ORNL-TM-3041.txt

Contract No, W=74-5-eng-26

Molten-Salt Reactor Program

 

ORNL-TM~-3041

 

MSRE OPERATOR TRAINING AND OPERATING TECHNIQUES

R. H. Guymon

AUGUST 1973

NOTICE

This report was prepared as an account of work
sponsoted by the United States Government. Neither
the United States nor the United States Atomic Energy
Commission, nor any of their employees, nor any of
their contractors, subcontractors, or their employees,
makes any warranty, express or implied, or assumes any
legal liability or responsibility for the accuracy, com-
pleteness or usefulness of any information, apparatus,
product or process disclosed, or represents that its use
would not infringe privately owned rights.

 

 

OAK RIDGE NATIONAL LABORATORY

Oak Ridge, Tennessee 37830
operated by

UNION CARBIDE CORPORATION
for the

U, S. ATOMIC ENERGY COMMISSION

-
-
L

=3
i
iii

CONTENTS
Chapter Page
ABSTRACT . & v v v v 4 v v 4 6 s v o o o o & o o o o o o v
ACKNOWLEDGEMNTS ¢ » * . » o . . e . . * . . * . . * . * Vi
1. INTRODUCTION . . + & v 4 &« o o o o o o o o o o o o o o & 1
2., DESCRIPTION OF THE MSRE
3. CHRONOLOGY . . +v + & & o o & o o o o s o o o
4. OPERATING PERSONNEL . . &+« v & o & o o o o
5. TRAINING . . v v 4 v ¢ o 4 o o o o o o o o o« s o o o o 13
Initial Training . . . e e e e e e 15
Precritical Training and Certlficatlon « 48 e e e 19
Prepower Training . . . . e e e e e e e e e e e e 24
Training of Chief Operators e e e e e s e s e e e e 27
Training Before 233y Startup . . . v 0 v e e e e e e 27
Other Training . « ¢« ¢ &+ ¢ ¢« ¢ o o« & o o o o o o o 28
Training of Replacements. . . . + ¢ « ¢ ¢ « « ¢ « o« & 30
Equivalence to AEC Examination . . . . . . . . . . . 30
6. OPERATION OF THE REACTOR + . v v ¢« & ¢ ¢ o« v o o o o o & 31
Planning . . . e e . s e e e e 31
Instructions to and Communlcatlons w1th Operators . . 32
Data Taking and Record Keeping. . . . . . . . . . . . 33
7. APPROVAL OF MAINTENANCE AND DESIGN CHANGES . . . . . . . 34
8. INFORMATION AVAILABLE TO OPERATORS . . « + v & « o« « + 35
Design and Operations Report. . . . « « ¢ ¢« ¢ o o« « & 35
Other REPOTES « & v v « & + o & & s o s o o o o o o 39
Drawings . . « + o« ¢ o 4 o« e 4 s 4 e e a4 e e e e e 39
Miscellaneous . + + « « o + o s+ o o & 4 2 e s s e . 39
9. RECOMMENDATIONS. . « ¢ « ¢ 4 o o o o o o « s o o s o « 40
REFERENCES . v & & ¢ ¢ o o 4« o o s o s o s s o s o o o s 51
Appendices
A. Schedule for Initial Training. . . . « « « + « ¢« « + . . 53
B. 1Instructions for Flowsheet Checkout. . . . . . . . . . . 57
C. Schedule for Precritical Training. . . « « « ¢ « &+ &+ & 59
D. Precritical Examinations e h e e s e e e e e e e e e 63
E Schedule for Pre-Power Training. . . . . . ¢« « « « .+ . . 75
F. Pre~Power Examinations . . . C v e e e e e e e e e s 77
G. Chief Operator Training Schedule t e e e e e e e e e s 79
H. Chief Operator Review Test . . + & v v ¢ + ¢ o o o o« o« 81
I. MSRE Operator's Examination — May, 1969. . . . . . . . . 83
J. Examples of Instructions and Communications. . . . . . . 99
K. Examples of the Control-Room and Building Log o s s e s 121
L. Forms Used in Making Changes . . + ¢ o ¢ o ¢ o o o « + & 131
M. Example of MSRE Daily Report . + o o ¢ o ¢ o o o o o o & 137
ORNL-TM-3041

MSRE OPERATOR TRAINING AND OPERATING TECHNIQUES

R. H. Guymon

ABSTRACT

The MSRE was a unique, fluid-fuel reactor that operated successfully
at ORNL from 1965 through 1969. MSRE operators and supervisors, mostly, -
veteran ORNL employees, were trained and examined within the Molten-Salt
Reactor Program. Formal training sessions were held at the beginning of
prenuclear testing, just before initial criticality and before the ap-
proach to power. Training of replacements and retraining of operators
was a continuing effort. This report describes the training, the infor-
mation provided for use by the operators, operations planning and adminis-
tration, and the use of procedures. Recommendations by the author (former
MSRE operations chief) conclude the report.

