ORNL-TM-2111.txt

   

 

OAK RIDGE NATIONAL LABORATORY
operated by

UNION CARBIDE CORPORATION - %
NUCLEAR DIVISION
for the
U.S. ATOMIC ENERGY COMMISSION

ORNL- TM- 2111

<31

 

MSRE DESIGN AND OPERATIONS REPORT
Part V-A

SAFETY ANALYSIS OF OPERATION WITH 233U

P.N. Haubenreich
J.R. Engel

C.H. Gabbard
R.H. Guymon
B.E. Prince

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 therefore does
not represent a final report.

RISTRIBUTION OF THIS COCUMENE 1§ UNLIMITED
-——— LEGAL NOTICE

 

 

 

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

nor the Commission, nor any person 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 centained in this report, or that the use of
any information, apparatus, method, or process disclosed in this report may not infringe
privately owned rights; or

B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of

any informotion, apparatus, method, or process disclosed in this report.
i As used in the above, ‘‘perton acting on behalf of the Commission’’ includes eny employes or
contractor of the Commission, or employee of such contractor, to the extent that such employee
ot contractor of the Commission, ot employee of such contractor 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-2111

Reactor Division

MSRE DESIGN AND OPERATTONS REPORT
Part V-A
SAFETY ANALYSIS OF OPERATION WITH 233y

. Haubenreich
. Engel

. Gabbard

. Guymon

. Prince

Qg
HidHy s

FEBRUARY 1968

OAK RIDGE NATIONAL LARORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U.5. ATOMIC ENERGY COMMISSION

LEGAL NOTICE

This report was prepared as an account of Government spenscred work. Neither the United
States, nor the Commission, nor any person acting on behalf of the Commission:

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

B. Assumes any linbilities with reapect to the use of, or for damages resulting from the
use of any information, apparatus, method, or process disclosed in this report.

As used in the above, ‘‘person acting on behalf of the Commiasston® includes any em-~ e
ployee or contractor of the Commission, or employ?e of such contractor, to the extent that ’
aych employee or contractor of the Commission, or employee of guch contractor prepares,
disseminates, or provides access to, any informatioh pursuant to his employment or contract
with the Commission, or his employment with such contractor.
CONTENTS
PREFACE ¢ &« ¢ ¢ ¢ o v ¢ o & 4 o 4 o o o o o o o o o
LTIST OF FIGURES « ¢ « & + &+ ¢ 4 o o o s s o »
INTRODUCTIONG« « + o o o o = « o o o o o o o s o o o
1. REACTOR SYSTEM o ¢ o o ¢ o o o o & o o o o o
1.1 Fuel and Primary System Materials . . . . .
1.1.1 Salts. « « « ¢« v o « &
1.1.2 Salt Container Material. . .
1.1.3 Moderator Material . . . . « . . . .
, 1.1.4 Compatibility of oalt, Hastelloy-N, and Graphite
1.2 System Components ¢ o s e o s o s s o &
2.  CONTROLS AND INSTRUMENTATION . . . . . . . . .
2,1 Control Rods and Drives « « « « « « . « .
2.2 Safety Tnstrumentation. .
2.3 Control Instrumentation « « « « « « + ¢ « &
o4 Neutron SOUTCES « o « « = « o o o o o o o &
2.4.1 Sources Inherent in Fuel Salt. . . .
2.4.2 External Source. « « « ¢ o o o o o .
2.5 Electric Power System « « « « « « o + o « o
2.6 Physical Layout of Instruments and Controls .
3.  PLANT LAYOUT T
L.  CONTATNMENT e e e e e e e e e e e e e

iii

L.l Description « « « o« « o o o o o o o o o .
4,1.1 Contaimment During Operation . .
4,1.2 Contaimment During Maintenance .

J,2 FEXPEri€NnCes « o o o o o o o o« o o o o o o

4,2.1 Contaimment During Maintenance . . .

»

L.2.2 Primary Contaimment During Operation .

4.2.3 Secondary Contalmment During Operation .

SITE o ¢ ¢ o o ¢ o o o o &

vii

= woww

10
10
11
13
13
16
16
17
17
18
18
19
19
20
20
20
23
23
23
23
2L
27
lO.

iv

OPFRATION o 4 & » a4 & 2 4 s sas ¢ 8 e s+ o s
6.1 Staff and ProcedUresS. « « « « « o o o s o o &
6.2 Chronological Account . . . . . . . . .

6.3 Evaluation of Experience. . .
HANDLING AND LOADING =°3U. . . . . . . . . . . .
T.1l Production. . « ¢« « o« « ¢« o &« + &

T.2 Major Additions Through a Drain Tank.

7.3 ©Small Additions Through the Sampler-Enricher,
SAFETY OF ROUTINE OPERATIONS . . « + . + .+ + « & .
BREACH OF FPRIMARY CONTATIIMENT.

9.1 Damaging Nuclear Incidents. . . . . . . . .

9.1.1
g.1.2
9.1.3
9.1.hL
9.1.5
9.1.6
9.1.7
9.1.8
9.1.9
9.1.10
9.1.11
9.2 Damage
g.2.1
9.2.2
9.2.3
9.2.k4
9.2.5

General Considerations . . . .
Uncontrolled Rod Withdrawal . . . . .
Sudden Return of Separated Uranium .
Fuel Additions . . . . . . . .
Graphite Effects . . . . .

Loss of Load &+ &« « &+ + . . .

Toss of Flow .+ « « & & v & o « « o &
"Cold-Slug" Accident .

Filling Accident . . . . . . . .
Afterheat . . . . . . . « ¢« . &« o .
Criticality in Drain Tanks .

from Other Causes. . « « « &+ o« + &+ &
Thermal Stress Cycling .

Ireezing and Thawing Salt., . . . . .
Excessive Wall Temperatures

Corrosion. « o v « v v v ¢« & 4 & 4 o

Radiation Damage to Container Material

RELEASE FRCM SECONDARY CONTATNMENT . . . .
PREFACE

This report is one of a series that describes the design and opera-
tion of the Molten Salt Reactor Experiment. All the reports have been

issued with the exceptions noted.

ORNL-TM-T728 MSRE Design and Operations Report, Part I,
Description of Reactor Design by
R. C. Robertson

*
ORNL-TM-T729 MSRE Design and Operations Report, Part II,
Nuclear and Process Instrumentation, by
J. R. Tallackson

ORNL-TM-T730 MSRE Design and Operations Report, Part III,
Nuclear Analysis, by P. N. Haubenreich,
J. R. Engel, B. E. Prince, and H, C. Claiborne

X
ORNL-TM-T731 . MSRE Design and Operations Report, Part IV,
Chemistry and Materials, by F. F. Blankenship
and A, Taboada

ORNL-TM-732 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

ORNL-TM-2111 MSRE Design and Operations Report, Part V-4,
Safety Analysis of Operation with 233U, by
P. N. Haubenreich, J. R. Engel, C. H. Gabbard,
R. H. Guymon, and B. E. Prince

ORNL-TM-733 MSRE Design and Operations Report, Part VI,
Operating Limits, by 5. E. Beall and
R. H. Guymon

*
ORNL-TM-907 MSRE Design and Operations Report, Part VII,
Fuel Handling and Processing Plant, by
R. B. Lindauer

 

*
These reports are in the process of being issued.

**These reports will not be issued.,
ORNL-TM-908

ORNL-TM-909

ORNL-TM-910

ORNL-TM-911

vi

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

MSRE Design and Operations Report, Part IX,
Safety Procedures and Emergency Plans, by
A, N. Smith _

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

MSRE Design and Operations Report, Part XIT,
Test Program, by R. H. Guymon,
P. N. Haubenreich, and J. R. Engel

MSRE Design and Operations Report, Part XII,
Lists: Drawings, Specifications, Line Schedules,
Instrument Tabulations (Vol. 1 and 2)
Figure
1.1

1.2
1.3
1.k
1.5
2.1
L.

L.2
6.1
6.2
6.3
7.1

9.1

9.3

9.h

9.5

9.6

vii

LIST OF FIGURES

Comparative Stress-Rupture Properties for Irradiated
Hastelloy-N at 650°C.

Comparative Rupture Strains for Irradiated Hastelloy-N
at 650°C.

Comparative Tensile Properties of Irradiated and Un-
irradiated MSRE Surveillance Specimens, Heat 5085.

Comparative Creep Rates for Surveillance and Control
Specimens at 650°C,

Hastelloy~N Surface from Exposed MSRE Surveillance
Samples. Surface Deposit from Hastelloy-N in Near
Contact with Graphite.

Control Rod Worth in MSRE.

Schematic of MSRE Secondary Contaimment Showing
Typical Penetration Seals and Closures.

cecondary Contaimment ILeak Rates.

MSRE Activities from July 1964 through December 1965,
MSRE Activities in 1966,

MSRE Activities in 1967,

Arrangement for Adding £7°U Enriching Salt to
f'uel Drain Tank.

Observed Response of Nuclear Power to Small Step
Changes in Reactivity at Various Initial Powers
With 35U Fuel.

Results of Uncontrolled Rod Withdrawal with No
Safety Action, 2337 Fuel,

Results of Uncontrolled Rod Withdrawal with No
Safety Action, Z7°U Fuel.

Results of Uncontrolled Rod Withdrawal, with Scram at
11 Mw, =33U Fuel.

Time Dependence of Reactivity Addition due to Sudden
Resuspension of Uranium in Lower Head of
Reactor Vessel

Temperature Excursion Caused by Sudden Resuspension of
Uranium Equivalent to 0.25% t8k/k if Uniformly
Distributed; Initial Power, 1 kw; No Safety Action

12
15

22
25
29
30
31

37

k3

L8
50

51

54

56
Figure
9.7

9.8

9.9

9.10
9.11

9,12

viii

Effect of Magnitude of Reactivity Recovery on Peak
Pressures and Temperatures during Uranium
Resuspension Incident with No Safety Action.

Effect of Initial Power on Peak Pressure Rise Caused
By Sudden Resuspension of Uranium Equivalent to
0.25% 8k/k if Uniformly Distributed;

No Safety Action.

Effect of Magnitude of Reactivity Recovery on Peak
Pressures and Temperatures During Uranium Resuspension
Incident with Rod Scram at 11.25 Mw., Py = 1 kw.

Regulating Control Rod Motion During #7°U Fuel Capsule
Addition at Full Reactor Power.

System Response to Load Increase from 2 to 7 Mw at
Maximum Rate.

Chromium in Fuel Salt Samples

5T

58

62

65
6
INTRODUCTION

The Molten Salt Reactor Experiment is an important step in a pro-
Ject whose ultimate goal is a thermal breeder reactor operating on the
thorium—uranium-233 cycle. The breeder project is the outgrowth of ex-
tensive development of molten salt technology in the Aircraft Nuclear
Propulsion Program of the 1950f's. The MSRE was built to demonstrate that
the molten salt technology had advanced to the point that many of the
features of the proposed breeders could be incorporated in a reactor that
could be operated safely and reliably and could be maintained when neces-
sary. The MSRE began nuclear operation in June 1965, reached full power
in May 1966, and now has passed 8000 equivalent full-power hours of opera-
tion, In a large measure, it has met its objectives. It is now proposed
to extend its usefulness by experimental operation of a sort not contem-
plated in the original planning and safety analysis. In order to obtain
information directly relating to tThe neutronic and stability analyses of
233y breeders, we propose to remove the present uranium from the fuel salt
and substitute Z>>U. After the replacement of the uranium, the reactor
would be taken to full power agein and operated for the better part of a
year to obtain data on 233 eross sections.

This report presents the data and the analyses that have led us to
conclude that it is safe to load the MSRE with Z23U and pursue the pro-
gram of experimental operation. It leans on the MSRE Design Reportl and
the original safety analysis repor"c2 for much of the detailed description
of the reactor components and the site. A comprehensive report on the

instruments and controls is being issued concurrently,3 sc no attempt is

 

1R. C. Robertson, MSRE Design and Operations Report, Part I —
Description of Reactor Design, ORNL-TM-728 (January 1965)..

Z5. E. Beall, P. N. Haubenreich, R. B. Lindauer, and J. R. Tallackson,
MSRE Design and Operations Report, Part V — Reactor Safety Analysis Report,
ORNL-TM-T732 (August 196k4).

°J. R. Tallackson and R. L. Moore, MSRE Design and Operations Report,
Part IT-A — Nuclear and Process Instrumentation, ORNL-TM-729 (January 1968).
made here to give a complete description of those systems, What this
report does include is a summary of relevant experience and new inflorma-
tion and an assessment of the safety of operation with 233 , taking into
account that experience, the physical condition of the system, and the

different neutronic characteristics with #>°U in place of 2397,
1. REACTOR SYSTEM

At the time the MSRE design report and the original safety analysis
report were issued, construction of the reactor was essentially complete.
The description of the components and the mechanical systems given in
those reports therefore is as-built, is still valid in all essential
respects, and will not be repeated here. There is new information on the
materials, however, as a result of further testing and experience and this

is discussed below.

1.1 Fuel and Primary System Materials

1.1.1. Salts

The original safety analysis considered the possible use of fuel
salts of three different compositions. One of these, Fuel C, has been
used throughout all the operation to date, and the composition therefore
has been proved in use. The present mixture contains 75U as the fissile
material, diluted with 2387 to provide a total uranium concentration of
about C.9 mole percent. This mixture will be fluorinated to remove that
uranium, then 233UF4-LiF eutectic will be added and nuclear operation re-
sumed. With most of the non-fissile uranium removed, the operating
uranium concentration with 77U will be gbout 0.2 mole percent, otherwise
the chemical composition of the fuel salt will be practically unchanged.
There should be no significant difference in the chemical stability of
the salt. The higher uranium concentration was desired originally because
at that time it was considered possible that the fission products from
one fission might use up more than four fluorine atoms or that fluorine
might be lost by some other process, causing some reduction of UF, to UF=.
This process, if allowed to continue, could lead ultimately to precipi-
tation of metallic uranium. The higher concentration of UF, gave more
time for careful analysis and determination of the actual situation before
uranium precipitation could occur. It turned out that the fission pro-
ducts from one fission tie up less than four fluorine atoms, not more,
and there is no significant loss of fluorine by other reactions, so the

need for UF4 much in excess of the mininmum regulred for criticality does
not exist, The slight amount of flucorine liberated as a result of fission
gradually oxidizes some of the UFx in the salt to UF4. A reducing environ-
ment must be maintained to prevent attack of the container walls, so the
UF= concentration is held at approximately one percent of the total
uranium by exposing a rod of beryllium metal in the sampler-enricher at
intervals of several weeks. (The reaction is 2 UFy + Be — 2 UF=s + BeFo.)

During the operation of the MSRE, corrosion products and fission pro-
ducts in the fuel salt have not built up to concentrations that could have
any deleterious effect on chemistry. Moisture has been effectively ex-
cluded from the salt systems as evidenced by fuel salt analyses which have
consistently shown only about 50 ppm oxide. Since this is far below the
solubility of ZrOz, no precipitation of ZrOz is expected. Fluorination
of the salt to remove the original charge of uranium will produce ad-
ditional corrosion products, but the salt will be given further treatment
(probably reduétion and filtration) to insure that concentrations are
acceptably low when the salt is returned for use in the reactor.

