 x 1077 4.7 x 1077 3.0 x 107°
750 7.7 x 1074 3.2 x 107° 2.3 x 1077 1.8 x 1072
1,000 4.8 x 1074 3.5 x 1077 1.4 x 1077 1.2 x 107?
1,200 3.5 % 1074 4.5 % 1077 1.0 x 1077 9.1 x 107°
2,000 1.5 x 1074 6.2 x 1077 4.0 % 107° 3.8 x 107°
3,000 7.7 x 107 4.9 x 1077 2.0 x 107° 2.0 X 107°
4,000 4.8 X 107° 3.6 x 1077 1.2 x 107° 1.2 x 107°
5,000 3.3 x 107° 2.7 x 1077 8.0 x 1077 8.0 x 1077
10, 000 1.1 % 107° 1.0 x 1076 2.3 x 1077 2.3 x 1077

 

% X release rate (curies/sec) = activity concentration in air

(pe/ce).

Calculation of Dose Rates Outside Building

 

Dose Rate from Building

 

The building is assumed To be a point source with radiation of 1l-Mev

average energy E. The dose rate R (in r/hr) at 1 ft is given by the
270

formula’
R = 6CE ,

were ¢ 1s activity concentration in uc/cm?. With the fan on the maximum
activity in the building is 183 curies at 1.1 hr. With the fan off the
maximum activity in the building i1s 932 curies at 4 hr after the maximum
credible accident (see Tables A.l and A.2). The radiation level is as-

sumed to vary inversely with the square of the distance.

Dose Rate from Radicactive Cloud

The maximum release rate with the fan on is 3501 curies per day at
1.1 hr after the maximum credible accident (see Tables A.l and A.2). With
the fan off the maximum release rate is 93.2 curies per day at 4 hr after
the maximum credible accident. The dose rate R (in r/hr) is calculated

from
R = 935 cE .

The release rate in curies per second multipiied by the diffusion factor,

X, from Table A.4 is the activity concentration in uc/cm?.

Calculation of Total Integrated Dose

 

The total integrated thyroid snd bone doses are calculated for stack

3 4

release” and for building seepsage.

Total Activity Released

 

For stack release with the fan on the activity released from the

building in the first 4 hr is

t
j‘ (curies in building at © hours) X (release rate per hour) .
0

 

Radiation Safety and Control Manual, p. 3-3, Ozk Ridge National
Laboratory, June 1, 196l.

 
271

For iodine (see p. 266) the release in 4 hr is

4 |
30.2 [ (1= e 8750 ~ 0.2205¢) at X 1.875 = 96.5 curies

After 4 hr, no additional iodine leasks into the building from the cell,
and the 3.55 curies in the building at 4 hr (Table A.1l) is the only ad-
ditional release after 4 hr. The total iodine release with the fan on
is therefore 100 curies.

For solids the release in 4 hr is

4
164.2 j; (2 — e=1-875% 0 2205t) at

X 1.875 x 0.001L (99.9% retained on filters) = 0.53 curies s

to which 0.1% of the 19.3 curies in the building at 4 hr is added to give
a total solids release of 0.55 curies.

For release by building seepage with the fan off the activities are

 

calculated in the same manner.

Tor iodine the release in the first 4 hr is

&
7.95 X 10° J; (1 — e70-004% _ 3 937 % 1073%) at x 0.004 = 1.06 curies

From Table A.2 it 1s seen that 98.8 curies of iodine are in the building
at 4 hr, and it 1s assumed that this is all. released, although a large
vrercentage will undoubtedly plate out on the building walls. The total
iodine release is therefore ~100 curies.

For solids the release in the first 4 hr is
t
4.32 x 10® x 0.004 [ (1 —e70:004% _ 3,937 x 1072¢) dt = 5.43 curies .
0

From Table A.2 it is seen that 537 curies of solids are in the building

at 4 hr, and therefore the total release is 543 curies.

Maximum Permissible Doses

 

Thyroid: 300 rem (ref. 4); 1484 mrem per microcurie of I131,

Bone: 25 rem (ref. 4); 113 mrem per microcurie of solids.
272

These doses are based on an irradiation time of 365 days, using re-

vised exposure values representing current estimates of dose per inhaled

component microcurie.

1,2

Stack Release

For

For

For

For

For

2@ X breathing rate

 

 

TIDmax T X e X wind speed X stack height
iodine,
-4
2 X 100 x 5 x 10 = 2.04 x 107° curies
3.14 X 2.72 X 2.3 x (50)?
normal atmospheric conditions,

2,04 X 107¢ X 1.484 X 10° = 3 rem .

inversion conditions,

 

3 x 242 = 4.6 rem
1.5
solids,
) Q=%
2 x 0.55 X 5 X 10 = 1.12 x 1078 curies
3.14 x 2,72 x 2.3 X (50)°
normal atmospheric conditions,

1.12 x 1078 x 0.113 x 102 = 1.3 mrenm .

inversion conditions,

1.3 X 24% = 2 mrem .

i_J
273

Building Seepage

 