Keywords: #*MSRE + *®operation + *procedures + *training +
administration + communications + examinations + operators +
qualifications + reactors + startup -+ testing
vi

ACKNOWLEDGEMENTS

The author wishes to acknowledge the efforts of S, E. Beall and C. E.
Wolfe in supplying very able operating personnel and express appreciation
to the entire crew for their contributions and fine team spirit during
training and throughout the operation of the MSRE. Special thanks is due
to P. N. Haubenreich for his leadership of the project and his suggestions
on this report.

J. R. Engel and B, H, Webster deserve much of the credit for the co-

operation between operations, experimentation, and maintenance.
 

1. INTRODUCTION

The objective of the Molten-Salt Reactor Program was the development
of a practical breeder reactor in which the fissile and fertile materials
are incorporated in a molten mixture of fluoride salts.l! A major step in
that development was the Molten-Salt Reactor Experiment (MSRE), a 7.4-MW
reactor that operated at ORNL from 1965 through 1969. The main purpose
of the MSRE was to demonstrate that the desirable features of the molten-
salt concept could be embodied in a practical reactor that could be con-
structed and maintained without undue difficulty and one that could be
operated safely and reliably. This purpose was accomplished, as the MSRE
operated for long intervals and with a high overall availability.?s3

Among the factors contributing to the success of the MSRE were the
very thorough preparations for operation, including careful selection and
training of operating personnel, and the disciplined manner in which the
operation was planned and conducted. It is the purpose of this report to
describe these aspects of the MSRE. Only brief descriptions of the physi-
cal plant and its history are included, since these have been widely re-

ported.

2. DESCRIPTION OF THE MSRE

The MSRE was a single-region, graphite-moderated, thermal reactor
which produced heat at a rate of 7.4 MW(th). The fuel was UF, (originally
235y and later 233U) in a carrier salt of LiF-BeF,~ZrF,. At the operating
temperature of 1200°F, this salt was a liquid with good physical proper-
ties ~~ viscosity, 8 centipoise (about like kerosene); density, 135
1b/£ft3; heat capacity, 0.57 Btu/lb °F; and a very low vapor pressure .of
<0.1 mm Hg. The liquidus temperature of the fuel salt was 813°F, so the
equipment and procedures had to be designed to prevent freezing. The salt
also had to be protected from contact with air to minimize corrosion and
accumulation of oxides.

The design conditions for full-power operation are shown in the flow

diagram (Fig. 1). The general arrangement of the plant is shown in Fig. 2.
 

 

||

STACK

)

FAN

ABSOLUTE

OFF-GAS
HOLDUP

FILTERS
BLOG.
T2 VENTILATION
© FROM
==t . COOLANT
SYSTEM

 
 
 

PUMP I ‘ o ) LEGEND

 

fl.,
|
I

 

 
  
 

 

—;—-uu- COOL ANT

 
 
 

 

 

200,000 ctm

FREEZE VALVE (TYP)

 

 

 

 

 

 

MAIN
BED -

 

AUX.
BED -

 

PP

 

¥

 

 

 

WA';ER STEAM

 

 

 

 

WA'I;ER STE

 

 

 

ORNL-DWG 6511410

e FUEL SALT

SALY

— HELIUM COVER GAS
———— RADICACTIVE OFF -GAS

 

 

   
 

 

 

 

 

 

 

 

 

 

 

 

 

| i
SODIUM
FLUORIDE BED

Fig. 1. MSRE Flow biagram,}.