In summary, no problems have been encountered with the fuel salt

chemistry and none are expected in the 33U operation.

1.1.2 Balt Container Material

 

All the salt piping and vessels in the MSRE are made of the nickel-
base alloy Hastelloy-N (also called INOR-8) which was especially developed
to be corrosion-resistant in molten fluorides and to have good high-
temperature physical properties. IExperience with and testing of Hastelloy-N
since the construction of the MSRE have shown that it is indeed corrosion
resistant, but that certain cof its high-temperature physical properties
suffer under prolonged neutron irradiation. Corrosion experience is
summarized in Section 9.2.4. Effects of irradiation are discussed below.

Irradiation of Hastelloy-N has little effect on the yield strength
and the secondary creep rate, but causes drastic reduction in the rupture
ductility and the creep rupture life. Rupture ductility in creep tests
may be reduced from strains of 8 - 12% to as little as 0.5 to 4%. Rupture
1life may be reduced by as much as a factor of ten at high stress levels.
The damage is believed to stem from n-0 reactions preoducing helium that

collects in grain boundaries and promotes intergranular cracking. This
type of damage is quite general among iron- and nickel-base structural
alloys and can be caused by n,0 reactions of fast neutrons as well as by
thermal neutron absorptions in boron, However, in the Hastelloy-N in the
MSRE, helium production is predominantly from boron. Thus the degree of
damage is primarily a function of thermal neutron fluence and practically
saturates at 105% n/cm2 or less.

A comparison of stress-rupture characteristics of irradiated and un-
irradiated Hastelloy-N is given in Figure 1.1. The rupture strains ob-
served in these tests are shown in Figure 1.2. The irradiated specimens
were from four commercial heats of metal used in the fabrication of the
MSRE reactor vessel. The specimens irradiated in MSRE were removed in
August 1966 after exposure to a thermal neutron fluence ranging from
0.5 x 102° to 1.3 x 10°° n/em® (Reference 4). (Hastelloy-N specimens have
since been irradiated in the MSRE core to higher doses, but these speci-
mens were of heats modified by the addition of 0.5% Ti or Zr to greatly
reduce radiation damage and so are not directly relevant to the condition
of the MSRE vessel.) Those marked ORR were exposed in a helium atmosphere
in that reactor to 1.4 x 10%° to 5.2 x 10%° n/em®,

Figure 1.3 illustrates that yield strength was not affected and that
ultimate strength was not drastically reduced by the irradiation in the
MSRE. The total elongation in these tensile tests was reduced, but not
nearly so much as in the creep-rupture tests. For example, at 650°C
(1200°F) the elongation was 13% in the tensile test at a strain rate of
0.05 min~1 compared to elongations of 1 to 4% in the longer-term tests
shown in Fig. 1.2. Figure 1.4 shows that there was practically no dif-
ference between the secondary creep rate of irradiated specimens and un-
irradiated control specimens,

The effects of neutron irradiation must be considered against the
background of allowable stresses used in the MSRE design and the antici-

pated service life of the reactor. When the MSRE was designed, Hastelloy-N

 

4. E. McCoy, Jr., An Evaluation of the MSRE Hastelloy-N Surveillance
Specimens — First Group, ORNL-TM-1997 (November 1967).
ORNL-DOWG 67-7940R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70
Y
T THIN T
‘YP?CAL UNIRRADIATED DATA
60
\\
50 h
N
¢ \
g ‘\T\'—-..__h_hq-., i‘IH
g 40 — e SN
2 AVERAGE |RRADIATED DATA N \\
0 E || tLO j\ol‘--:kno N
wn i ! b i N
y 30 T " | © S \\
i : i t My
B MSRE  ORR Pl NN
o 5065 o : \ N
20 A 5067 \j\\\
8 o 5085 \ 1 N, b
¢ 50814 o _ ‘ Pl
o : ‘ !
10 L e
| b
L L L
0 Ll L | ;
10" 10° 10! 102 Tok 104

RUPTURE TIME (hr)

Figure 1.1. Comparative Stress-Rupture Properties for Irradiated
Hastelloy-N at 650°C.
ORNL-DWG 67-7944

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

& MSRE ORR ,
A A 5085
o 5065
9 ’ 0 5081 T
B o 5067 u ’
=
Z 4 L
= . X
-
m i
w3 1
P
% qa
=2 r |°j’ |
o 2 ; ‘ 4 1 ]
' 4 ‘ ) it
_4] ‘ d ]
A ‘ D o t H :
: V“ A : [
! ik il
o il P , o
! ° a L b
o . 1 L Ll P
10~ 10° 10" 10° 10° 10*

Figure 1.2.

at 650°C.

RUPTURE T!ME (hr)

Comparative Rupture Strains for

Irradiated Hastelloy-N
ORNL-DOWG 67-2459A

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1.2
1.0 ! —_— . ®
o — —— i
& | | :
‘5_-'0-08 i ——g=— ; b N ]
S| : T - ~a
e N
e | _r T
£|2 0s T
2 2 | i \\. j
ElE | ® YIELD STRESS Nl
04 f— - A ULTIMATE STRESS ‘L-\
= TOTAL ELONGATION .
‘ é=0.05 min ' "\
02 |- - '~
N
: ..‘
; \
O i
-0 100 200 300 400 500 600 700 800 900 1000
TEST TEMPERATURE (°C)
Figure 1.3. Comparative Tensile Properties of Irradiated and

Unirradiated MSRE Surveillance Specimens, Heat 5085.
ORNL-DWG 67-7939

 

70 I

 

 

60
| A

 

<
Qo
N

 

H
o

¢
N

 

STRESS (1000 psi)
ol
o
N\
N

 

 

 

 

 

 

 

 

 

 

 

 

 

 

N
o

508t 5085
CONTROL o A

10 ‘ SURVEILLANCE e A

0
w0 ? 102 102 . ot 10° 10!
MINIMUM CREEP RATE (%/hr)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Figure 1l.4. Comparative Creep Rates for Surveillance and Control
Specimens at 650°C.
10

had not yet been considered by the ASME Code Committee, so a curve of
maximum allowable stress as a function of temperature was prepared using
code criteria and the physical properties of unirradiated Hastelloy-N.5
Below 900°F the allowable stresses were governed by the tensile and yield
strengths, from 900 to 1150°F the 100,000-hour rupture stress was limiting,
and above 1150°F the stress that produces a secondary creep rate of O.l%

in 10,000 hr governed. ©Subsequently the ASME Boiler and Pressure Vessel
Code Committee approved Hastelloy-N for construction under the Unfired
Pressure Vessel Code and the Nuclear Vessel Code. Maximum allowable
stresses approved under the codes are essentially those on the MSRE design
curve, Maximum allowable primary stress at 1200°F is 6000 psi and at 1300°F,
3500 psi, Actually, in the design of the MSRE, primary stresses were
generally limited to 2750 psi or less except in a few locations where lower
temperatures justified higher allowable stresses.

The data in Fig., 1.1 suggest that the difference in the rupture life
of dirradiated and unirradiated Hastelloy-N decreases as the stress is re-
duced and that it may be very small at the MSRE design stresses.

Implications of the effect of irradiation on the serviceability of

the MSRE primary containment are discussed in Section 9.2.5.
1.1.3 Moderator Material

Further information on the MSRE graphite has come from exposure of
surveillance specimens in the MSRE core (discussed in the next section)
and from irradiations in the ORR to doses far beyond those anticipated in
the MSRE. The irradiated specimens showed practically no dimensional
changes at doses that may be reached in the MSRE and no other changes of

any consequence to the MSRE,
1.1.4 Compatibility of Salt, Hastelloy-N, and Graphite

Analyses of several hundred samples of fuel salt, taken over more

than two years of operation, and examination of two sets of metal and

 

5Robertson, op.cit, pp 119 - 120,
11

graphite specimens exposed for thousands of hours in the MSRE core have
further demonstrated the compatibility of the fuel, the moderator, and the
container materials, |

Interaction between the salt and the Hastelloy-N appears to have been
1imited to deposition of an extremely thin layer of noble metal fission
products on loop surfaces and an inconsequential amount of corrosion, i.e.,
leaching of chromium. This is discussed more fully in Section 9.2.h4.

Two sets of graphite specimens exposed in the MSRE core showed no
attack by the salt in 2800 and 4300 hr. There was no change in the surface
finish, no intrusion of salt into the pores and no further cracking of the
graphite, Radiochemical analyses and examinations with electron micro-
probes showed that noble metal fission products were deposited at the sur-
face (less than 0.3 mil deep) and products of xenon and krypton decay were
distributed throughout the specimens. Although of great interest, the
effects of these fission products in the MSRE are insignificant.

Where graphite and Hastelloy-N are in direct contact in the MSRE core,
some carburization of the metal was expected. Thus, when the core was as-
sembled, sacrificial metal inserts were included at contact points. The
surveillance specimens showed that some reaction does take place where
surfaces are in contact, Figure 1.5 is a section through a Hastelloy-N
surface that was in contact with graphite through 4800 hr at 1200°F in
the first set of surveillance specimens., (The affected layer is small
compared to the thickness of the inserts between graphite and structural
metal in the core.)

Substitution of 233U should in no way affect the compatibility of the

materials in the primary system.
1.2 System Components

There has been no modification of any of the reactor components
described in Section 1.2 of the original safety analysis report, so the
descriptions given there are still valid. The heat transfer performance

of both the primary heat exchanger and the coolant radiator proved to be
 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

   

  

  

 

 

 

     

   

 

 

' i
, , ,
. $§
'k
. 'g O
ol
P - > ‘
A v
6
1z -
" Figure 1.5. Hastelloy-N Surface from Exposed MSRE Surveillance
Samples. Surface deposit from Hastelloy-N in near contact with
graphite. , L | T
. . \
13

less than predicted, however.® The important conseguence of this is that
the steady-state reactor power has been limited to about 7.5 Mw instead

of the 10 Mw originally contemplated in the safety analysis.

2 CONTROLS AND INSTRUMENTATION

The system of instrumentation and controls remains essentially as
described in the original safety analysis report, and experience has
shown that it performs as intended. There have been a few changes, how-
ever, notably the addition of a period scram of the control rods. Also
the change to 33U has some control implications. These points will be

discussed in this section, which follows the outline of the original report.

2.1 Control Rods and Drives

 

The mechanical description of the MSRE control rods and drives pre-
sented in the original safety analysis report is still valid since no
changes have been made.

The measured worth of the control rods with £5U fuel in the reactor
is slightly greater than was predicted.” The worth of one rod was found
to be 2.26% Bk/k, compared to a prediction of 2.11% 8k/k. The observed
worth of three rods was 5.59% 8k/k; the predicted worth, 5.46%. When
2337 is substituted for the present partially-enriched uranium, the neutron
diffusion length will be longer and the rod worth greater by a factor of
1.3. The worth predicted for one rod is 2.75% ak/k; for three rods,
7.01% ®k/k. (These worths are for the 5l-inch travel between limit

switches, with the critical concentration of uranium in the fuel., The

 

SMSR Program Semiann. Progr. Rept. Aug. 31, 1966, ORNL-L037,
pp. 35-39.

7B. E. Prince, et al., Zero-Power Physics Experiments on the MSRE,
ORNL-4233 (February 1968).
1k

one-rod worth is with the other two rods fully withdrawn). Figure 2.1
shows measured and predicted worth curves with the original fuel and the
predicted curve with 32U fuel.

Experience has shown that mechanically the rods are quite reliable,
There have been 28 unscheduled control-rod scrams (through November 1967)
when fuel salt was in the reactor vessel. (None was caused by a process
variable actually exceeding the scram setpoint.) Scrams for testing
purposes brought the total for each rod to well over 100 scrams. Never
has a rod failed to drop on request.

Rod drop time with the core hot has ranged from 0.97 down to O0.71 sec.
depending on the rod and its length of prior service. (The rod becomes
more flexible with use, leading to shorter drop times.) The drop time
corresponding to the delay and acceleration assumed in the safety analysis
is 1.4 sec., One rod drive was replaced in September 1966 because the drop
time was slower than normael., It had been 0.96 sec, at the end of a run
and during the shutdown with the core and rod cold the drop time approached
1.3 sec. After replacement of the rod itself (see below) did not improve
the drop time, the cause of the slow drop was found to be a bent air tube
in the drive unit that rubbed the inside of the hollow rod.

Two of the three rods currently in the reactor have been in service
since the start of nuclear operation in May 1965. The rod that was re-
placed in September 1966 had developed a tendency to hang on withdrawal
about two inches above the fully inserted position. The hanging was at-
tributed to interference between a sharp corner on the bottom fitting on
the rod and the lower end of a guide rib in the rod thimble. The upper
corners of the end fitting on the replacement rod were rounded and no
further difficulty has been encountered. Aside from this instance the
rods have always moved freely in either direction.

The gadelinium loading in the poison elements is such that the rods
should remain black to neutrons for much longer than the expected opera-
tion of the MSRE, In July 1967, tests verified that there had been no

change in sensitivity due to poison burnout.
REACTIVITY ADDED (% Ak/k)

2.8

2.4

20

16

1.2

0.8

0.4

15

ORNL-DWG 68~965

 

/

 

 

 

0BSE RVED,
~~ 235

 

 

233U

  

,///1/

€ALCULATED

i

 

(CALCULATEy ///
/
/ //

 

 

/
/
7

 

 

 

 

 

 

 

 

 

 

 

Figure 2.1.

16 24
CONTROL ROD POSITION (inches withdrawn)

32

40

48 56

Control Rod Worth in MSRE.
16

2.2 Safety Instrumentation

 

An up-to-date description of the safety system is given in Part IT-A
of the MSRE Design and Operations Report, recently issued.®

Shortly before the beginning of nuclear operation with Z7°U, a period
scram was added to the safety system described in the original safety
analysis report. The control rods are scrammed when there are indications
of a positive period shorter than 1 second. A period signal derived from
the output of a safety chamber is included, along with flux level and core
outlet temperature, in each of three trip channels. These channels are
connected so a trip on any two channels scrams the rods.

Other changes made after the original safety analysis report was
issued are in the flux level trips. After full power proved to be about
7.5 Mw, the high level trip was set down to 11.25 Mw (150% of T.5 Mw).
This is the trip point if the fuel pump is running. If the fuel pump is
off, the level trip points are automatically reduced to 11.25 kw, and each
of the three channels must be reset manually to the higher level after the
pump is started.

Other than the changes described above, there have been no changes in
the functions of the safety system since the original safety analysis, As
will be shown in Chapter 9, no changes will be necessitated by the substi-
tution of 233U for the #7°U in the fuel.