Q X breathing rate
T X wind speed X oy X Oz

 

Inhaled curies =

Q = total curies released from building
Breathing rate = 5 x 10~% m?/sec
Wind speed (normal) = 2.3 m/sec

Wind speed (inversion) = 1.5 m/sec

1
o, (vertical concentration deviation) = — C d(l—n)/z

2

| L (1-n)/2

o, (horizontal concentration deviation) = — C,d

V2
Cy (vertical diffusion coefficient) = 0.3 normal

= 0.3 inversion

C, (horizontal diffusion coefficient) = 0.3 normal

0.033 inversion

Table A.5., Doses Resulting from Building Seepage

 

Quantity of Activity and Dose

 

 

 

 

Distance
Lrom Activity  Normal Conditions  Inversion Conditions
Building
(m)
Le rem e rem
1.00 Todine L 6 66.3 1075 1600
100 Solids 242 27 .4 5840 660
200 Iodine 13.2 19.5 348 517
200 Solids 13.8 1.56 405 45.8
1000 Iodine 0.75 1.12 24,1 35.8

1000 Solids 4,07 0.46 131 14.8

 
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Flow Sheet‘;

No.

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40881 .
40882

40883

40884

40885

40887
40888
40889

275

Appendix B

PROCESS FLOWSHEETS

Procéés flow sheets follow for the systems listed below:

7

System

 

. Fuel System
Coolant System
| Fuel- Drain Tank System

Off Gas System and Containment
- Ventilation ,

Cover Gas SyStem

011 Systems for Fuel and‘CQQl-
ant Pumps o

Fuel-Précessing System
Liquid Waste System
Cooling-Water System

O

 

 

 
 

Unclassified
ORN'L-DJG 63=7'752

 

276

 

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TO WPOR
CONDENSN
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NORTH. ENO OF BLDG.

USED)
COOLING TOWER
PUMP NO.1.

EXISTING

 

TREATED WATER

PUMP NO. 2.

WATER ROOM

PROCESSING SYSTEM
GAS SYSTEW & COMTAINMENT VENTLATION
SYSTEM INSTR, APPLICATION DMGRAM

SYSTEM FOR FUEL & COOLANT PuwPs

£, DRAN TANK SYSTEM

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HS TTYNH3HL H0LOV3H OL

TREATED WATER PUMP DATA

 

COOLING TOWER PUMP DATA

PROCESS FLOW SHEET

O-AA-A-40800-8

JOB 433-1,0

 

 
277

 

 

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ORNL~-IMG 63-7751 ,

 

FLTER nr{

SERVICE TUNNEL

 

 

’ .

 

VENTILATION STACK — e !2)——_.
’ -

“ln puct fLOOR"‘"‘”@"—-—“_—-

RADIATOR STACK

COOLANT DRAIN CE |- (352~
(3550

SEMVICE ROOM (£358) '
SERVICE TUNNEL—(D59) -

 

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LIQUID WASTE CELL

 

 

 

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TANK SYSTEM.

 

PUEL PROCESIMS SYSTEM
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LIGUID WISTE SYSTEM INST. APPLICATION DiA.  [D-A-B-40308

 

 

 

OFF A% SYSTEM B CONTARMENT VeNT
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&
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ABSORBER DATA AREA OUTSIDE 7303 WEST OF DTC

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278

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ORNL-WG 63="7750

40 PSIG HELIUM
T 40-35S FROM LINE SO0

TRAILER CONTAOL N WATER ROOM

’VINT TO
ATMOS.

}
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AREA WEST OF FPC

FUEL STORAGE TANK DATA
FUEL
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BLOWER HOUSE

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816 FUEL LOADING & STORASE SYSTEW
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VALVE WDICATION FOR
NORMAL OPERATION OF
PARTICULAR EQUIPMENT

.
S

OAMPER (0. G

FUEL H!DCESS!NG CELL I
S SRS SR S GBS SE——

NERSCE Shamnes ' B, W
Oalt et NATIONAL LABORATORY

FUEL PROCESSING SYSTEM
PROCESS FLOW SHEET

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- LUBRICATING OiL. I §-40- n:s)
- . | . — s )
y-40 -8 {
* . | D' BREATHER coMcmn-—-b 20
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b pe - 2 = = Go>————]
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‘ tE SYSTEM ONLY
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b . . .
HELIUM —E35> 535 -
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pi . . n SR - g':’ec;oso‘r.o
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OIL SUPPLY TANK DATA - - !I ' ®
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COOLNG TOWER o ‘37405 ve 8~} QUICK DISCONNECT CONMECTION e NOTES:
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\ . . ‘ R S | I..FDUEELTAGND mc ctn'ntr 'ruur OIL SYSTEMS AME
l . IDENTICAL EX OR LINE AND EQUIPMENT
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R 2 MATEMAL CAROON 3T . SHOWN 1N PARENTHESIS. o
it TYP CyuNo EF ' 2. ALL LEAK DETECTION FOR LUBE OIL SYSTEM
' i CARTRIDGE re | IS LOCAL. TUBING IS NOT CONNECTED TO THE
COOLING TOWER £ REMOTE LD SYSTEM IN THE TRANSMITTER
WATER IN i o ROOM. ,
g ,-
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COOLING WATER SYSTEM - D-AA-A-4Q8 89
. . SYSTEM FOR FUEL PUMP INSTRUMENT
APPLICATION [WAGRAM | D-AR-2-40304
E COOLANT SYSTEM D-AA-A-40881 |
\ - frus svsrew oax-s:40090
_L . . nerEnancs BRAWGS . we
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L Y v ] TE‘E.‘H--'-'&- OAR RIDGE NATIONAL LASORATORY )
2o, we * oo TR £ s LaNT PUMPS
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‘ ' fpulm |1 Y, s : M.S.RE
1481 EE_OCN Fedi /e & ——
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280