 

COOLANT
DRAIN
TANK

 

 
ORNL-DWG 63-1209R

     
  
 

j::}TREMOTE MAINTENANCE : j
— /| CONTROL. ROOM

REACTOR CONTROL
ROOM

A7

1. REACTOR VESSEL 7. RADIATOR
- - 2. HEAT EXCHANGER 8. CCOLANT DRAIN TANK
3. FUEL PUMP 9. FANS
4. FREEZE FLANGE 10. FUEL DRAIN TANKS
5. THERMAL SHIELD 11, FLUSH TANK
Lo 6. COOLANT PUMP 12. CONTAINMENT VESSEL

13. FREEZE VALVE

Fig. 2. General Arrangement of MSRE,
In the reactor primary system, the fuel salt was recirculated by a sump-
type centrifugal pump through a shell-and-tube heat exchanger and the re-
actor vessel. The heat generated in the fuel salt as it passed through
the reactor was transferred in the heat exchanger to a molten LiF-BeF;
coolant salt. The coolant salt was circulated by means of a second sump-
type pump through the heat exchanger and through the radiator where the
heat was dissipated to the atmosphere. The rate of heat removal was con-
trolied by using either one or two blowers and by adjusting doors in the
radiator air stream and dampers in a bypass stream. Drain tanks were pro-
vided for storing the fuel and coolant salts at high temperature when the
reactor was not operating. Drain and transfer lines included freeze valves,
where salt could be frozen or thawed to block or permit flow. The salts
were drained by gravity, and were transferred back to the circulating sys-
tems by pressurizing the drain tanks with helium. Electric heaters were
used to keep the salt molten in the tanks and to preheat the piping system
before filling. Diesel~powered generators provided emergency power for
heaters and salt pumps. The salt systems normally operated with the cover
gas at 5 psig.

Most of the fission products remained in solution in the fuel salt.
One class, known as "noble metals,'" deposited on graphite and metal sur-
faces. The gaseous fission product s, krypton and xenon, were removed con-
tinuously from the circulating fuel salt in the fuel pump tank. There they
transferred from the liquid to the helium cover gas and were swept out of
the tank by a small purge stream. This stream passed through holdup piping,
a filter, and long, water-cooled beds of activated carbon. The passage of
krypton and xenon through these beds required days to weeks, by which time
all the radioisotopes but the ®3Kr had decayed so that the stream could be
safely diluted with air and discharged to the atmosphere.

The fuel and coolant systems were provided with equipment for taking
samples of the molten salt while the reactor was operating at power. The
fuel sampler was also used during operation for adding small amounts of
uranium or plutonium to compensate for burnup. An array of graphite and
metal specimens at the center of the core was removed and replaced peri-

odically during shutdowns.
The negative temperature coefficient of reactivity of the fuel and
graphite moderator made nuclear control of the system very simple. How-
ever, three control rods were provided for adjusting temperature, compen-
sating for buildup of fission products, and for shutdown.

Instrumentation was installed to adequately monitor all variables and
to automatically handle any anticipated malfunctions. This, together with
collection of much data by an on-line computer, made it possible to operate
with a minimum operating crew,

Because of the fission-product activity, the fuel-salt equipment was
contained in heavily shielded cells that were kept sealed except when main-
tenance was being done with long-handled tools through openings in the cell
roof. The fuel off-gas system was also shielded and contained. Radiation
zones were established around the coolant salt system and a few other lo-
cations, but normal operation of the reactor did not require entry into
these zomnes.

Although not shown in Fig. 2, the plant included a simple processing
facility for treating full 75-ft3 batches of fuel salt with hydrogen fluo-
ride or fluorine gases. The hydrogen fluoride treatment was used for re-
moving oxide contamination from the salt as H,0 vapor; the fluorine treat-

ment, for removing the uranium as gaseous UF6.

3. CHRONOLOGY

Design of the MSRE was started in the summer of 1960. Construction
of the primary system components and modifications of an existing building
to house the MSRE began in 1962 and by mid-1964, installation of the
salt systems was completed.8

Early in the design and development of the MSRE, some of the engineers
involved were assigned to become members of the operating staff. As design
and construction progressed, these engineers made periodic reviews and in-
spections to assure that proper consideration was being given to opera-
bility. They also began writing procedures for checkout and operation of
the systems and components. The engineer who was to head the operations

staff was given the responsibility for detailed planning of the training
program, and he and the other engineers prepared the instructional material.
As indicated on Fig. 3, the initial operating staff was assembled by July,
1964 and basic operator training was started on July 6.

As construction was completed on more parts of the plant, the effort
spent on checking, calibrating, and testing increased. In late September,
operations were placed on a 24-hour, 7-day basis with four rotating-shift
crews. Flush salt was added to the fuel flush tank, coolant salt was added
to the coolant drain tank, and auxiliary equipment was put into operation.
The fuel and coolant circulating systems were filled with salt and the
plant was operated during most of January and February, 1965. The loops
were then drained to hydrofluorinate the flush salt, to load fuel carrier
salt, and to complete other preparations for zero-power nuclear operation.
During this precritical shutdown, several weeks were spent in advanced
training of the reactor operators, with emphasis on nuclear aspects, and
the administration of comprehensive certifying examinations by Molten-Salt
Reactor Program supervision.