2.3 Control Instrumentation

 

Since the time of the original safety analysis report, several changes
have been made in the control system. These changes affect normal operating
procedures, but none are of significance from the standpoint of the safety

analysis, No changes are anticipated because of the =°°U loading.

 

8J. R. Tallackson and R. L. Moore, MSRE Design and Operations Report,
Part II-A — Nuclear and Process Instrumentation, ORNL-TM-729 (January 1968)
L

2.4 Neutron Sources

The presence of a neutron source is important because of its influence
on the course of excursions from reactivity additions that begin with the

reactor subcritical.”

2.4.1 Sources Inherent in Fuel Salt

 

Alpha particles emitted by heavy elements in the fuel salt interact
with the lithium, beryllium and fluorine to produce an abundant source of
neutrons within the salt. In the original fuel salt most of the energetic
alpha particles come from the decay of 23*U, which constitutes 0.3% of the
uranium. The neutron source in the amount of fuel salt required to fill
the core was predicted to be L4 x 10° n/sec. (Ref. 10) This strength was
verified approximately during the experiments at the beginning of nuclear
operation.tt

After 237U is substituted for the present uranium, the inherent
alpha-n source will be much greater. The 74U concentration in the salt
will be higher, but its contribution to the alpha-n source will be insig-
nificant compared to that of the 77y and the daughters of “°2U. There
is a chain of short-lived alpha emitters descending from 1.9-y 228 whose
activity builds up with that of the thorium following chemical purification
of the uranium. TIn the spring of 1968, the ““8Th (and daughters) associated
with the uranium will be at about three-fourths of saturation. The pre-
dicted alpha-n source in a core full of fuel salt at that time is
3 x 108 n/sec, a factor of 70O higher than in the original fuel salt.

There will also be a substantial photoneutrbn source in the fuel salt
from interaction of fission product gamma rays with the beryllium., Tmmedi-
ately after high power operation, this source will emit more than 10° n/sec
in the core, but within about a day will have decayed below the predicted

alpha-n source.

 

®3. H. Hanauer, Role of Neutron Source in Reactor Safety, Nuclear
Safety 4(3): 52-54 (March 1963).

10p, N. Haubenreich, Inherent Neutron Sources in Clean MSRE Fuel Salt,
ORNL-TM-611 (August 1963).

118, E. Prince et al., Zero-Power Physics Experiments on the MSRE,
ORNL-L4233 (February 1968).
18

The inherent alpha-n source is particularly valuable from the stand-
point of safety because it is absolutely dependable. Wherever there is

£33y, there is the neutron source.

2.4.2 External Source

 

An external neutron source, located in the thermal shield, is pro-
vided for convenience. This source permits checking that the nuclear
startup instruments are working properly before the fuel salt is brought
out of the drain tanks into the reactor vessel., After only a small
fraction of the vessel is filled with fuel, the neutrons from the inherent
source completely overshadow the effects of the external source and give
a strong counting rate on both the wide-range channels and the startup
channel,1®

The external source is an alpha-n source consisting of a mixture of

2431pm, ®4%Ccm and Be. When the source was new the strength was 102 n/sec,

mostly from 24%¢

m alphas, The flux of fission neutrcns at the source
tube during power operation is not enough to keep the 2420y regenerated
by production from the Z4lAm, Thus the source decays after installation
with practically the 163-d halflife of the 2420n originally present. The
first source was installed in May 1965 and remained adequate through May
1967. Another source of the same type and original intensity was in-
stalled on top of the first in June 1967. This will remain adequate

through the anticipated operation of the MSRE.

2.5 Eleectric Power System

 

The electric power system at MSRE i1s essentially the same as described
in the design report, but some improvements have been made to reduce the
likelihood of power failures interfering with operation of the reactor.
These include better lightning protection for the 13.8-kv feeder lines,
installation of a battery-powered 50-kva static inverter for uninterrupted
instrument power, and provision of independent power supplies for each of

the three channels in the safety system.

 

127, R. Tallackson and R. L. Moore, op.cit,
19

2.6 Physical layout of Instruments and Controls

 

The location of the reactor controls and instrumentation is as
described in the original safety analysis report (ORNL-TM-732). DNot
described was the fire protection system. In the main control area are
three detectors, each a combination of rate-of-rise and fixed-temperature
(136°F) devices, connected to the building and plant alarm system. Carbon
dioxide fire extenguishers are readily available to all control areas.

The automatic sprinkler system in the building also covers the control

areas and computer room with fog nozzles triggered by 212°F fusible plugs.

3. PLANT LAYOUT

The plant layout was described in the safety analysis report
(ORNL-TM-732). Construction was essentially complete at the time of that
report and there has been no sighificant change from the original

description.
20
4. CONTAINMENT

The MSRE design aimed at zero leakage from the system of piping and
vessels that is the primary containment for the fission products. 1In
addition, a secondary containment system was provided to limit the release
of fission products to the environs in the event of a failure in the pri-
mary containment. Stringent leakage criteria had to be met by the secon-
dary containment, because the potential release from the primary contain-
ment in the ultimate accident might conceivably amount to practically the
entire inventory in the reactor.

The MSRE has met the criterion for primary contaimment. By the use
of welded construction with a minimum of gasketed joints, and those
pressure-buffered, zero leakage has been attained during all periods of
operation. No accident has ever occurred to test the secondary contain-
ment, but tests of variocus kinds have shown that the specified design cri-
teria have been met and routine, frequent measurements indicate that the
reactor has always operated within a satisfactory secondary containment.

Containment during ©3°U operation will be the same as in the Z75U

operation.

L.1 Description

Most of the fission products remain in the fuel salt, but the fuel
offgas is alsoc intensely radioactive, containing noble gases and part of
the noble metal fission products. Some fission products deposit on sur-
faces in contact with salt or offgas, thus presenting radiation and con-
tamination problems in maintenance and inspection. Conteinment is always

provided for these sources, the nature depending on the situation.

4,1.1 Contaimment During Operation

 

The salt is contained in piping and vessels of Hastelloy-N. This
system was designed for long operation at 50 psig and 1300°F without ex-
cessive creep and is therefore capable of containing much higher pressures
and temperatures for short periods of time. The fuel system contains only
a few flanges. In the fuel circulating loop there are three "freeze

flanges", with a metal ring seal backing up a frozen salt barrier. As in
21

all other primary containment flanges, the groove under the ring is pres-
surized to 100 psig with helium to provide a buffer zone and continuous

leak detection. There is also an access nozzle on top of the reactor

vessel with a frozen salt seal backed by a buffered ring-joint flange,

The piping and vessels in the cover gas and offgas system are of Hastelloy-N
and stainless steel, with several flanges, all leak-detected and buffered.
This grade of containment extends through the charcoal beds, where prac-
tically everything but 10.6-y 8°Kr decays, to the point where the offgas

is mixed into the ventilation stack flow.* This then is the primary con-
taimment during operation,

The sealed reactor and drain-tank cells are the secondary containment
for the fuel salt during operation. The lines and vessels through which
the cell atmosphere is recirculated by the component coolant pump are in
effect extensions of the reactor cell, The cells are held at -2 psig by
venting about 70 scf/d of the component coolant pump output tc compensate
for inleakage and deliberate inputs of nitrogen. The evacuation flow
passes a radiation monitor and an automatic block valve, then through high-
efficiency particulate filters before passing up the ventilation stack
past another set of monitors. The fuel offgas line (outside the reactor
cell) and the charcoal beds are inside enclosures through which ventilaticn
air flows directly to the stack filters., The fuel sampler-enricher, which
when it is being used is an extension of the offgas system, is in an en-
closure swept with helium that exhausts through a charcoal trap to the
stack filters. All service lines penetrating the secondary containment
are eguipped with closure devices, the type depending on the application,
as indicated in Fig. 4.l. Most of the safety block valves are actuated
by radiation monitors, but many close if the reactor cell pressure goes

above atmospheric.

 

e 85Kr concentration in the stack gas is presently & x 10~ yc/cc,
Just tolerance for immersion for 4O hours a week. After 277U is substi-
tuted, the yield of 85kr will be up by a factor of 2.8 and the offgas con-
centration will be higher by that factor. Atmospheric dispersion from
the stack makes concentrations at the ground quite insignificant in
either case.
MEMBGANE SBEoL ™

OREL DWG. 67-2688

. ~ JLNELD LOCKS
i e ®

 

 

 

 

 

 

 

 

 

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COFTOEATED WELDS
cv
HELLM  SUPoLY 1 o
-— —ept b N—GT
SARETY VA J
SEALED CONTAINMENT
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SLARETY VA
p—
Ol
VLT
P g eaTER TanK
L \ J
O STALW ) o

 

S[ACETY VA,

  
 
 

CONTAINTMENT
ENCLOSULE-

 

Figure 4.1.
and Closures.

Schematic of MSRE

 
  
    
    

 

 

 

OFFGAS

ELECTRICAL

 

‘t SO R N
=

=
METAL -TO -CERQAMI

‘\:- AR SUPPLY

 

SAFETY VA,
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osconnEcTd L_'r%_‘;—_\&'f

 

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I S EAL ' Ng_ SuRP v
LT P SARETY VA. @ SARETY VA, :
! HERMOCOUPLES LES (VY cell evacuaTion Yo e
i
i
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IOAINTENANCE ™3

   
 

 
 
 
     
 

 

 

 

A0 BUTTERFLY VA

TUBRET, SEALS

ouPTUE
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|<"TO vAPOD, CONDENSING
SygTEM

Secondary Containment Showing Typical Penetration Seals

cc
23

L,1.2 Containment During Maintenance

 

Most maintenance does not entail opening the contaimment described
above, When it is necessary to perform work inside the reactor cell, a
30-inch line is opened, connecting the cell to the ventilation stack
through the high-efficiency filters. Then when the opening necessary for
maintenance is made in the cell roof, air is drawn down into the cell
from the work area. The work area is maintained slightly below atmos-
pheric pressure to provide another line of defense against unfiltered re-
leases. Contaminated equipment and tools are bagged in plastic or sealed

in cans before being withdrawn from the containment.

L,2 Experience

4,2,1 Containment During Maintenance

 

Experience with regard to contaimment during maintenance can be
briefly summarized. In no case has personnel exposure exceeded normal
occupational limits, and the maximum release of fission products to the

envirorment in any week has been less than 0.2 curie.
4,2.2 Primary Contaimment During Operation

The primary contairmment of the fuel salt has been perfect. Buffer
pressure has been maintained on the freeze flanges at all times, ensuring
zero leakage of salt. The tightness of the system is indicated by buffer
gas leakage, which is less than 5 x 107* cm®/sec from each flange buffer
zone when the system is hot.

There has been no significant release from the radiocactive gas sys-
tem. Occasionally during fuel salt sampling, minute quantities (<10 pc)
of fission products have been vented to the stack when the sampler en-
closure is purged, but large releases of this kind are impossible because
the source is limited. The radiation block valve on the offgas line has
been called on to block release of activity on only one occasion., In
October 1966, as a result of excessively rapid changes in the fuel pump

pressure, radiocactive gases entered the pump shaft seal vent line.l® The

 

1°MSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4119,
pp. 28-29,
2

radiation monitor blocked the line before there was a measurable (10 nc)
release to the stack. Subsequently a charcoal filter was installed in
this line even though the need for the troublesome pressure changes was
eliminated.

The primary system is routinely pressure-tested by applying 65 psig
helium pressure in the pump bowl with flush salt circulating at 1200°F.

Never has there been any indication of leakage.

4,2,3 Secondary Containment During Operation

 

In 1962, soon after the construction of the reactor cell and drain
tank cells was completed, they were tested hydrostatically at L8 psig
(measured at the tops of the cells) to assure they would withstand the
design pressure of 40 psig. The first complete leak test was made in
1965, after the vapor-condensing system was connected. All individual
service line block valves and check valves that could become secondary
containment in case of a catastrophic failure in a cell were tested and
proved satisfactory. A check at 1 psig showed no leaks in the welded
membranes covering the cells., After the top blocks were installed, all
openings closed, and penetrations sealed, the cells were tested successively
at 20, 30, 10, and -2 psig. At each positive pressure level all pene-
trations, cable seals, tube fittings and external parts of valves com-
prising secondary contaimment were checked for leakage. Leakage rates
were measured with the results shown in Fig. k4.2,

The safety analysis had assumed that the leakage rate from the cells
would reach l% per day at 39 psig and 260°F. Corresponding leakage rates
at other pressures would be approximately as indicated by the curves on
Fig. 4.2; the exact relationship between pressure and leakage would depend
on the relative contribution of wvarious types of leaks. The secondary
contaimment was judged to be acceptable since all the leakage rates
measured at positive pressure fell well below even the straight line (which
is a conservative way to extrapolate to higher pressures). Subsequent tests
at positive pressure also showed acceptable rates. In the fall of 1966,
two tests at 10 psig gave 65 and L3 scf/d at 10 psig. In June 1967, a
test at 20 psig showed only 35 scf/d.
CONTAINMENT LEAKAGE RATE (scfd)

25

ORNL -DWG 66-4753
500

 

 

400

300 ‘ @ /

7

 

N

 

N Y,

 

 

 

 

 

 

 

 

 

 

 

®
.
100 '
¢
0 () ASSUMED LINEAR RELATIONSHIP
° @ ORIFICE FLOW
(® TURBULENT FLOW
-100 / @ LAMINAR FLOW 7]
® EXPERIMENTAL DATA
-200 . . .
-10 0 10 20 30 40

CONTAINMENT PRESSURE (psig)

Figure. 4.2. Secondary Containment Leak Rates.
26

Leakage rates with the cells at positive pressure are measured by
changes in differential pressure between the cell atmosphere and a ref-
erence volume within the cell. During operation, when the reactor cell
is held at -2 psig, the inleakage rate is determined by a balance on
measured inputs and exhaust rate with corrections for small changes in
pressure, Although large changes in inleakage are detectable almost im-
mediately, an accurate determination at the normal rate takes about a
week of data under fairly steady conditions.

When the cell is at negative pressure, gas enters through sump bub-
blers and by leakage from pressurized service lines and penetrations, but
the important component is the cell inleakage through routes that repre-
sent possible outleakage paths at positive pressure. This inleakage has
been less than 50 scf/d during all periods of operation. On three oc-
casions other inputs have caused anomalously high measurements, but on
investigation the actual cell inleakage was found to be acceptably low,
In May 1966, the reactor was shut down when the apparent inleakage in-
creased above 100 scf/d and it was found that nitrogen was leaking into
the cell from the pressurized thermocouple penetratlion sleeves. Gas
cannot leak out by this route so a flowmeter was installed and this input
was factored into the balance. In November 1966, another indication of
high inleakage was found to be leaks in the cell from pneumatic valve
operator lines. Flowmeters were installed to measure this inflow and the
actual inleakage was again found to be low. Over the next two months,
however, the air line leaks increased until the error in measurement be-
care so large that the actual cell leakage could not be determined with
satisfactory accuracy. In January 1967, the reactor was shut down and
the air line disconnects in the reactor cell were all replaced, stopping
the leaks. (The elastomer seals in the original disconnects had suffered
radiation damage and were replaced by metal-to-metal seals.) Since then
the cell leak rate measurements have always been acceptable, usually

below 25 sef/d.
27
5. SITE

The MSRE 1s situated in Melton Valley, about a mile across a ridge
from the main X-10 Area of Oak Ridge National Laboratory. A detailed
description of the site, including surrounding population densities and
geophysical features, is given in Chapter L of the original MSRE Safety
Analysis Report. There have been minor changes in population within the

plant areas; otherwise the original description is still valid.