 

Unclassified
ORNL~DWG 63-7748

 

CYLINCERS

EMERGENCY
HELIUM
SUPPLY

 

6-?00 5T, CLF

 

 

iT-oes-88

 

@ soorsig (1D \' .1 1258 -
89 17

w }-00-88

 

 

NORTH OF
DIESEL’ HOUSE

FRESH HELIUM SUPPLY
. (TRALER)

9070 STD. CL FT.
2400 PS)

T0°F 0
30-9f# x 2 crumoEns

 

 
    

 

 

 

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$1-0.083-33

  
   
  
 

 

 

 

 

 

 

 

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TTER ROOM r a— 4 |
' HIGH BAY AREA

 

SERVICE
AREA

'—— _‘..; —

  
 

    

 

SYSTEM

BUBBLER

 

 

 

 

 

 

 

     
  

 

 

 

 

 

 
    
 

NOTES:

i
1~ALL EQUIPMENT EXCEPT HELIUM TRAILER IS LOCATED N

DIESEL HOUSE

I ..4
, L4

WES YO COOLANT]
TANK

  
 

 

 

 

 

 

 

 

 

 

 

 

AL ARRANCEMENT — W & L€ COvER
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FUEL PROCESSING SYSTEM MK~

OIL SYSTEM FOR FUEL & ¢ OLANT PuvPs  loaa-a-408383
£ SY5T CONTAINMENT VEN DAMA

DRAIN_TAWK SYSTEM -AA-Ar 4 Q

CNOLANT SYSTEM D-AM-#-40

FUEL SYSTEM D-AA-A-40880
OVER GAS SYSTEM {NSTR. APPLICATION OiAGRAM l0-AAB-4030

REFERENCE BRAWHES [——

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

¥ - S e OAK RIDSE NATIONAL LABORATORY
o e S - COVER GAS SYSTEM
foc % - —r PROCESS FLOW SHEET
H - FRbln (Y |20 _M.S.RE. :
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281

  
   
 
    

Unclassified
ORNL-IMG 63=7747

 
 

FROM COOLANT
& OiL SYSTEM

 

 

VENT HOUSE

) CNNLANT DRAIN CELL
S SIS— SIS SIS

CHARCOAL BED CELL

    
 

STACK FAN NO.t
21,000 CFM AT
107 w0

PONT SF I

 
  

=40

1

 
   

AUXILIARY
CHARCOAI, BED

   

   
  
     
 
    
  
  
 
    
  
   
 
 
      
 
  
   
     
    
  
   
    

   

CHARCOAL BED NO. 1A

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CHARCOAL \ BED NO.28

    

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CHARCOAL BED DATA -

 
 
 

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t

 

INFILTRATION
870 CFM MAX.
e

OJ TO 0.3 WATER PRESSURE

SPECIAL
EQUIPMENT

ELECTRICAL
SERVICE AREAS

|
REACTOR

CELL

 
   
 
      
   

 
     

   

   

 

-2.0 PSIG

 
 
  

EQUIPMENT
STORAGE CELL
SFARE CELL:

ELFL PRy CEuoiNG

MAINTENANCE
PRACTICE
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CELL

 
 

TUNNEL LEARASE

 
  

110N CFM

    

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- 1500 CFM EACH DURIN
AR TERANCE ! . . WATER ROOM

  
 
 

INLET AIR
FILTER
HOUSE

  
 
   
 

SYSTEM MST. APPLICATION DHAGRAM

 

HIGH BAY AREA ORAIN TAMK SYSTEM

SYSTEM

 

      
 
  
    

CAx TDOE NATIONAL LABORATORY

NFF GAS SYSTEM 8 CONTAINMENT VENT.
PROCESS FLOW SHEET

 
   
   

16,000 CFM (MAX)

   

Jom 43340 08 3-8

 

 

 
 
  
  
  
 
     

   
 
     
 
 

 
 

 

 
 
 

   

2w st e e ¢

 
 

 

»

282 .

   

Unclassified
ORNL~-DWG 63-7746

FUEL PuMP TO AUXILIARY
: CHARCOAL. BED

-} -40-38

 
 
 

  
 
 

 
 

DRAIN TANK CELL {~0-38

  
  
  
 
 
 

Ig-40-1

 

}un_i | 40-3

 

 
 
 

      
 
   
  

 

 
   
  
 
 
 

    
  

  
 
  
 
 
 

 
 
  
 
 

  
   
    
 

   
 
  
      
   
 

 

   

 

 
   

 

 

 

      
 

 

 
 

 

 
  
     
  
 

 
 
 
  
 

 
    
 
       
  
  

  
  
   
         

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| T e e B, |
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&
: W
: TANK DATA
| N
l ,
- ) ELECTRICAL SERVICE AREA
DRAIN TANK CELL
ERLLnad wo.
' ‘ 'DRAIN TANK CONDENSER DATA r
304 l
I I DRAIN TANK DATA
. | .
| I LEGEND
:::l. o
I ==~ NORMAL L/M
VALVE POSITIONS ARE FOR NORMAL
REACTOR OPLRATION
' I (H) 13 L HEATER cONTROL
CIRCUIT
y
l 70 940 LAVEL MALL ,_@1—:-\
.‘ : AND DUCT 937 8 ¢
| ' I pio [~ —— ELECTRICAL SERVICE AREA
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LINE ‘ REPERENCE SRAVNSS
801 OAR IDSE MATIONAL
FUEL. DRAIN TANK SYSTEM ( ]
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|

TO FUEL
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FREEZE WALVES

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28

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Unclassified
ORNL-DWG 637745

RADIATOR DATA
INOR~-8

,
|

COOLANT CELL ¢

1
T et

LEGEND

~—=pP8IG

— NORMAL ¢/M

M = e s » -

HOUSE

@ 13 1 HEATER CONTROL GIRCUT

PROCESSING SYSTEM

COOLANT SALT SYSTEM INSTRUMENT
APPLICATICN DIAGRAM

MOLTEN SALT REACTOR EXPERIMENT

COOLANT SYSTEM
PROCESS FLOW SHEET
REACTOR DIVISION

 

RBM.J\ {SJ-) "‘y"fl’

JOB 433+4.0 I L N-AA-A-ADRAL~ R
 

   

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9

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DRAIN 1ANK CELL 3

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TANK SYSTEM

 

—
SAPECIAL EQUIPMENT ROOM
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S

CONTAIKMENT ENCLOSURE KO.2
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o K !

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NUCLEAR
INETRUME
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18 FLANGE PAIRS
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THERMAL SHIELD

AUTO-RESISTANCE
HEATING LUG

-

 

 

284

TO ENRICHER SAMPLER

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A
2

 

Unclassified

REVISIONS

SEE DCN 2491
SEE OCN 268)

. e eyt ol o i ol
CC ot on, SPECIAL EQUIPMENT ROOM

I V7204
vr208
WORH
; WASTE

oIL
RQCEIVER

ORNL~DWG 63-7744 .

TC OFF
SYSTEM

 

COOLANT
DRAIN

. CELL

39 GAL

HEAT EXCHANGER
HX

REACTOR
CFLL

@ =L EA% DETECTON SEAVES 2 FLANGES
THROUGH CONNECTING L TUBING

(!Ill-—n:umt GAS LINE DVSTONNECT

 

J 1025*F

Hoo*F (v

KOLTEN SA T “EACTOR EXTERIMVENT

FUEL SYSTEM

PhOCESS FLOW SHEET

.

QAN RIDGE NATIONAL LABGRATORY

FUTURE FLANGES
FE-200 & FFR201

BWO. o,

7803 !

“ows o
BRSION oF BN

JOB 433-1.C

3.
i-31.4%

“ e | D-aA-A-40280

 

 
285

Appendix C

COMPONENT DEVELOPMENT PROGRAM IN SUPPORT OF THE MSKRE

The reliable performance of components and auxiliary equipment used
in the circulation of molten salts was established in over 200,000 hr of
accumulated loop operations and formed the basis for the specification
of components for the MSRE. As added assurance of reliability and safety,
prototypes of critical MSRE components were operated out-of-pile under
conditions resembling those of the reactor. Facilities for this testing
included salt systems of various sizes and different degrees of complexity

and model tests in which hydraulic and mechanical processes were analyzed.

Core Flow

A one-Tifth scale model of the MSRE core was operated with water as
an ald in establishing the design. Variations of several design features
were investigated in the model before the design was completed. Details
of the entrance plenum were studied to assure even circumferential flow
distribution into the core-shell cooling annulus, and baffles to kill the
swirl produced by the entrance conditions in the lower vessel head were
tested. Preliminary measurements were made of the heat transfer coeffi-
clent at the surface of the lower head. Upon completion of the design,

a full-scale model was constructed and operated with 1250 gpm of water

to verify the adequacy of the flow distribution in every section of the
core vessel assembly. As a result of these tests, minor modifications
were made to the graphite support lattice to correct an uneven flow dis-
tribution found between mutually perpendicular chamnels in the graphite.
Tests were also conducted with various sizes and densities of solid parti-
cles introduced into the inlet stream to determine the probable distribu-
tion in the system. It was found that particles below about 300 P passed
through the core vessel adequately; however, particles over 300 p tended
to settle out in the lower head and to accumulate around the drain pipe
at the vessel center line. The design of this drain pipe was changed to

reduce the possibility of plugging.
286

A Tew tests were conducted while using a thickening agent in the
water to wverify the analytical translation of the results with water to
the galt conditions. The results of these tests indicate that the flow
distribution within the core vessel assembly will provide adequate cooling
to all parts of the assembly.