6. OPERATION

Operation of the MSRE has attained most of the objectives of the ex-
periment. Mechanical problems in the operation of some components and
systems have been met, and overcome, but experience has shown no de-

ficiency with regard to safety.
6.1 Staff and Procedures

The MSRE is operated routinely by four crews on rotating shifts,
each crew consisting of a minimum of one supervisor and two operators.
Supervisors and operators are trained, examined and formally certified
before being assigned to a crew, During periods involiving experiments,
sampling or other such operations, the crews are augmented by additional
trained members of the MSRE staff. Analysis and maintenance support is
as described in the original safety analysis report.

There has been very little turnover of personnel, Of the persons
presently on the operating crews, only two operators were not part of
the MSRE staff from the beginning of nuclear operation.

- Operating procedures are contained in Part VIII of the MSRE Design

and Operations Report.l4 Loose-leaf versions, formally updated, are used

 

14R. H., Guymon, MSRE Design and Operations Report, Part VII —
Operating Procedures, ORNL-TM-908, Vol. 1 and 2 (December, 1965).
28

by the operating crews. ©OGafety procedures and emergency plans constitute
Part IX of the Design and Operations Report.'® Part VI (Ref. 16) sets
forth the safety limits on the operation.

Any modification of the system or any change in the operating pro-

cedures must first be appropriately reviewed and formally approved.l7

6.2 Chronological Account

 

The operation of the MSRE has consisted of the following phases:
pre-critical testing, initial critical measurements, low-power measure-
ments, and reactor capability investigations.l8 The last phase covers
both the approach to full power and sustained operation at high power.

Figure 6.1 outlines the activities for the 18 months beginning with
the arrival of the operating staff at the reactor site in July 196L.
Subsequent operation is outlined in Figures 6.2 and 6.3.

Precritical testing served both to check out the equipment and train
the operators. Coolant and flush salts were successfully loaded in the
molten state and few difficulties were encountered in the integrated
operation of the system with these salts. It was found, however, that the
radiator dcoors would not operate properly after being heated, which would
require modifications before power operation. The final test before be-
ginning nuclear operation was loading, circulating, and sampling the fuel
carrier salt containing 150 kg of depleted uranium.

The reactor was first made critical on June 1, 1965. Highly enriched
£357 was added as the LiF-UF, eutectic, with four batches totalling
69-kg 27U added through the drain tanks and 0.6 kg added in small capsules

 

1SA, N, Smith, MSRE Design and Operations Report, Part IX — Safety
Procedures and Emergency Plans, ORNL-TM-909 (June 1965).

163, E. Beall and R. H. Guymon, MSRE Design and Operations Report,
Part VI — Operating Safety Limits for the MSRE, ORNL-TM-733 Rev. 2,
(September 1966).

17R. H. Guymon, Op.cit., Sections 13B and 13C.

18R. H. Guymon, P. N. Haubenreich, and J. R. Engel, MSRE Design and
Operations Report, Part XI — Test Program, ORNL-TM-911 (November 1966).
ORNL-DWG 67-978

 

 

 

 

 

 

 

INSPECTION AND PRENUCLEAR TESTS FINAL PREPARATIONS
PRELIMINARY TESTING OF COMPLETE SYSTEM FOR POWER OPERATION
A A A,
FINISH LEAKTEST, INSTALL INSTALL CORE SAMPLES
INSTALLATION PURGE & HEAT CONTROL RODS INSPECT FUEL PUMP
OF SALT SYSTEMS SALT SYSTEMS SAMPLER-ENRICHER HEAT-TREAT CORE VESSEL
C LTI BRI T
TEST SECONDARY
TEST AUX, SYSTEMS CONTAI T
E FINISH VAPQOR-COND. SYSTEM
TEST TRANSFER, QPERATOR MODIFY CELL PENETRATIONS ADJUST & MODIFY
OPERATOR TRAINING FILL & DRAIN OPS. TRAINING REPLACE RADIATOR DOORS RADIATOR ENCLOSURE
3 T T SIS I SIS | vrdrsinsiinns U 77777777 o]
LOAD SALT
LOAD U-235
INTO DRAIN TANKS LOAD & IN ZERO-POWER LOW- POWER
COOLANT FLUSH CIRCULATE CIRCULATE NUCLEAR (0-50 kw)
SALT SALT C & FL SALTS CARRIER EXPERIMENTS EXPTS.
[# | o A SISI SIS C7777) Crrdd7777d) 3
{ | | 1 | | ] ] | l | i 1 L | 1 l 1 |
[ T . I 1 L | | T 1 | T [ | [ T { B!
J A S O N D J F M A M J J A S 0 N D
1964 1965

Figure 6.1. MSRE Activities from July 1964 through December 1965.

6
ORNL-DWG 66-12792R

POWER (Mw)

@

»

D

N

o

INVESTIGATE GO TO OPERATE REMOVE CORE SPECIMENS REMOVE
OFF-GAS PLUGGING FULL POWER AT POWER  REPLACE MAIN BLOWER SALT PLUG
CHANGE REPAIR THAW FROZEN LINES CHECK
TEST CONTAINMENT “( -
FILTERS AND VALVES SAMPLER ST CO EN CONTAINMENT
LOW-P —
DYNAMICS CHECK OPERATE
TESTS AT POWER

CONTAINMENT

[ S

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

—
SALT

£l

F

1

{
CIRCULATING

M

|

1

 

  

T

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F H O

 

 

 

 

B T S

 

 

 

  

A M J J A

Figure 6.2. MSRE Activities in 1966.

o
POWER (rMW)

8

ORNL-DWG 67-13661R

 

REPAIR
INSTRUMENT

CHANGE CORE SPECIMENS

 

AlIR LINES
oz,

  

 

 

A M J

TEST CONTAINMENT
e e

 

 

 

 

 

 

J

REPAIR

 

 

FUEL

FLUSH O

 

A

 

Figure 6.3. MBSRE Activities in 1967.

REPAIR
SAMPLER

 

 

— SAMPLER
Pttt N ———

 

 

 

WIRING
———

 

 

 
 

T¢
30

through the sampler-enricher to reach criticality. Another 6.6 kg was
added in 85-g capsules while measurements were made of control rod worth
and various reactivity coefficients.

After the zero-power physics experiments, the fuel was drained and
stored while final preparations were completed for operation at power,

The service life of the reactor was reevaluated and some steps were taken
to prolong the life, The penetrations of the coolant salt lines through
the reactor cell wall were modified to reduce thermal stresses and increase
the permissible number of thermal cycles. Piping and vessel supports

vere adjusted to minimize stresses, and strain-gage analyses were made of
questionable points. The reactor vessel closure weld was heat-trated

in situ to improve the physical properties. Data on stresses, neutron
fluxes and the results of experimental measurements on the effects of
irradiation on Hastelloy-N were combined to establish an expected service-
able life for the reactor vessel., Before this shutdown, the fuel pump

had circulated salt for more than 2000 hours. The rotary element was re-
moved for inspection and to provide a final test of an important remote
maintenance operation, It was reinstalled when it was found to be in
excellent condition. The original, heat-warped radiator doors were re-
placed and the door guidance mechanisms were modified and adjusted to
provide reliable, free operation, hot or cold. On the radiator enclosure,
alr leakage paths were reduced, thermal insulation was improved, wires
were relocated, and cell ventilation was modified to eliminate overheating
of the surroundings.

Late in the prepower shutdown, the reactor secondary contaimment was
sealed, the vapor-condensing system was connected and the combined volumes
were leak-tested. Leaks, which were confined to service penetrations of
the reactor and drain tank cells, were repaired, after which tests over a
range of pressure showed the leakage was well within acceptable limits.
(See Chapter L4.) Meanwhile, analysis of the zero-power experiments was
completed, furnishing values for the characteristic coefficients needed
to monitor and interpret subsequent operation.

Wuclear operation resumed in December 1965 with low-power tests., A

month later the escalation of the power was started, only to be interrupted
33

at 1 Mw when small valves and filters in the fuel offgas system plugged.
Investigation disclosed a few grams of heat- and radiation-affected or-
ganic matter, presumably from oil that had leaked in through the fuel pump
rotary element, A seal-welded unit was readied, but was not installed
when larger valves and a new type of filter reduced the problem to a
manageable nuisance. After this delay, power escalation was resumed and
in May reached the capability of the heat removal system — about 7.5 Mw,

The first weeks of power operation were interrupted briefly to repair
an electrical failure in the fuel sampler and to investigate apparently
high leakage into the reactor cell. Then in July, after 7800 Mwh of power
operation, one of the air blowers used to remove heat from the coolant
salt broke up from mechanical stress, Cracks were found in the other
blower and the spare (all left over from the Aircraft Reactor Test),
necessitating procurement of three new units. While the reactor was down
for the blower replacement, the array of graphite and metal specimens in
the core was removed and new specimens were installed. The special filter
assembly in the fuel offgas line was also replaced so the first assembly
could be examined to further identify the material that had caused plugging
before the filter was installed. A complete test of the secondary contain-
ment was also completed during this shutdown.

Power operation was resumed in October with one blower, then in
November the second blower was installed and the reactor was taken to
full power., After a shutdown to remove flush salt that had accidentally
gotten into a gas line at the fuel pump during the July shutdown, the re-
actor was operated for 30 days without interruption at full power in
December and January. This run was terminated to inspect the new blowers,
to install an improved offgas filter and to replace leaking air-line dis-
connects in the reactor cell, Full-power operation was resumed late in
January and continued into May for 103 days of nuclear operation., During
this time numerous samples were taken to elucidate fission product be-
havior, long-term effects on reactivity were studied, and enriching cap-
sules were added for the first time with the reactor at power. The re-
actor was Tinally shut down for the scheduled removal of specimens from

the core.
3k

During the May-June shutdown the core sample array was reinstalled
with some new specimens, minor maintenance and inspection were carried
out, and the complete annual tests of controls and containment were
completed,

After 7 weeks of full-power operation, the reactor had to be shut
down to repair the fuel sampler mechanism and to retrieve a latch from
the sampler tube at the pump bowl. Power operation was resumed on
September 15 and continued through the end of 1967 with only brief inter-

ruptions.

6.3 Evaluation of Experience

 

From the standpoint of reactor safety, experience with the MSRE has
been most gratifying. The chemistry of the fuel salt has borne out ex-
pectations that it would be quite stable. Nor is there any trend in the
chemical analyses thai gives cause to expect instability in the future.
Very close monitoring of the reactivity has shown that all changes in
normal operation are described by the analytical mecdel to within
+ 0.05% 8k/k, indicating excellent precision of measurements and compu-
tations and no anomalous physical behavior in the system. (The only times
this difference has been exceeded was during experiments when unusual
amounts of gas were entrained in the fuel, causing the xenon poisoning to
deviate from the model by about 0.2% s8k/k.) The neutronic characteristics
of the system agree very closely with predictions. The heat removal and
nuclear dynamics are such that the system is stable at all power levels
and quite easy to control. The operations of filling, going critical,
and changing power level are simple and well-governed by control inter-
locks. An indication of the doeility of the system is that in over
10,000 h of nuclear operation, there has never been a control rod scram
because any process variable went out of limits. Corrosion has been
practically nil and, aside from tolerable changes in the physical properties
of the reactor-vessel material, there has been no deterioration of reactor
materials. Dependability of major components has generally been gocd.
Some delays were encountered in early operation because of the offgas sys-
tem and the main blower faillure, but these did not prevent the accomplish-

ment of the planned experimental program or require any undesirable
35

compromises in the operation. The safety system is reliable; no safety
circuit or component has ever failed so as to decrease the intended pro-
tection. Thus there is nothing in the experience to date to cause reser-

vations about operating with =27U.

7. HANDLING AND LOADING =37U

After the partially enriched uranium now in the fuel salt is stripped
by fluorination in the MSRE storage tank, =>3U will be added to the re-
maining carrier salt as the eutectic LiF-UFy (73 - 27 mole %). Adequate
precautions will be taken to prevent accidental criticality in handling
and storing the eutectic; techniques proven in maintenance of the MSRE
will be used to cope with the problems of radiation and contamination

which prohibit direct handling of the material.
7.1l Production

The enriching salt will be prepared in the Thorium-Uranium Recycle
Facility (TURF) by hydrofluorination of uranium oxide in the presence of
molten lithium fluoride. The process must be carried ocut remotely in a
shielded cell because of the intense activity of the daughters of =34y,
which constitutes 220 ppm of the uranium. (The isotopic composition of
the 233U feed material is given in Table T7.1l.) The molten salt will be
transferred from the process vessel to small containers for transport to
the MSRE. A total of 35 kg of 2337 will be loaded into nine 2-1/2—inch-OD
cans ranging in size from 0.5 to T kg £330 each. Forty-five capsules,
like those used previously for “°°U additions through the sampler-enricher,

will be filled with eutectic salt containing a total of 4.0 kg 237U,
36

Table 7.1

Isotopic Composition of 2337 Feed Material

Abundance
U Isotope ‘ (atom %)
232 0.022
233 91.k9
23k 7.6
235 0.7
236 0.05
238 0.1k

7.2 Major Additions Through a Drain Tank

While the uranium is being stripped from the fuel and the fuel drain
tanks are empty, equipment for adding cans of salt will be attached to the
access flange of one drain tank. This equipment is shown in Fig. T7.1l.

The procedure for making the additions is as follows. The carrier salt
will be divided between the two drain tanks, bringing the levels somewhat
below the tank centerlines. The pressure of helium in the fuel system
will be lowered slightly below the pressure in the containment enclosure
attached to the access nozzle, One can of salt will be brought from the
nearby TURF building to the MSRE in a shielded, bagged carrier. From the
carrier it will be lowered through a temporary opening into a storage
well in the contaimment enclosure. After the enclosure is sealed and
purged to reduce moisture and oxygen, the isolation valve will be opened
and the salt can will be taken from the turntable and lowered into the
upper part of the drain tank, above the salt surface. The can will re-
main suspended in the tank until the salt has melted and drained. After
a weight measurement has verified that the can is empty, it will be placed
in a storage well in the turntable and the isolation valve will be closed
until the next addition., The charging cans are individually safe from
criticality and no more than one loaded can will be in the enclosure at

any time.
37

   

PURGE GAS
SUPPLY

    
   
 
 

MAINTENANCE SHIELD =%~ /

i
K

I-_:—r- g
U/

TURNTABLE AND
STORAGE WELLS- <

 

  

  
  

CONTAINMENT ENCLOSURE
AND STANDPIPE ASSEMBLY

Figure 7.1.
Drain Tank.