The full-scale core hydraulic model was operated with an analog simu-
lator to study the effect of system pressure oscillations on core reac-
tivity. It was found that power oscillations of less than 1/2% would .
result if the questionable assumption were made that the pressure oOscil-
lations in the core were similar to those in the model. Additional studies
are planned in this area.

The possible interaction of fuel salt and full-size core graphite
stringers was investigated in the S-in.-diam vessel that is part of the
Engineering Test Loop (& faclility for testing many MSRE prototype compo-
nents and overating vrocedures,. The flow conditions in the channels were
similar to those in the MSRE, and after 5000 hr of operation at 120C°F

there were no significant changes in the appearance of the graphite.

Ileat Exchangers

 

Heat exchangers as complex as the MSRE exchanger were fabricated in
Inconel for the ARE and the ART. The fabricebility of INOR-& into tubes
and tube-to-tube sheet assemblies was demonstrated. The MSRE heat exchanger
and radiator sassemblies, as now designed, do not reguire high-temperature

y The prenuclear operation of the entire

3

develorment beyond that plamned 1
system. However, uncertainties iz the fluid flow design required that
hydraulic tests be conducted on the heat exchanger. An excesslve pressure
drop was FTound on the fuel side of the tuve bundle and alterations were
mede to the shell in both the entrance and exit regions to reduce the loss
in pressure. Vibrations of a significant amplitude were encountered during
the tests, and this condition was corrected by "tightening' the tubes by
the insertion of appropriate spacers.

The air-cooled radiator was operated at full air flow without detect-

gble vibration.
287

Freeze Flagges

Freeze flanges have been a part of molten-salt engineering develop-
ment for several years. They were found to be satisfactory in a large
circulating salt system — the Remote Maintenance Demonstration Facility.
It was shown that the flanges could withstand repeated thermal cycling
and the joint could be broken and reassembled. Twoc flange pairs with
integral cooling were cycled between room temperature and 1300°F for more
than 100 times without failure., Salt leakage from a freeze flange has
not been experienced to date, although leakage of helium through the secon-
dary gas buffer seal was encountered in the thermal-cycling tests.

As a result of these preliminary tests, the design of the flanges
was improved to provide greater strength toward axial loads, to reduce the
thermal stress, to improve the tightness of the buffer seal, and to re-
move the requirement for integral cooling. Flanges of the improved design
were thermal cycled between room temperature and 1300°F for more than 100
times, during which the helium legkage from the buffer zone remained less
than 5 X 10™% cc (STP)/sec and without any significant dimensional changes
that would indicate excessive thermal stress. These flanges were also
subjected to severe thermal shock without damage. It was found that a
new flange would makeup and seal satisfactorily with a flange that had
experienced several room temperature-to-1300°F thermal cycles. Under
conditions of pump stoppage, it was found that the salt in the bore of
the flange would not freeze solid if the temperature of the pipe on either
side of the flange was maintained above 1000°F. Eguipment has been de-
vised for the remote operation of the clamps, for the replacement of the
ring gasket, and for sealing the open pipe during the maintenance period.

In addition to the work on freeze flange disconnects, small, remotely
maintainable, buffered seal disconnects were developed for use in the
sampler-enricher and in the various gas lines of the MSRE. These discon-
nects were operated to satisfactorily demonstrate their reliability under

extreme temperature conditions of service.
288

Freeze Valves

Since no reliable mechanical valve was avallable at the start of the
project, it was decided to use frozen plugs of salt to control the trans-
fer of salt. Development of these "valves" produced a unit in which it
is posgsible to establish and maintain reliably either the open or closed
condition. For the normally open valve, approximately 10 min is required
to thaw the frozen valve, even under the conditions of a total power fail-
ure. The normally closed valves, with the requirement that they remain
closed during a power failure, are allowed to cool to ambient temperature,
thereby insuring that there is not enough energy to open them, except on
demand. The time required to open this type of valve is in excess of 1 hr.
Both types of valves require an average of 25 min to freeze.

The freeze valves to be used in the reactor system were designed to
ensure that the thawing operation would proceed from the molten zones at
the ends of the valve to the frozen zone at the center. The purpcose of
this feature is to prevent expansion damage to the valve as a result of
thawing at the center of an excessively long frozen zone. Prototype valves
of this design have been operated through more than 100 freeze-thaw cycles

without dimensional cheange.