ORNL-DWG 68-967

  
     
 
       

HIGH EFFICIENCY FILTER

EXHAUST BLOWER

Arrangement for Adding 233U Enriching Salt to Fuel
38

After each can of enriching salt is added, the salt in the other
drain tank will be transferred back to distribute the uranium throughout
all the salt. Experience in the #°°U critical experiment indicated that
such transfers provide excellent mixing of the salt. If more uranium is
then to be added, half of the mixture will again be transferred to the
second drain tank. However, if the next step is filling of the fuel loop,
no further transfers will be made, Although not shown in Fig. 7.1, two
neutron-sensitive chambers will be suspended in the drain tank cell and
the count rates will be analyzed to monitor the suberitical multiplication
in the tank,

The approach to the critical loading of the reactor will be the same
as in the #75U startup.t®,%° The predicted minimum critical loading at
1200°F is 34.6 kg ©7U. After three T-kg cans have been added to the
drain tank and the salt mixed, the core will be filled and neutron count-
rates determined. This will be repeated after another T-kg can has been
added. The size of two subsequent additions will be determined by the
extrapolation of inverse count rates. The last major addition should
bring the 233y content to about 0.5 kg below the projected minimum critical
loading. At this point the turntable with all the empty cans, the con-
tainment enclosure, and the standpipe assembly will be removed and the
blank will be installed on the access nozzle. The drain tank cell will
then be sealed and the shield blocks will be installed before the reactor

is made critical.

7.3 Small Additions Through the Sampler-Enricher

 

As in the original experiment with #°°U, the final approach to criti-
cality will be made by adding capsules through the sampler-enricher with

the fuel circulating. Some changes will be necessary in handling the

 

19MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNL-3872, pp. 8-9.

203, T, Prince et al, Zero-Power Physics Experiments on the MSRE,
ORNT~4233 (February 1968).
39

capsules on their way to the sampler-enricher because of the neutrons and
gamma radistion from the enriching salt. Latch keys and cables will have
been attached before the capsules are filled and the filled capsules will
be drilled to expose the frozen eutectic before they leave the TURF cell.
Six capsules at a time, each containing about 88 g £33U will be moved to
the MSRE in a shielded, bagged carrier. One capsule at a time will be
renoved from the carrier and lowered directly into the sampler-enricher
enclosure. This transfer can be made without shielding because of the

relatively small amount of salt in a single capsule,

8. SAFETY OF ROUTINE OPERATIONS

Operation of the MSRE involves handling substantial amounts of radio-
activity in fuel sampling, offgas sampling, removal of core specimens, and
maintenance of radiocactive systems. In these there is the potential for
radiation exposure or activity release that could affect the personnel
operating the reactor and delay the experimental program. Therefore, each
of these operations follows careful, formally approved procedures, and
uses equipment designed to provide adequate protection. But at any rate
the hazards of misoperation or improper functioning of equipment are local
in nature and are no different with 22U fuel than they have been in
operations to date. The next two chapters consider conceivable incidents
that threaten serious damage to the reactor or activity releases hazardous
to the public.
L0
9., BREACH OF PRIMARY CONTAINMENT

A gross failure or breach of the primary containment of the fuel salt
would have a serious impact on the program even though personnel would be
protected by the secondary contaimment. The possibility of such an occur-
rence has therefore been reconsidered, taking into account the different
characteristics of the reactor with %>y fuel, the changes in safety cir-
cultry since the original safety analysis, and the condition of the salt

containment after more than two years of operation.

9.1 Damaging Nuclear Incidents

 

The most significant effect of changing from =>5U to #33U fuel is the
change in the nuclear characteristics of the system, particularly the dy-
namies. It was necessary, therefore, to reexamine carefully the response
of ‘the reactor to incidents that could cause nuclear excursions. The
original safety analysis considered in some detail the complete spectrum
of nuclear incidents that could be postulated. As expected, some kinds
of incidents proved trivial in the MSRE because of the nature of the re-
actor. The safety analysis for ©2°U therefore only briefly touches on these
inconsequential cases and focusses primarily on those incidents that could

conceivably have significant potential for damage.

9.1.1 General Considerations

 

The basic neutronic characteristics that determine the dynamic be-
havior of the system are presented in Table 9.1 for both the projected
£33y loading and the current loading with partially enriched 235y, For
the 2357 loading both the predicted and the observed values are listed
for purposes of comparison. The characteristics for 233U fuel were calcu-
lated by the same procedures as the °°U predictions, and the probable
errors are about the same,

The smeller fraction of delayed neutrons from the £2°U fuel may sug-
gest a decrease in the inherent relative stability of the system. But

stability is a function of many system parameters, A detailed analysis=t

 

21g, J. Ball and T. W. Kerlin, Stability Analysis of the MSRE,
ORNL-TM-1070 (December 1965),
Table 9,1

Neutronic Characteristics of MSRE with 233U and 235U Fuel Salt at 1200°F

 

Minimum Critical Uranium Loading™
Concentration (g U/liter salt)
Total Uranium Inventcry (kg)d
Control Rod Worth at Minimum Critical Loading (% &k/k)
One Rod
Three Rods
Prompt Neutron Generation Time (sec)
Reactivity Coefficientsf
Fuel Salt Temperature (°F)~*
Graphite Temperature [{°F) 1]
Total Temperature [(°F)-1]
Fuel Salt Density
Graphite Density
Uranium Concentration®
Effective Delayed Neutron Fractlons
Fuel Stationary
Fuel Circulating
Reactivity Change Due to Fuel Circulation (% &k/k)

2337 mel

15.82
32.8

-2.7
-7.0%
.0 x 107%

-6,13 x 107°
-1,23 x 1075
-9.36 x 1073
4+ 4L7
4+ il
+.389

2,64 x 10°°
1.71 x 1073
-0.093

2355 puel

33.06° [32.85 + 0.257°
207,5°

-2.11 [2,26 * 0.07]
-5.46 [5.59 * 0.07]
2.k x 104

-4, 1 x 107° [(-4.9 £ 2.3) x 107%)
-4.0 x 107°
-8.1 x 107® [-7.3 x 107%)
0.182
0.767
0.234 [0.223]

6.66 x 1073
L hh x 1072
-0.222 [-0.212 % 0,004]

 

®Fuel not clrculating, control rods withdrawn to upper limits.

b
235y only.

cValues in brackets are measured results, The others are predicted,
dBased on 73.2 £t of fuel salt at 1200°F, in circulating system and drain tanks.

®Based on a final enrichment of 33% 225y,
£

EHighly enriched in the fissionable isotope (91.5% 233U or 93% 275U),

At initial critical concentration., Where units are shown, coefficients for variable x are of the form
ok/kdx; otherwise, coefficients are of ‘the form x5k/kbx.

T
Lo

predicted and subsequent experiment522 confirmed that in the MSRE among
the most important parameters are the prompt and delayed temperature feed-
backs, Consequently the larger temperature coefficients of reactivity
with 222U fuel give the system a larger stability margin, particularly at
low powers. Figure 9.1 shows experimental results with Z°5U fuel, namely
observed responses to small step changes in reactivity at different initial
power levels., The importance of temperature effects is evident in the
lighter damping at low powers. Calculations have been made for =°-U
fuel, > using the techniques proved in the 235U operation. These indicate
that relative to the behavior in Fig. 9.1, the damping will be much
greater at the low powers because of the larger temperature coefficients
of reactivity for 77U fuel. The higher gain of the neutron kineties will
appear primarily as shorter natural periods of oscillation. The important
conclusion from these results is that the small perturbations and reac-
tivity fluctuations that cccur in any reactor will not lead to divergent
nuclear behavior that could damage the MSRE. Thus if there is any severe
nuclear transient, it will have to be caused by an independent, large and
persistent reactivity perturbation.

For an incident in which reactivity is added continuously (such as
uncontrolled rod withdrawal), the severity of the power transient is
greater at lower initial power levels. A lower initial power allows the
insertion of more reactivity, and hence the establishment of a shorter
positive period, before the power level gets high enough for power feed-
back shutdown mechanisms (e.g. fuel temperature coefficient of reactivity)
to become effective. Thus, the source power level is an important para-
meter in defining this type of incident. In the MSRE the principal sources
of neutrons are those in the fuel salt. At the time of the 37U startup,
the fuel salt will contain only a small fraction of the fission products

from prior nuclear operation and the photoneutron source will be completely

 

£50. W. Kerlin and S. J. Ball, Fxperimental Dynamic Analysis of the
MSRE, ORNL-TM-1647 (October 1966)

#3MSR Program Semiann. Progr. Rept. Aug. 31, 1967, ORNL-4191, p 6L1.
43

ORNL-DWG 66-1 4
40 66-1028

0075 MW

0465 MW

POWER CHANGE (%)

 

 

 

7.5 MW

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

O 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 34 36
TIME {min)

Figure 9.1. Observed Response of Nuclear Power in Small Step
Changes in Reactivity at Various Initial Powers with‘235U Fuel.
n

overshadowed by the very intense -n source in the £33 fuel. The a-n
source, alone, in the 77U mixture will be a factor of 700 stronger than
the minimum source (also a-n) that was available in the 2357 loading.
Thus, while the lowest power at which the reactor could pass through
criticality was about 2 milliwatts with %7°U fuel, the corresponding value
for the 77U fuel will be more than a watt,

Aside from the basic properties of the reactor, the action of the
reactor safety system is important in limiting the severity of the various
incidents to be considered. Control-rod scrams are actuated by three
reactor parameters: high neutron flux level (over 11.25 Mw if the fuel
pump is running or 11.25 kw if the pump is off), positive reactor period
less than 1 sec, and reactor outlet temperature greater than 1300°F.

The efficacy of a control-rod scram depends on the specific behavior of
the rods. In analyzing the various incidents, we assume that (1) only

2 of the 3 control rods actually drop on request, (2) a delay of 100 msec
occurs between the scram signal and the start of rod motion, and (3) the
control-rod acceleration is 10 ft/sece. All of these assumptions are con-
servative since no control rod has ever failed to drop, the clutch-release
time is about 20 msec, and the actual rod acceleration is about 13 ft/sec®,

Nuclear excursions severe enough to threaten damage produce responses
from the reactor system that are similar in important respects, almost
regardless of the cause of the reactivity excursion. Typically, the re-
activity must increase rapidly until the reactor is well supercritical.
There is then a brief excursion to high power which causes a rapid in-
crease in the temperature of the fuel salt at a rate locally proportional
to the fission distribution. At the same time there is a pressure surge
as the heated salt expands. Inertial effects of acceleration of fluid in
the outlet pipe, momentary increase in friction losses in the pipe and
compression of the gas in the pump bowl are the components of the pressure
surge, with inertial effects predominating during very rapid heating. The
power excursion is brief because of negative reactivity feedback from the
rising temperature and the effects of the control rods being dropped by
the safety system, The termination of the power excursion leaves hotter
salt in the core, which moves on up through the channels, giving up heat
to the graphite, and then mixes with salt from other channels in the upper

head.
1

In the original safety analysis, somewhat arbitrary limits of 50-psi
pressure increase or 1800°F maximum fuel temperature were used to define
accidents that would not be expected to cause damage. In light of the
effects of neutron irradiation on the mechanical properties of the vessel
material, demage mechanisms and thresholds have been reconsidered in more
detail, _ ,

Damage could conceilvably result from either of two mechanisms:

- stresses caused by the pressure surge or fhermal stresses caused by the
rapid change in the temperature of salt in contact with surfaces., Con-
gideration of the details of mechanical design and the system response in
an excursion lead to the conclusion that thermal stresses in the top head
of the reactor vessel near the outlet pipe are controlling. The control
rod thimbles are exposed to greater temperature changes, but they are
relatively thin and transient thermal stresses are lower there than in
the top head. 1In an excursion, the pressure surge is over before the
rising salt temperatures affect the top head. Therefore the effects of
pressure and temperature on the top head can be considered independently
to determine which is limiting.

In calculating the most severe excursion tolerable from the stand-
point of thermal stresses, we chose 25,000 psi as the limit on the com-
puted thermal stresses, The rupture life at this stress level is at
least an hour at temperatures to lhOO°F, and the yield stress for rapidly
applied strains is greater than this at temperatures on up to 1600°F or
so. Wall temperatures will be less than 1U00°F in the limiting cases and
the high stresses will be of briéf duration, so 25,000 psi is a conserva-
tive limit under these accident conditions. Thermal stresses in the top
head were calculated assuming an instantaneous rise in salt temperature
and a heat transfer coefficient of 300 Btu/hr-ft2°°F between the salt and
the head. It was found that for step changes of up to lTSfF stresses were
less than the 25,000-psi limit. (For a 178°F step change, the temperature
difference through the wall reaches a maximum in about 16 seconds and de-
creases to only 10°F in 6 minutes.) An increase of 178°F in the tempera-
ture of the salt leaving the vessel corresponds to an increase of about
343°F in the temperature at the exit of the hottest channel through the
L6

core, Therefore, 1f a nuclear incident does not cause the hot channel
outlet temperature to rise more than 340°F, thermal stresses will not
damage the top head.

The fuel system was designed for 50 psig (more in some parts) with
primary stresses of 3500 psi or less. A nuclear incident that would cause
a temperature excursion approaching the limiting thermal stresses would
produce a pressure surge of less than 90 psi. Thus the pressure alone
would produce primary stresses of no more than 10,000 psi. Therefore, the

thermal stresses are limiting, not the pressure surge,
9.1.2 Uncontrolled Rod Withdrawal

The control rods are the most direct means of increasing the reac-
tivity. The amount of excess reactivity held down by the rods can be as
much as 2.8% sk/k (with 23U fuel) and the speed of the rods is such that
the reactivity can be increased fairly rapidly. There are, of course,
many restrictions on the rods. In the order of increasing relisbility
they include: administrative procedures, control interlocks that inhibit
withdrawal at a 25-sec period and insert the rods if the period reaches
5 sec, and the safety system that scrams the rods. It is conceivable,
although very unlikely, that some combination of misoperation and control
system malfunction could result in a reactivity excursion with the po-
tential for damage but, in such an event, the safety system can be de-
pended on for its design action.

Uncontreoclled rod withdrawal would have the greatest effect if it began
with the reactor subcritical, i.e., with the fission rate very low, and
passed through criticality with all three rods moving in unison through
the region of maximum sensitivity. This accident was analyzed in detail
in the original safety analysis. With ®>7U fuel, however, the power,
temperature and pressure excursions resulting from uncontrolled rod with-
drawal would be different because of differences in rod worth, inherent
neutron source strength, delayed neutron fraction, and temperature coef-
ficilents of reactivity. The rod worth and sensitivity are about 30% higher
with #77U fuel so the reactivity increase would be faster. On the other
hand, the stronger neutron source in the #>7U fuel would tend to bring the

fission rate into the range where safety interlocks (and temperature
b7

feedback) can act at an earlier point in terms of the excess reactivity
that has been introduced. These factors affect the ability of the safety
system to suppress the excursion. If there were no safety action, but only
the temperature feedback to shut down the reactor, the delayed neutron
fraction and the temperature coefficient of reactivity would also be of
great importance. 1In the case of a sustained reactivity ramp such as

this the smaller delayed neutron fraction is an advantage in that the
reactor becomes prompt critical and the power begins to rise rapidly when
there is less excess reactivity that must be cancelled by rising tempera-
tures. The larger temperature coefficient further reduces the‘temperature
rise necessary to turn down the power.

In the analysis of this accident we assumed that the rods would be
poisofiing 2.8% sk/k when the reactor is just critical. (This will be the
condition at the end of the zero-power experiments, in which one rod will
be calibrated over its entire travel.) For this condition, the reactor
would be subcritical by L4.2% 8k/k when the rods are fully inserted, and
would go critical with the three rods at 28 inches withdrawal, slightly
above the center of their range and very near the position of maximum
differential worth. Following continuous withdrawal of the rods from full
insertion, the fission power when criticality is attained would be somewhat
greater than 1 watt, which was used in the analysis. The speed of with-
drawal is 0.5 in./sec, giving a rate of reactivity increase of 0.093%
8k/§/$ec at criticality. If the rods were not scrammed, this rate of in-
crease would continue for approximately 16 sec after criticality, then
would gradually slow down and finally stop when the rods reach their upper
limits at 51 inches withdrawal.

The first step in the analysis of the effects of the uncontrolled
rod withdrawal was to compute the response in the absence of a rod scram,
These results were needed to determine the point at which the reactivity
effect equivalent to two rods scramming should be started in the compu-
tation. Although unrealistic and not directly applicable to the safety
evaluation because the reliable scram of the rods cannot be ignored, the
results of the computation with no safety action are of some interest.
Figure 9.2 shows these results for the reactor fueled with 37U, The
48

ORNL—-DWG 68—968

 

1800

/
L

 

1700

LA
1600 7
MAXIMUM FUEL TEMPERATUR’E)‘

1500 /

_/ OUTLET TEMPERATURE OF
HOTTEST CHANNEL
A

 

 

 

 

TEMPERATURE (°F)

 

1400 /

 

1300 /

 

1200 /A
500

 

 

400

 

 

 

 

 

 

 

 

300

 

 

POWER {Mw)

 

200 -

PRESSURE RISE (psi)

 

TIME (sec)

 

 

{00

 

 

 

 

 

 

 

 

 

 

 

 

 

4 6 8 {0 12 149 16 8
TIME (sec}

Figure 9.2. Results of Uncontrolled Rod Withdrawal with no Safety
Action, 233U Fuel.
49

digital program used for the calculations includes a detailed numerical
treatment of the axial convection of heat by fluid motion during the

power transient.®* Both the temperature of the fluid at the hottest point
in the reactor channels and the outlet temperature of the hottest channel
are plotted in Fig. 9.2. The calculated pressure rise in the core during
the period of very rapid heating is also shown.

To elucidate the effects of the differences in the important neutronic
properties, we performed calculations similar to those of Fig. 9.2, using
235y characteristics but assuning that the reactivity addition rates are
identical (.093% 5k/§/%ec), and also that the power levels at the time of
criticality are identical (1 watt). With these simplifications, the calcu-
lated nuclear excursions differ only because of the differences in delayed
neutron fractions, fuel temperature coefficients of reactivity, and prompt
neutron generation time. The resulting transients calculated for the 235y
fuel are shown in Fig. 9.3. It is evident that the rapid-rise portion of
the transient occurs later in time than with the 227y fuel, since more re-
activity must be added to reach the prompt critical condition. The tempera-
ture excursion in this admittedly fictitious case is greater with =°°U
because more reactivity must be cancelled by this mechanism to stop the
power rise and the temperature coefficient of reactivity is smaller than
when Z33U is the fuel.

Although the calculations shown in Fig. 9.2 indicate that the fuel
temperature and core pressure rise incurred during the rapid portion of
the transient would be inconsequential, it is clear that counteraction of
rod withdrawal would ultimately be necessary to prevent overheating in the
core. Figure 9.4 shows the results of actuating the rod scram mechanism
in the 33U case when the neutron flux level exceeds 11.25 Mw. This action
would be effective in reducing the transient to inconsequential proportions.
Therefore the runaway rod accident will not damage the primary contain-

ment,

 

240, W, Nestor, Jr., ZORCH — An TEM-TO90 Program for the Analysis of
Simulated MSRE Power Transients with a Simplified Space-Dependent Kinetics
Model, ORNL-TM-345 (September 1962)
50

CRNL—DWG 68-969

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

7
1800 7
1700 ’ /
MAXIMUM FUEL TEMPERATURE
1600 7 /
@
e
wt
v
>
2 !
< 41500 ;
wl
a
=
L
-
1400 / /
/-—- OUTLET OF HOTTEST GHANNEL
1300 /
1200 /
500
400 = 20
‘@
a
=45
wl
s
T 300 x 4o
wi
2 ¢ ]
3]
= & 5 ’
= 6
g 200 a0
, 6 8 10
l TIME {sec)
(00 |
|
I .
0
4 6 8 {0 {2 {4 (6 i8
TIME (sec)

Figure 9.3. Results of Uncontrolled Rod Withdrawal with no Safety
Action, 232U Fuel.
51

ORNL~DWG 68—-970

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1230
T
o 1220 MAXIMUM FUEL TEMPERATURE
%
[ ,
L
o[ .
S 1210 1 i
o
AUTLET OF HOTTEST GHANNEL\
1200 o | | | \\\
60 :

7 40
=
x
W
5 fl
& 20

0

a4 6 8 10 12 14 16 18

TIME (sec)

Figure 9.4. Results of Uncontrolled Rod Withdrawal, with Scram at
11 Mw, <330 Fuel. '
52

The present safety system provides an additional margin of safety by
actuating the rod scram when the reactor period decreases below one second.
The digital calculations for the transient without safety action indicate
that a one-sec period is reached at 1.l sec after criticality, the 11.25-Mw
level is not reached until 5.3 sec, only 0.5 sec before the maximum power
is reached at 5.8 sec. The actual time of actuation of the period-safety
scram device is not simply arrived at since it lags the attainment of the
l-sec period by an interval that depends on the ion chamber current and
the rate at which the period is decreasing. However, calculations based
on quite conservative assumptions on the actual time of actuation of the
period scram show that the power transient would be reduced by at least
two orders of magnitude below that shown in Fig. 9.4 and would be quite

insignificant.

9.1.3 Sudden Return of Separated Uranium

 

Two remote possibilities exist for separation of uranium from molten
fluoride fuel salt. If fluorine should be lost from the salt, the
UF;/UF4 ratio would increase, possibly to the point that metallic uranium
would be produced by disproportionation of the UFs to UF4, and U. Second,
if enough moisture or oxygen were introduced, ZrOz would be produced and
precipitate until the Zr**/U*t ratio fell below 2; after which some UOz
would form along with additional ZrOz. (Ref. 25)

Neither of these mechanisms was expected to cause separation in the
MSRE, and experience has supported this expectation, Furthermore, there
is no trend in the fuel chemistry that would indicate that precipitation
of uranium or uranium oxide is likely in future operation. There has been
no detectable loss of fluorine to increase the UFx. In fact, as explained
in Section 1.1.1, UFs gradually decreases during power operation, Analysis
of the fuel for oxides at intervals of approximately one month since the
summer of 1966 has shown that the oxide content has been practically steady
at about 50 ppr. This level is more than a factor of ten below the solu-
bility of ZrOz and even farther below the point at which U0z would begin

 

25W. R. Grimes, MSR Program Semiann, Progr. Report, July 31, 196k,
ORNTL-3708, p. 230.
23

to form. The fluorination of the original fuel charge will not produce
any uranium-bearing precipitate. Verification of proper composition and
state of the salt will be obtained by analyses after the complete pro-
cessing and before 2337 additions begin,

Clearly if present conditions persist through the operation with
233U, as we expect them to, there will be no gradual drift toward uranium
separation. Nor 1s precipitation of U0z because of accidental gross con-
tamination of the fuel with oxygen likely, for the reasons discussed in
the original safety analysis report. If however, despite all precautions,
UCz should separate from the circulating fuel, there would be a detectable
effect on reactivity before a hazardous situation could develop, This
conclusion is based on the analysis that follows.

If U0z solids were to appear in the fuel, being more dense than the
salt, they would tend to accumulate in lower-velocity regions such as the
lower head or the vicinity of the core support lugs. (Detection of de-
posits in these regions is discussed on Page 8L4.) The reactivity worth
of the separated uranium would almost certainly be less than when the
uranium was dispersed in the salt, so the reactivity would tend to go down.
Normally the regulating rod would be withdrawn automatically to keep the
reactor critical. If the separated U0z were by some mechanism suddenly
resuspended in the salt, the flow would carry it through the core, pro-
ducing a reactivity excursion. The magnitude and the time variation of
the reactivity would depend on the amount of uranium returning and the
details of how it entered the circulating stream and the core, We have
analyzed a hypothetical case in which an increment of uranium is instan-
taneously dispersed throughout the 10 ft> of fuel salt in the lower head
of the reactor and then is carried up through the core with the flowing
fuel. TFigure 9.5 shows the time dependence of the added reactivity, cal-
culated from the flow velocities observed in the MSRE hydraulic mockup
and the computed spatial variation of nuclear importance ifi the reactor
vessel, The reactivity effect in this figure is normalized to Ako, the
reactivity effect of the increment of uranium when it is uniformly dis-
persed throughout the 70 ©t° of salt in the fuel loop. In our asnalysis

we varied the size of the increment and the initial power level to ge-
termine the maximum amount of uranium that could be introduced in this

manner without causing damaging temperatures or pressure.
54

ORNL—DWG 68—97¢

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6
5
o
E-4
<]
~
k4
a4
: /T 7\
=
- / | \
53 :
< 1
ul
= |
W i
> i
- 2 '
< |
—’ .
Lt .
- |
{ i ; \
o I
o 2 4 : 6 8 10 12 14

TIME ({sec)

Figure 9.5. Time Dependence of Reactivity Addition due to Sudden
Resuspension of Uranium in Lower Head of Reactor Vessel.
52

Some results of a typical calculation are shown in Fig., 9.6. In
this case Ak, is 0.25% (giving a pesk added reactivity of 1.2% &k/k), the
initial power is 1 kw, the core inlet temperature is 1200°F, and there is

no safety scram of the control rods. ©Shown is the temperature of the fuel

 

at the hottest point in the core (which moves with time) and the tempera-
ture of fuel leaving the hottest channel. During the time when the maxi-
mum temperature is rising most steeply, the pressure in the core increases
briefly by 39 psi.

The magnitude of the temperature and pressure excursions depend on
the amount of uranium resuspended and also on the initial power. Figures
9.7 and 9.8 illustrate this dependence for cases in which the effects of
scramming the rods were not included. Figure 9,7 shows the variation of
temperature rise with the amount of uranium recovered for two initial
power levels: near full power and 1 kw. (The latter is the lowest power
considered because it 1s a factor of ten below the lowest steady-state
power at which the reactor is routinely operated. The reactor is critical
below 10 kw only briefly during startups.) For sizeable recoveries, i.e.
Ako greater than about 0.25% Bk/k, the initial power makes very little
difference in the outlet temperature rise during the excursion., The pres-
sure excursion is worse for lower initial powers, as 1llustrated in
Fig. 9.8 for Ak, = 0.25%. When the safety action of scramming the rods
is taken into account, the picture is completely changed. Figure 9.9
shows the effects of rod scram at 11.25 Mw. Initial power in these cases
is 1 kw, which, with only a level scram, results in larger pressure and
temperature excursions than would occur if the initial power were higher,
Actually scram due To short period would considerably precede the 11.25-Mw
level and the excursions would be much less than indicated in Fig. 9.9,
particularly at the low initial power. Thus, this figure is a conservative
upper limit on the disturbances in temperature and pressure that would
result from recovery of various amounts of uranium.

Based on Fig. 9.9, recoveries up to Ak, = 0.78% at least will not
cause the hot-channel outlet temperature excursion to exceed the 3U3°F
criterion adopted to limit thermal stresses to safe values. Nor would the

pressure excursion be serious at this Ak . So Ak, = 0.78% is a conservatively
56

ORNL~-DWG €8- 966
1700

 

 

1600

 

MAXIMUM FUEL

TEMPEflb

1500 -- A \

1400 7/ OUTLET TEMPERATURE OF N\
/ HOTTEST CHANNEL \

1300

J// ™

2 q 6 8 {0 2 14 16
TIME (sec)

 

 

 

 

TEMPERATURE {°F)

 

 

 

 

 

 

 

 

 

 

 

 

Figure 9.6. Temperature Excursion Caused by Sudden Resuspension of
Uranium Equivalent to 0.25% 5k/k if Uniformly Distributed; Initial Power,
1 kw; No Safety Action. '
57

ORNL-DWG 68-972
120

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

600
500 [——— HOT - CHANNEL OUTLET 100
TEMPERATURE
e
__ 400 // 80
L
< —
un

& o
x ol
& y &
2 300 60 W
<l o
@ =
ul w2
s i ¢
i PRESSURE T

200 40

100 20

0 o
0 0.4 : 0.2 03 0.4 0.5

Ako, EQUIVALENT UNIFORM REACTIVITY (7 A k/k)

Figure 9.7. Effect of Magnitude of Reactivity Recovery on Peak
Pressures and Temperature during Uranium Resuspension Incident with No
Safety Action.
58

ORNL —DWG €68-973

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

PEAK PRESSURE RISE (psi)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

|
4-—-

Figure 9.8. BEffect of Initial Power on Peak Pressure Rise Caused by
Sudden Resuspension of Uranium Equivalent to 0.25% Sk/k if Uniformly
Distributed; No Safety Action.

 

 

INITIAL POWER LEVEL (Mw)
59

ORNL-DWG 68-974

 

 

 

 

 

 

 

 

 

 

 

 

 

500 100
PRESSURE’,/””’/”
~
400 /////,f 80
: /
o -
-~ . e ' ‘5
% 300 0T CHANNEL OUTLET | g9 &
@ / .~  TEMPERATURE w
a
/ -u_)
o @
R W
— A
g ’ >
w 200 ~—— 40 @
o Wl
= / 2
L o
'.._
4100 20
0 0
0.4 0.5 0.6 0.7 0.8 0.9

Akg , EQUIVALENT UNIFORM REACTIVITY (% Ak /k )

Figure 9.9. Effect of Magnitude of Reactivity Recovery on Peak
Pressures and Temperature during Uranium Resuspension Incident with
Rod Scram at 11.25 Mw. Py = 1 kw.
60

safe 1limit on the amount of uranium that could be resuspended suddenly
without causing damage to the fuel contalnment.