Heaters

Prototype heaters for pipes and vessels have been fabricated and
tested for many thousand hours of successful operation. The heaters for
the pipes are enclosed in all-metal reflective insulation, which was chosen
because of the dusting problem normally encountered with refractory ma-
terials where much handling is required. Heater-insulation units of this
type have operated with the interior at 1200°F for more than 3200 hr with-
out apparent changes in the heat loss of 600 W/lin ft. The heaters are
connected in such a way as to reduce the effect of the failure of any one
heater on the temperature distribution along the pipe. Additional confi-
dence in the heater design was obtained when one of the heaters shorted
to ground and shut itself down without any detectable damage to the empty

5-in. pipe.
289

A prototype of a single heater for the drain tank has operated at
1200°F over 10,000 hr without difficulty. The failure at the beginning
of the run of an interconnecting lead wire was attributed to inadequate
inspection of a weld. The design and inspection of the weld was improved
and no further trouble was encountered.

A mockup of a drain taank cooler for removing afterheat from the fuel
was operated through the equivalent of 1000 startups without any indica-
tion of damage to the INOR-8 tubes. Previous tests on a system of a
slightly different design, constructed of Inconel, had caused failure of
the tubes after only 250 startups, The results of the tests indicate that
there should be no trouble with these cooler tubes through at least 100
reactor shutdowns from full power.

Power tests on the heater rod to be used in the fuel pump furnace in-
dicated that less than 50% of the rated capacity of the heater would be
transferred if the rod surface temperature was to be held to 1400°F. As
a result, heaters were added to the furnace to prevent overloading of the
individual units and to ensure a longer life.

A mockup of a section comprising 15% of the heaters for the reactor
furnace was operated for 20,000 hr at 1250°F without incident. The ex-
perience with the test furnace was incorporated in the design of the re-
actor furnace.

All the potentially dusty refractory insulation material to be used
in the furnaces was examined by activation analysis to eliminate any that
would cause a contamination problem by spreading radioactive dust when

equipment is removed from the cells.

Control Rods

A prototype control rod drive was operated in a 150°F environment
for more than 60,000 cycles of 102-in. travel in each cycle. Several
combinations of materials were tested in a worm to worm-gear combination
to obtain satisfactory life. Numerous small changes were made to improve
the final design.

The prototype control rod was tested at 1200°F in conjunction with

operation of the drive test. Development of a pneumatic position indicator
290

for the in-pile end of the control rod was completed, and satisfactory

operation was demonstrated.

Xenon Migration

 

Experiments designed to measure the xenon stripping efficiency in the
fuel pump bowl were conducted on a pump mockup using CO; and water, and
preliminary results indicated that the stripping efficiency could be as
low as 25% under normal operating conditions. Using this value of 25%,

a study was made of the xenon distribution expected in the MSRE system,
and it was found that from 20 to 70% of the xenon could migrate to the
graphite of the core, depending on the choice of values for the xenon
mass transport through the salt and the xenon diffusion in the graphite.
Additional studies of the xenon migration will be made on the MSRE system

during precritical operation with salt.

Engineering Test Loop

 

The engineering test loop, a collection of simulated MSRE components,
was operated for extended periods without serious difficulty. The oper-
ability of freeze valves, level indicators, and gas-handling systems was
demonstrated under operating conditions. A method was tested for removing
oxide from the salt with HF and H, treatment in the drain tank. The locop
was altered later to include a large container for full-size graphite
specimens and was operated with an LiF-BeF, salt for extended periods with-
out difficulty. The adequacy of a method of gas purging and salt flushing
of the system to remove oxide before operation with fuel was demonstrated.
The operation of a frozen-salt joint, which provides access to the graph-
ite in the loop, was successfully demonstrated, and a procedure was pre-
pared for operating 2 similar Jjoint for access to the graphite samples in
the MSRE.

The loop was operated with LiF-BeF,-ZrF,-UF, salt in excess of 5000
hr, mainly for testing the sampler-enricher mockup and following the in-
ventory of uranium added to the loop. The uranium concentration was varied

from 0.1 mole % at the beginning of the period to about 0.7 mole % at the
291

end of the period. Metal and graphite samples were removed periodically

for examination, and no unusual changes were noticed.

Helium Purification

 

A full-scale model of the oxygen-removal unit was tested under maxi-
mum flow conditions (10 liters/min) with varying inlet oxygen concentra-
tions and bed temperatures. The results indicated that, with a bed tem-
perature of 1200°F and an inlet oxygen concentration of 100 ppm, an outlet
concentration of <l ppm of oxygen was maintained until 58% of the titanium
sponge had been consumed. The average oxygen concentration at the inlet
to the purifier for the MSRE is not expected to exceed 10 ppm, hence an
average bed life of 900 days may be predicted for the oxygen-removal unit.