Before uranium could be resuspended in the lower head, it would have
to first separate from the fuel, causing a reactfl‘ity decrease, If one
assumes that all the separated uranium comes back, then the reactivity
decrease will equal Ak,. If only a fraction of the separated uranium is
resuspended, as is more reasonable, then the reactivity decrease will have
been larger than Ak,. Thus the separation of enough uranium to cause po-
tential damage by its sudden and complete suspension would be attended by
a decrease of at least 0.78% sk/k.

Separation of uranium would show up as an anomalous change (in the
negative direction) in the residual term in the computed reactivity balance
that is used routinely to monitor nuclear operation.25 Normally the compu-
tation is done at 5-minute intervals by the on-line digitél computer. The
precision of the measurements and computation is about +0.02% 8k/k and an
anomalous change of 0.2% 8k/k would be clearly distinguishable from normal
variations. An administrative safety limit will be imposed to prohibit
nuclear operation when the residual term in the reactivity balance is too
large. The prescribed limit will be something less than 0.78% sk/k, pro-
viding an added safeguard‘against the development of a situation with the

potential for damage due to resuspension of uranium,
9.1.4 Fuel Additions

The possible reactivity, power and temperature effects of a fuel ad-
dition through the sampler-enricher are quite mild because the émount of
uranium and the rate at which it can be introduced into the core are
limited by the physical system.

The enriching capsules for £33y operation will each contain 97 grams
of uranium (88-g #33U). This amount of uranium, uniformly dispersed in
the circulating fuel salt will produce a reactivity increase of 0,12% Bk/k.
This could be compensated by 2 to 3 inches of regulating rod insertion or

by an increase of 13°F in the core temperature.

 

£6J. R. Engel and B. E. Prince, The Reactivity Balance in the MSRE,
ORNL-TM-1796 (March 1967).
61

The rate at which added uranium mixes into the core has been observed
during twenty-seven capsule additions of 2357 with the reactor operating
at full power. Fach time the regulating rod was servo-controlled to keep
the core outlet temperature constant. Figure 9.10 shows a plot of regu-
lating rod position as a function of time during a typical capsule ad-
dition. (The plot was made on-line by the MSRE digital computer and most
of the indicated changes smaller than 0.1 inch are not real shifts of the
rod.) The lag and the transient indicate that the enriching salt metls
and disperses rapidly in the salt in the pump bowl, then mixes into the
circulating stream with a time constant close to the residence time in
the pump bowl. The same behavior was repeated in each of the twenty-seven
additions. The reactivity increase from a capsule of 27U will be about
b timés as great as those from =35y additions, but the regulating rod can
easily keep up with the change., If for any reason administrative control
or the servo system failed and the regulating rod was not driven in, the
temperature and power would start to rise, probably causing a level scram

at 11.25 Mw. Certainly there would be no damage.

9.1.5 Graphite Effects

 

As indicated in the original safety analysis, loss of graphite from
the core is extremely unlikely and would in any event cause no hazardous
nuclear excursion. This conclusion is still valid. Substitution of £°°U
fuel salt for an entire stringer of graphite would cause the reactivity
to increase less than 0.2% 8k/k and this could not occur very rapidly.

Graphite distortion because of irradiation effects would not be
hazardous, but the most recent data on the kind of graphite in the MSRE
indicate that exposure through the proposed operation with £33 should
produce practically no distortion.

Salt penetration of the graphite has proved to be no problem., Speci-
mens removed after exposure during 24,000 Mwh of operation showed weight
gains of 0.03% and only occasional salt penetration into cracks that

happened to extend to the surfaces.®” Tt was calculated from analyses of

 

£75. S. Kirslis and F. F. Blankenship, MSR Program Semiann. Progr.
Rept. Aug. 31, 1967, ORNL~-4191, pp 121-12k,
62

ORNL-DWG 67-10432
39

 

PUMP BOWL

 

 

W

@
o
=
T
w
C
—
m
<

 

 

REGULATING ROD POSITION
(inches withdrawn)
W w
M ~
f

 

 

 

 

 

 

 

 

 

35
O 1 2 3 4 5 6 7

TIME (min)
0=14450 hr APRIL 20,1967

Figure 9.10. Regulating Control Rod Motion during 2337 Fuel Capsule
Addition at Full Reactor Power. -
63

these specimens that the total amount of £7°°U in all the core graphite
was less than L g, a quite inconsequential amount. There is no reason to

expect any change with substitution of 27U for the 75U in the salt.
9,1.6 Loss of Load

Several locad scrams from full powér have shown that sudden inter-
ruption of air flow through the radiator has no ill effects on the system,
The coolant system heats up at a moderate rate and, because the large gas
volume in the drain tank is connected to the pump bowl surge space, there
is no detectable pressure rise. The load scram is accompanied by auto-
matic control action that reverses the rods when the blowers stop until
the nuclear power goes below 1.5 Mw. This rod action prevents any rise
in core temperature, but as shown in the original analysis, the temperature
rise would be at most L4O°F without any corrective action. With the £37y
fuel and its larger temperature coefficient of reactivity, the temperature

rise would be even less,
9.1.7 Loss of Flow

Interruption of fuel circulation produces two immediate effects in
the core: delayed neutron precursors are no longer swept out so the re-
activity tends to increase, and heat is not carried out of the core by
fuel flow so the temperature begins to rise. It was shown in the original
safety analysis that even in the absence of safety action, fuel flow
interruption would cause no damage. The situation is better with Z33U fuel
for two reasons: +the change in effective delayed neutron fraction is only
0.08% instead of 0.21% for 35U fuel, and the temperature coefficients of
reactivity are Jarger. These differences would cause the fission rate
and core heating to decrease more rapidly. From a practical standpoint,
however, there is little or no difference; the safety system would prevent
undesirable excursions with either =2°U or 77U fuel. With the pump off
the scram at 11 kw prevents fission heat from contributing much and without

fission heat the core temperature will rise very slowly if at all.
6l

9.1.8 '"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 because of the negative tempera-
ture coefficient of the fuel. Of course, something of the sort occurs
whenever the power is raised by withdrawing more heat at the radiator.

But physical limitations of the load system make the reactivity rates
moderate and not disturbing even in the absence of automatic rod action.
Figure 9.11 shows system responses in just this situation, where the heat
removal was increased from 2 to 7 Mw as quickly as possible while the
control rods were kept stationary. So for there to be a "cold-slug"
worthy of the name there must be some sort of flow interruption, cooling
and flow resumption. Another important fact about the "cold-slug" acci-
dent is that it cannot happen if the control rods are inserted. The fuel
loading and the rod worth are such that the core could be cooled to the
salt liquidus temperature without going critical if all the rods are fully
inserted.

In principle, a cold-slug could result from interruption of either
the fuel or the coolant flow. Suppose the coolant flow were interrupted
and part of the coolant loop cooled down while the fuel pump continued to
run. Then if the coolant flow were resumed, a cold slug would hit the
heat exchanger and would show up as a fairly fast reduction in core inlet
temperature., But this is prevented by interlocks which scram the load and
stop the fuel pump if the coolant flow drops. A sharper cold-slug could
result if the fuel pump were stopped while the coolant flow continued.

The fuel in the heat exchanger could be cooled down and then be introduced
to the core by festarting the fuel pump. In this case a decrease in re-
activity due to loss of delayed neutron precursors would be superimposed
on the increase due to fuel temperature as flow is resumed. Protection
against a power excursion in this event is provided by an interlock which
requires that all three control rods be fully inserted before the fuel
pump can be started.

For the foregoing reasons we believe a serious cold-slug accident is

practically impossible. But in any case, the nuclear excursion associated
65

ORNL—DWG 68-—-964

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

15
310 - | -
o REACTOR POWER
= S— '
g s ‘;r”/" e e A_1_~_*—”
0 ! 11
1250 l [ l
REACTOR OUTLET |
. §200 et ) : ‘ : E
! | REACTOR INLET
e | N~ |
50 : - ;
g N I
@x N . ?
A o __RADIATOR INLET |
=2 X . .
- “ ‘ { “
1 ' : _]F__#
1050 <T__RADIATOR OUTLET ~
. | E
| |
1000 .
40 80 120 160 200 240 280 320
TIME (sec)

Figure 9.11.

System Response to Load Increase from 2 to 7 Mw at
Maximum Rate.
€6

with an incident of this type should not be damaging to the system. This
conclusion is reached by the following argument. The total volume of fuel
salt in the circulating system outside the reactor vessel furnace is only
12 £t® — about half the volume of salt in the passages in the graphite
core. If this much subcooled salt were pumped through the core, the
shape of the reactivity-time curve would be very nearly that shown in

Fig. 9.5 for the resuspended uranium. If the entire core were suddenly
filled with fuel at 900°F, which is only slightly above the liquidus
temperature, the excess reactivity would be 1.8% sk/k. This equals the
peak reactivity from resuspension of uranium equivalent to 0.37% Bk/k
when uniformly dispersed. As shown in Section 9.1.3, a uranium-resuspension
accident of this magnitude would not cause damage. Therefore, the possi-
bility of a cold slug causing a damaging nuclear excursion can be dis-

missed.

9.1.9 Filling Accident

 

The original safety analysis report included detailed analyses of
accidents in which the reactor became supercritical while the core was
being filled with fuel salt under various abnormal conditions, The only
acclident of any conseguence was found to be one in which the core was
filled with fuel with a uranium concentration substantially higher than
normal, The possibility of such an accident was suggested by the equi-
librium crystallization path of the fuel mixture in which the last phase
to freeze is rich in uranium. It was postulated that there was partial
freezing of the salt in a drain tank followed by physical separation of
the solid and liquid phases, then a series of operator and equipment
malfunctions,
| Since the original analysis, there have been experiments duplicating
as nearly as possible the situation in the drain tanks during very slow

cooling and freezing.®® Results of these showed that the degree of

 

£8MSR Program Semiann, Progr. Rept. Feb. 28, 1965, ORNL-3812,
pp 126-127,
67

concentration and separation originally postulated are unrealistic and
led us to the conclusion that a seriocus filling accident will not occur
even if other malfunctions are assumed.

Although no credivle filling accident threatens damage, the adminis-
trative procedures to prevent any kind of abnormal fill and the automatic

actions to terminate any such fill will be retained.
9.1.10 Afterheat

As discussed in the original safety analysis report, problems associ-
ated with decay heat in the MSRE are quite moderate and require no rapid
emergency action, The afterheat in the proposed operation will be less
than was considered in the original safety analysis, because heat trans-
fer has limited full power to about 7.5 Mw rather than the 10 Mw that was
anticipated. (Because of differences in fission product yields, the
afterheat from 233U fuel will be about 7% greater than from “7°U fuel
operated at the same power.) Testing has verified that the cooling sys-
tem on the drain tanks has ample capacity and that the salt can be
drained reliably, but that a drain is not essential to afterheat removal
because heat losses from the reactor vessel are enough that overheating
can be prevented simply by turning off the electric heat to the furnace,

Therefore, afterheat poses no threat to the primary contalnment.

9.1.11 Criticality in Drain Tanks

 

The nuclear reactivity of the unmoderated fuel salt with the partially
enriched 23°U currently in use 1s somewhat lower than it will be when =33y
is substituted., In the original safety analysis, the drain tanks were
shown to be critically safe even with the assumption of some highly un-
realistic conditions to increase the reactivity. Because of the greater
reactivity of the Z2>U mixture these assumptions have been reevaluated
in terms of conditions that are physically attainable,

The most reactive situation in a drain tank would occur if the entire
fuel charge were stored in one drain tank and allowed to cool to room
temperature. An important increase in reactivity would result if water
were supplied for neutron moderation and, since the drain-tank cooling

thimbles use water, it must assumed that the thimbles will be full of
68

water for the worst condition. An external water reflector around the
tank would also increase reactivity but this cannot be attained. The

only water available to the drain tank cell is that in the treated water
system and the total amount that could collect in the cell would not
reach even the bottom of the lower head of the drain tank., The reactivity
of a full drain tank at room temperature is sensitive to the bulk density
of the frozen salt, For small amounts of salt, this density has been
estimated to be 1.1k times that of liquid salt at 1200°F. Although the
bulk density for a large mass of salt will be less because of pores and
cracks, to be conservative we used the density named above in calculating
reactivity. In the calculations we did not include any effect of uranium
inhomogeneity because the rapid heat removal rates during freezing associ~
ated with the presence of water in the thimbles would produce frozen salt
that is homeogeneous from the nuclear standpoint.

Under normal fuel storage conditions, with all the salt in one drain
tank at 1200°F and no water in the cooling thimbles, the neutron multipli-
cation factor was calculated to be 0.85. Using the most reactive conditions
(tank at room temperature, water in the thimbles) calculations gave a
multiplication factor of 1.00 with all the salt in one tank. If the salt
is equally divided between the two drain tanks that are available,

Kepp = 0.88 at room temperature with water in the thimbles.

Because of the advantage in dividing the fuel, if freezing of the
salt is ever anticipated, it will first be divided equally between the
two tanks. In an emergency shutdown from power operation, the fuel salt
automatically drains to both tanks, so the salt would be left in a criti-
cally safe condition even if the operators had to leave immediately, before
the salt drained. Only if there should be an unplanned, extended building
evacuation during a shutdown in which all the salt is stored in a single
tank could there be a chance of criticality in the drain tank. In this
case it is possible that an electric power failure would allow water to
be admitted to the thimbles and the salt to freeze. Whether or not criti-
cality would be reached is questionable because there is some conservatism
and uncertainty in the calculated value of 1.00 for keff at room tempera-

ture. But criticality is conceivable, to say the least.
69

Since criticality in a drain tank cannot be absolutely ruled out
under all possible circumstances with 22U, some evaluation of such an
event is in order. If criticality did occur, it would not be until the
fuel salt was frozen and at a rglatively low temperature. At the low
temperature, the rate of temperature decrease and, hence, the rate of
reactivity increase as korp approached 1 would be very slow. Since the
mixture contains an intense inherent neutron source from “>2U and its
daughters, no nuclear excursion would result. Instead, the nuclear power
would rise slowly to a level just sufficient to maintain the salt at the
critical temperature. The drain tanks are inside the reactor secondary
containment with sufficient bilological shielding, so no radioclogical
hazard would exist in the reactor building from that source. Thus, it
would be possible to reenter the reactor building to stop the reaction
by remelting the salt and toc distribute it between the two tanks for safe

storage.