The performance of a commercial analyzer for the continuous detection
of moisture and oxygen in the helium supply was evaluated and found to be

satisfactory for use with the MSRE.

sampler-Enricher

 

The sampler-enricher mechanism was mocked up, first in part, to evalu-
ate critical areas for design, and then completely, to test the assembly
of parts as a system. The sampler-enricher mockup was tested in conjunc-
tion with the engineering test loop during a period of over 5000 hr of
operation, during which the uranium concentration was varied from O.1
mole % to about 0.7 mole % at the end of the period. The sampler-enricher
was used to successfully isolate over 50 samples and to make over 25 ura-
nium additions. Correlation of the chemical analysis of the sample with
the book inventory and with the results of other methods of sampling were
within an experimental error of 1 1/2%. The leakages through the buffer
seals in the valves, disconnects, and transport container remained within
an acceptable limit throughout the operation. The complete system for
sample transport from the sampler-enricher to the analytical chemistry
hot cell and for handling the sample in the hot cell was successfully
demonstrated. As a result of these studies, the designs of the samplers

for both the fuel and coolant salt system were finalized.
292

Maintenance

The practicality of the two general methods of maintenance designed
into the MSRE has been thoroughly demonstrated during the past 6 years.
Entirely remote maintenance, with the aid of stereo-televison, an over-
head crane, and a General Mills' manipulator, was demonstrated on a special
salt system (Remote Maintenance Demonstration Facility) of about the size
and degree of complexity of the MSRE. After this system was operated with
molten salt, the pump, the dummy core, the heat exchanger, and heaters
were removed and replaced with remote means. The system was shown to be
operable following the replacement.

Semidirect maintenance with long-handled tools operated through por-
table shielding has been used successfully on a number of operations at
Homogeneous Reactor Experiment No. 2, a reactor of the same general size
and activity level as the MORE.

Although the feasibility of maintenance was regarded as having been
established, additional development and practice was required to work out
detailed procedures. Because the MSRE was constructed in an existing
facility that was not readily adaptable to the manipulator type of remote
maintenance, the manipulator was not used in the MSRE. However, some of
the procedures include the use of the 'remotized" crane, the television
system, and viewing windows in carrying out entirely remotized operations.

Procedures, sufficiently detailed to be used by reactor personnel,
were written for maintenance of part of the components in the reactor and
drain tank cells. Design, fabrication, and testing was completed on over
20 individual long-handled tools. Television reguirements for the MSRE
were found to be better satisfied by a two-camera orthagonal system rather
than the stereo system previously studied because of better reliability,
less operator fatigue, and a simpler installation. All maintenance pro-
cedures were demonstrated during the construction and preoperational test-
ing of the reactor. Studies were made on methods of handling and dispos-
ing of the radiocactive components after removal from the reactor cell.

A one-sixth-scale model of the reactor system and maintenance area
was built as an aide in maintenance planning both before operation and

after cperations have begun.
293

Appendix D
CALCULATION OF ACTIVITY CONCENTRATIONS
RESULTING FROM MOST LIKELY ACCIDENT
Cell Volume
The total volume of the reactor and drain tank cells is

Reactor cell 11,300 £t3
Drain tank cell 6,700 ft°

Total 18,000 £t3 or 5 x 108 cc

lLeakage Rate of Cells

 

If the cell pressure is maintained at 13 psia after the most probable
accident, the inleakage for the lS,OOO—ft3 cell volume is a maximum of
0.43 t3/nr (ref. 1);

0.43 ft3/nr = 1.2 X 10% cc/hr .

Fraction of Total Fission-Product Inventory in 4 Liters of
Spilled Salt

 

4 liters
70 ££2 X 28.3 liters/ft?

=2 X 1073 ,

 

Total Fission Product Inventory for the Maximum Credible Accident
(see sec. 8,7.2)

 

Todine 2.5 X 10° curies
Solids 6.8 X 10° curies
Xenon and krypton 3.75 X 10° curies

Amount of Activity Released from 4 Liters of Spilled Salt

 

0.5 X 103 curies
1.4 X 102 curies

For 10% iodine: 2.5 X 10%® X 2 X 1073 X 0.10
For 10% solids: 6.8 X 10% x 2 x 10-2 X 0.10

"

The xenon and krypton can be neglected,

 

13, E. Beall and R. H. Guymon, "MSRE Design and Operations Report,
Part VI, Operating Limits," USAEC Report ORNL-TM-733, Oak Ridge National
Laboratory (to be issued).
29

Concentration in Cells

Todine: 0.5 X 10% + 5 X 10% = 107% curies/cc

Solids: 1.4 X 10% £ 5 %x 108 = 3 x 1078 curies/ce

The MPC, for a L0-hr/week exposure is 2 X 1072 uc/cc for iodine, and

thus the concentraticn of 1 uc/cc must be reduced by a factor of 108; for
the bone seeking solids, the factor is 1019,

For the iodine, the removal factors can be expected to be 104 for
the carbon bed, 10% for diluticn in the stack, and about 10° for disper-
sion in the atmosphere. Thus even without the charcocal bed, the lodine
concentration would be below the MPC,.

For the solids, the removal factors can be expected to be 10% at the
filters, 10 in the stack, and 10° in the atmosphere. These factors give

e total of 102, or 10° more thar needed to meet the MPCj.