9.2 Damage from Other Causes

 

The original safety analysis considered several possible causes of
damage to the primary contaimment other than nuclear incidents. These
other causes are not affected by the change to 2337 fuel. However, the
system has now been operated and there is experience pertinent tc each

damage mechanism. Therefore, they are re-examined here.
9.2.1 Thermal Stress Cycling

In a normal heating and cooling cycle of the salt systems, tempera-
tures may chenge from as low as TO°F to as high as 1300°F. The piping
was laid out with sufficient flexibility to avoid excessive stresses due
to expansion and contraction., Analyses indicated piping stresses were
TOOO psi or less except at a nozzle on the primary heat exchanger.
Strain gage measurements in September 1965 showed maximum stresses there
were 15,500 psi. Iven this is below the point at which stress cycling
could be damaging. Thus stresses caused by reaction forces from piping

are inconsequential.
70

Thermal gradients do prcduce high stresses and plastic strains in
some components, notably the freeze flanges,

A steep radial temperature gradient is inherent in the design of the
freeze flange. During thermal cycling this steep gradient and the thermal
inertia of the massive flange result in plastic strains at the bore.
Stresses are highest at the bore and decrease rapidly with increasing
radius so that damage due to thermal cycling would first appear as shallow
cracks at the bore. The flanges were analyzed on the basis of low-cycle
fatigue to predict the number of cycles of various kinds at which cracking
would be expected to begin. The calculated numbers of cycles were then

reduced by a factor of ten to obtain the feollowing permissible numbers of

cycles,
heating cycle 160
£ill cycle 58
power cycle (coolant flanges) 550

Although, as described below, these numbers are used to prescribe limits
on the operating life, the flanges should survive considerably more cycles
without consequential damage. First, the safety factor of ten on the
calculated cycles is conservative, ©Second, the initial cracking would be
superficial and many cycles would be required to propagate a crack through
the pipe wall.

An accurate history of thermal cycles is maintained and the effects
of the different kinds of cycles are combined by summing the fractions
of the permissible number of each kind of cycle that have been sustained.
Through the startup in September 1967, the fuel freeze flanges had reached
69% of permissible life on this bvasis. The anticipated operations, in-
cluding 233 startup experiments, will not exhaust the specified permissible
life,

As a supplement to the fatigue calculations for the freeze flanges,
a test flange was subjected to 103 combined heating and filling cycles,
Although the permissible number of cycles, calculated as for the reactor
flanges, was only 30 cycles, dye-penétrant inspection showed no evidence
of damage after the 103 cycles. This test facility has been reactivated

for continued cycling of the flange.
T1

No component other than the freeze flanges will approach a limit on
thermal cycles during the proposed operation of the MSRE. The component
with the next shortest life is the coolant pump and its predicted service
1ife is ten times that of the freeze flanges.

In summary, failure of the primary contaimment because of stress

cyecling does not appear credible.

9.2.2 Freezing and Thawing Salt

 

As the fuel salt melts its specific volume increases by at most
5 percent. Conceivably this could result in damage if a portion of salt
thawed and the expansion were confined by frozen plugs on either side.

Salt is routinely frozen and thawed in the freeze valves, but the
desigh is such that they are not damaged. The pipe in the valve is flat-
tened, permitting some expansion if required. The frozen plug is kept
short and thawing is done from the ends toward the center of the plug.
The adequacy of the design from this standpoint was proved by thorough
testing of prototypes. (Stresses are so low that fatigue is no problem
in the freeze valves.)

The MSRE fuel salt system is provided with heaters, emergency power
supply, and insulation to minimize the chances of accidental freezing. No
fuel salt has been frozen unintentionally since it was charged into the
MSRE. Furthermore, in case of freezing in a pipe it would in general be
possible to heat and thaw from the ends rather than in the middle., Thus
there is no significant risk of damage due to freezing and thawing the
fuel.

There is practically no change in the density of the flush salt or

coolant salt on thawing and thus no threat to the containment.
9.2.3 Excessive Wall Temperatures

Since the entire salt system of the MSRE is electrically heated with
an installed heater capacity somewhat greater than that actually required,
the possibility exists of heating the system to abnormally high tempera-
tures. Local overheating of the system by the electric heaters is most
probable when the system is being heated while empty. The possibility
of local overheating is greatly reduced when the system is salt-~filled

and local overheating is virtually impossible when salt is circulating.
T2

The operational high temperature limit for the reactor is 1300°F and
the following steps have been taken to avoid exceeding this limit. Me-
chanical stops are placed on all the heater coatrols to limit the heater
power to 110% of the power requirement for 1200°F. This in itself should
Linit the temperature to about 1300°F. Thermocouples are located under
each heater assembly, and the wall temperatures of the system are monitored
contirmously by the temperature scanner which gives an alarm if any of the
thermocouples exceed the preselected limit., In addition, the on~line com-
puter monitors numerous other thermocouples and gives an alarm if any of
the thermocouples exceed the glarm pecint, The neater settings are rou-
tinely checked and recorded every 4 hours so that any significant changes
in heater power would be promptly noted aund corrections could be made if
required.

Fastelloy-N has good high temperature strength properties, and the
design stress of the reactor system was selected on the basis of the 1300°F
creep rate. Actually much higher temperatures could be tolerated on a
short term basis. The yield strength at 1LE00°F is about 20,000 psi, and
the stresses in the MSRE are sufficiently low that temperatures of this
magnitude could be safely tolerated for a short time. Tests loops of
Hagtelloy-N have routinely operated for relatively long periods of time
at 1500°F and the reactor vessel was given a 100-hour heat treatment at
1400°F tc improve the mechanical properties of the closure weld.

Zn conclusion, the mechanical heater stops prevent the sgystem from
being overheated to actually dargerous temperatures. The possibiiity of
exceeding the 1300°F 1limit is minimized by the righ temperature alarms on
the scanner and computer and by the close surveillance of the temperatures
and heater settings by the operating personnel,

In addition to the possibility of overheating by the electric heaters,
excessively high temperatures in some areas might also occur as a result
of nuclesr radiation heating. The two areas of special interest are the
reactor vessel and the upper surface of the fuel pump tank.

Calculations indicated that heating of the reactor vessel and in-
ternal structures by gamma rays from fissions and fission products dis-

trivuted normally in the fuel salt would produce only trivial temperature
13

elevation and thermal stresses. Observation of thermocouples on the
outside of the reactor vessel have shown no effects of any consequence,
Particularly close attention is given to thermocouples on the lower head
and adjacent to the core support lugs, for it is here, if anywhere, that
any solids in the salt would accumulate., The on-line computer continuously
monitors the temperature difference between the reactor inlet line and
six locations on the lower head and four locations at the core support
lugs. An alarm is given by the computer if the temperature differences
exceed specified limits, These temperature differences have been 1.5
and 2.1 °F/Mw respectively for the lower head and core support lugs, and
there has been no significant change since the beginning of power opera-
tion to suggest heat generation from the buildup of a deposit,

Conservative design calculations indicated that radiation from fission
products in the gas space in the fuel pump could cause seriocus heating of
the upper surface of the tank. Thus the pump design included an air-
cooling shroud to limit temperatures and produce a distribution giving low
thermal stresses., OSustained operation of the reactor at power proved that
the temperature distribution was satisfactory with no forced air cooling
and this was adopted as the normal mode of operation. When £33 1s sub-
stituted in the fuel, the heat that must be removed through the upper pump
tank surface should increase by about 50 percent. This is a consequence
of the higher yield of the short-lived krypton isotopes from =337 fission
(about a factor of two over £35y yields). ©Some cooling air flow through
the shroud may be required, but a moderate amount, well within the capacity
of the system, will be adeguate to maintain tank temperatures at a suitable
level,

In coneclusion, radiation heating is not a credible cause of damage

to the primary containment.
9.2.4 Corrosion

There is abundant evidence that corrosion has not and will not weaken

the MSRE piping and vessels.
Th

First, there is the basic character of the corrosion process in
molten fluoride systems.®® No film of oxidation products develops in
these systems so corrosion protection does not depend on the integrity
of such a film, Instead corrosion is controlled by the thermodynamic
driving forces of the corrosion reactions. The fluorides that constitute
the salt are much more stable than the structural metal fluorides, so
there is a minimal tendency to corrode the metal. Thus the principal
source of corrosion becomes the trace impurities, such as HF, which can
be ceontrolled.

Corrosion data on Hastelloy-N in LiF-BeF- based salts have been
generated in thermal- and forced-convection loops and in inpile capsules,~©
Operation of 37 thermal-convection loops (17 for a year or more) demon-
strated the compatibility of Hastelloy-N with various fluoride mixtures.
Subsequently 15 forced-convection loops were operated at temperatures
from 1200°F to 1500°F, with a temperature difference of 200°F (except for
one loop with 100°F AT) for periods up to 20,000 hours. Metallographic
examination of surfaces exposed at 1200 to 1400°F showed no evidence of
attack during the first 5000 hr of operation; at longer times a thin (less
than 0.5 mil), continuous intermetallic layer was faintly discernable,

At 1500°F, the surface layer was depleted of chromium, as indicated by
moderate subsurface void formation to a maximum depth of 4 mils after

6500 hours. Numerous inpile tests involving capsules and forced-circulation
loops have shown no effect of radiation on the corrosion behavior of
Hastelloy-N in the fluoride salts,”t

Corrosion in the MSRE has been monitored by frequent analysis of the
salts for corrosion products and by examination of two sets of specimens

taken from the core, the first in August 1966 and the second in May 1967.

 

2%W. R. Grimes, Chemical Research and Development for Molten-Salt
Breeder Reactors, ORNL-TM-1853, (June 1967) pp. 37-L5.

°CH, E. McCoy, Jr. and J. R. Weir, Jr., Materials Development for
Molten-Salt Breeder Reactors, ORNL-TM-185L, (June 1967) pp. 18-26.

*W. R. Grimes, op.cit., pp. L46-56.
()

Chromium in the fuel salt is the best indicator of corrosion of the
Hastelloy-N, since corrosion selectively attacks the chromium and the pro-
duct, CrFs, is quite soluble in the fuel.”® Therefore the MSRE fuel has
been sampled and analyzed for chromium at least once a week during opera-
tion. Chromium analyses of fuel salt samples taken from the reactor over
a period of more than two years are shown in Figure 9.12. The increase
from 38 to T2 ppm corresponds to 170 g of chromium, which is the amount in
a 0.2 mil-layer of Hastelloy-N over the entire metal surface in the circu-
lating system. However, the data suggest that most of the chromium ap-
peared in the salt while it was in the drain tanks between runs. The
indication is that chromium has been leached from a 0.6 - 1.2-mil layer in
the drain tanks and from only 0.08 mils in the circulating system, There
is some reason to believe that an extremely thin layer of nocble-metal
fission products on loop surfaces is responsible for the virtual non-
existence of corrosion there. But in any event, the generalized corrosion
has been guite low.

The indication of extremely low corrosion in the loop was substanti-
ated by the condition of surveillance specimens exposed in thé core. The
first set was exposed to salt for 2800 hours, during which time the reactor
power generation amounted to 7800 Mwhr. None of the Hastelloy-N specimens
showed any evidence of corrosion.”> A second set, in which the Hastelloy-N
was slightly modified by the addition of 0.5% Ti or 0.5% Zr, was exposed
to salt for 4300 hours and 24,900 Mwh. The metal surfaces were only
slightly discolored, and metallographic examination showed no appreciable
corrosion.>*

Operation of the MSRE has also provided information on the corrosion

of the Hastelloy-N vessels and piping from the outside, that is by the

 

*2W. R. Grimes, op.cit., pp. LO-k3.

*3H. E. McCoy, Jr., An Evaluation of the MSRE Hastelloy-N Surveillance
Specimeng — First Group, ORNL-TM-1997, (November 1967) p. 50.

“4MSR Program Semienn. Progr. Rept. Aug. 31, 1967, ORNL~-4191, p. 203.
R
o

ORNL-DWG 67-11364

 

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o

 

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o

 

 

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o O

 

 

 

 

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4 8 12 6 20 24 28 32 36 40

MEGAWATT HOURS (x103)

Figure 9.12. Chromium in Fuel Salt Sample.

9
cell atmosphere. In June 1967, a set of specimens was removed from the
furnace around the reactor vessel after 11,000 hours at high temperature,
covering all the power operation up to that date. There was no evidence
of nitriding, and the maximum depth of oxidation was only about 3 mils, >
Visual examination of a control rod removed in January 1967 also disclosed
only moderate oxidation of the surface in the 8500 hours the rod operated

at high temperature in its thimble in the core.

9.2.5 Radiation Damage to Conteiner Material

 

The effects of neutron irradiation on Hastelloy-N were discussed in
Section 1,1.2 of this report. In summary, the irradiation effects at MSRE
temperatures are a reduction in tensile ductility and a reduction in the
fracture strain during stress rupture tests. The rupture life was also
reduced at high stress levels, but the available data at lower stresses
show only a small reduction (Fig. 1.l). The ultimate strength, yield
strength, and creep rate were not significantly affected in regard to MSRE
operstion.

The primary stress levels in the reactor vessel during normal opera-
tion are well below the range of the tests on irradiated material, but
extrapolation of the data indicate that the decrease in rupture life should
not be enough to shorten the service life below that contemplated for the
£33y operation. Calculations have indicated that the secondary stresses,
thermal stresses, and stresses from piping reactions are also satisfac-
torily low.?® Significant transient thermal stresses do not develop
during normal operation because of the relatively thin sections and the
slow thermal response of the reactor system to power and load changes.
Transient thermal stresses would have to exceed the yield point before the
life of the reactor would be reduced, and even then the stresses would be

relieved without actual failure of the vessel.

 

3S5Tbid.,

“6R. B. Briggs, Effects of Irradiation on the Service Life of the
Molten Salt Reactor Experiment, Trans, Am. Mucl. Soc., 10(1): 166-167
(June 1967).

 
T8

Since normal operations and the credible reactivity accidents (with
safety system action) do not produce high stresses in the reactor vessel,
we believe that the vessel can be used safely, decpite radiation effects,

for the proposed life of the experiment.
10, RELEASE FROM SECONDARY CONTATINMENT

Ultimate reliance for protection of the public from the consequences
of any credible accident in the MSRE is placed on the secondary contaimment
that surrounds the fuel salt system. The original safety analysis assumed
a containment leak rate that could probably be attained and assayed the
possibility of damage that would significantly increase the leak rate, Then
the analysis considered the situation that would place the most stringent
demands on the containment — the simultaneous spillage of gross quantities
of the fuel salt and water in the reactor cell. Calculations of leakage
and dispersal of fission products indicated that the secondary containment
adequately limited the consequences of this hypothetical event. It was
largely on this basis that the USAEC concluded that the MSRE could be
operated "without undue risk to the health and safety of the public.”

Periodic tests of the secondary contalrmment at high pressure have
invariafily shown lower leak rates than were assumed, and there has been
rothing to reduce confidence in the strength and reliability of the con-
tainment. Therefore from the standpoint of contaimment adequacy, any
differences between the operation originally approved and the proposed
operation with £330 fuel must lie wholly in the amounts of fission pro-
ducts that must be contained. Theré are differences, partly because the
yields from®33y fission are different and partly because the power is
limited to 7.5 Mw instead of the 10 Mw originally contemplated. Detailed
calculations show that there will be less of each of the important cate-
gories of fission products in the salt than was considered in the original
safety analysis. (There will be 9% less iodine, 27% less bone-seekers and
11% less kidney-seekers.) On this basis, then, we assert that the con-
clusion of the USAXC is still valid and the proposed operation will not
entail undue risk to the health and safety of the public.
.

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79

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b,
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