Ceonclusion

The Most Likely Accident could be tclerated, even 1T the charcoal

bec and the absolute filters were not effective,
295

Appendix E

TIME REQUIRED FOR PRESSURE IN CONTAINMENT VESSEL TO
BE LOWERED TO ATMOSPHERIC PRESSURE

The heat content of the steam and gas in the containment vessel at
various times after the maximum credible accident has been calculated.?
It was assumed conservatively that the nitrogen was swept from the cell,
leaving pure steam at the meaximum pressure of 37 psia, and thus the maxi-
mum temperature would be 262°F. The rupture disk should break after about
4 sec, when the pressure approached 40 psia, and the discharge of steam
would be over in about 200 sec. In calculating the time required for the
vessel pressure to return to atmospheric, a graphical solution to the
problem of heating a semiinfinite slab was used, as suggested by McAdams. @
In the case of the MSRE the containment vessel is the "slab'" and the ves-
sel atmosphere is the heat source. It was assumed conservatively that‘

the slab was insulated by the sand-water mixture in the annulus.

Using McAdams' notation,

t/ —-ta
{= —, (1)
t —-tb
where
t/ = initial temperature of the cell atmosphere, °F,
tb = initial temperature of the metal wall, °F,
ta = final temperature of the metal wall, °F,
and

(2)

W K
T r U’
m

 

1R. B. Briggs, "MSRE Pressure-Suppression System," internal ORNL
document, MSR-61-135.

“W. H. McAdams, Heat Transmission, 2nd ed., McGraw-Hill, New York,
1942.

 
296

where
K = thermal conductivity of steel, 26 Btu/hr-ft? (°F/ft),
r, = distance from midpoint to surface, 1/6 Tt,
U = film coefficient of heat transfer, assumed to be 50 Btu/hroft2a°F,
and
Ké
X, "Relative Time" ratio, = — , (3)
oC r°m
p
where
€ = time from start, hr,
o = density of solid, 480 1b/ft> for steel,
C, = specific heat of solid, 0.12 Btu/lb:°F,

an estimate was made of the amount of vessel heating that would take place
during the time salt was spilling and the vapor-condensing system was 1in

action. Then, taking t. as the vessel temperature before the accident

(150°F) and t’/ as the mzximum steam temperature during the event (262°F),
the vessel temperature (ta} at the end of the accident (i.e., 6 = 300 sec)
was found. For the value X = 1.32, obtained by solving Eq. (3), the value
of Y was found from the X-Y plot of Fig. 10 of ref. 2 to be 0.7. Based
on these values, Eq. (1) was solved and t, was found to be about 180°F

at the beginning of the cool-down of the cell atmosphere.

As an approximation of the unsteady state of the steam temperature
during cooling, Y was calculated at 5°F intervals of wall temperature in-
crease., For these calculations, the value of X was taken from Fig. 10
of ref. 2. The heat content per pound of steam remaining in the cell was
then determined and the equivalent pressure was read from the Steam Tables.
The temperature and pressure of the cell obtained for each interval are
given in Table E.1.

Although the input of heat from fission-product decay is not taken
into account for this short period, it becomes significant (approximately
10% Btu) over the 4-hr period assumed in the release calculation. However,

with the salt on the floor of the cell and in contact with the steel, which
297

Table B.1. Cell Pressure at Various Times
After the Maximum Credible Accident

 

 

Temperature
Interval (°F) Time Cell cell
. Presgure Temperature
———————  (min) : o
(psia) (°F)
t T
b a
180 180 0 37 262
180 185 0.8 32 254
185 190 1.5 28 247
120 195 2.3 24 238
195 200 3.4 20 228
200 205 5.6 16 215
205 210 14,4 14 210

 

is submerged in water,’ a modest heat flux of 10,000 Btu/hr.ft? would re-
move the heat thfough the bottom of the tank. (In the experiment involv-
ing the release of 500 1b of salt into a dished vessel, cited in Section
8.6, heat fluxes greater than 100,000 Btu/hr.ft? were found.) The 4.4
ft3/sec (max) steam generated under the hemisphere would be vented through
the &-in. pipe installed for this purpose.

 

*Memorandum from L, F. Parsly to E. S. Bettis, Jan. 23, 1962, "Con-
sequences of a Salt Spill into the Bottom of the MSRE Containment Vessel."
299

 

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5. R. F. Apple 51. A. I. Kraxoviak
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29. E. P. Epler 75. T. i. Row
30. A. P. Fraas 76. H. W. Savage
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33. C. H. Gabbard 7. J. H. Shaffer
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46. A, Houtzeel 92. R. E. Thoms
111-170.
171-172.
173-187.

H.

MWW EaROQ

. M.

H.

Q==

I\"i .

300

Tolson 100. G. D. Whitman
. Trauger 101. H. D. Wilils
. Ulrich 102-104. Central Research Library
Webster 105-106. Y-12 Document Reference Secction
. Weinterg 1207-10%9. Laboratory Records Department
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