Criticality was attained on June 1, 1965 and the zero-power nuclear
experiments were completed during that month. The reactor was then drained
for completion of shielding and containment provisions and minor modifi-
cations. During this shutdown, the operators underwent additional training
and testing with emphasis on power operation. Nuclear operation was re-
sumed in December, 1965. After some difficulties with the fuel off~gas
system were overcome, full power was attained in May, 1966. By December
of that year, the planned program of startup testing had been completed
and solutions had been found to the various problems that had arisen. As
indicated in Fig. 3, the reactor was at full power a large percentage of
the time for the remainder of the operation with 235U, which ended in
March, 1968. During the 5-month shutdown in 1968, core specimens were
changed, various maintenance was done, the flush and fuel salts were pro-
cessed to remove the uranium, and a charge of 233y fuel was added to the
fuel carrier salt. During these months, time was available for guided
self-study by the operators, and before nuclear operation was resumed, they
were reexamined.

Criticality was attained on October 2, 1968 and on October 8, USAEC

Chairman Seaborg took the reactor power to 100 kW, marking the first time
SALT IN LOOPS SALT IN LOOPS

SALT IN LOOPS

FIRST EXAM PRECRITICAL EXAM 235U CRITICALITY
PRECRITICAL ¥ PREPOWER  PREPOWER
PREPARE FOR OPERATION AND ASSEMBLE CREW INITIAL TRAINING e TRAINING + TRAINING EXAM
A : CHARG et =
CHARGE  FLUSH C_H“A’R“G“E"‘“
COOLANT S&\LT sézl_‘!_gT CARRIER SALT ADD FUEL TEST CONTAINMENT
—
== MODIFICATIONS AND
PROCES” FLUSHSALT COMPLETION OF CONSTRUCTIQ"
FUEL
FLUSH — — ¥
L cooLan )
J F M A M J J A S 0 N D J F M A M J J A s 0 N D
1964 1965
Low INVESTIGATE GO TO CHIEF
POWER  OFF—GAS FULL MODIFICATIONS AND REMOVE REPAIR OPERATOR
TESTS  PLUGGING POWER MAINTENANCE SALTPLUG MAINTENANCE SAMPLER TRAINING
T B ) R
FEsT ORE = REPAIR CORE ==
CONTAINMENT  SPECIMENS = CONTAINMENT AIR LINES SPECIMENS
== B 3 =3 Q

{

{f

ORNL-DWG 73308t

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

POWER (MW}
N A DD

   

 

 

  

 

   

 

    

 

 

 

 

 

 

 

 

 

 

 

FUEL
FLUSH
COOLANT
1967
233 CRITICALITY
TESTS AND MODIFICATIONS OPERATORS ‘ MAINTENANCE & GANMMA
OFCHEMPROCESQNGPLANT EXAM TENANCE SAMMA
REERERESE, i
CORE HOCESS SALTS e e
SPECIMENS FLUSH FUEL ADD 233U FuEL
— a T g )
8
Z6
g 4
2
S
FUEL o
LUSH
COOLANT ==

 

 

 

 

 

 

 

Fig. 3.

Chronology of Operator Training and Reactor Operation.

 

 

 
that any reactor had been operated with 233y fuel. After completion of
rod-calibration and other tests, the reactor was taken to full power in
January 1969. Investigations of the effects of gas circulating with the
fuel involved revisions to permit operation of the fuel pump at reduced
speed. Power operation continued until the first of June when the loops
were drained to change the core specimen array and to make preparations
for some special expefiments. |

The remainder of the operation involved extensive sampling, gamma
scanning, and varying of operating conditions to study the distribution
of fission products and tritium. Consideration of the limited funds
available to the Molten-Salt Reactor Program, the fact that the primary
goals of the MSRE had been achieved, and the funding needs of other molten-
salt breeder development led to a decision by program management to termi-
nate the MSRE operation. Accordingly the reactor was shut down on Dec. 12,

1969 and within 2 weeks the operating staff was disbanded.

4. OPERATING PERSONNEL

The initial operating staff was comprised of four crews, each con-
sisting of two engineers and three technicians, all under the supervision
of an operations chief and assistant operations chief, and augmented by
two day-shift technicians. Long-range planning, analysis of data, mainte-
nance coordination, and design changes were handled mostly by other groups
within the MSRE QOperations Department, whose original organization is out-
lined in Fig. 4. In addition to those shown, personnel of other ORNL di-
visions such as chemical analysis groups, health physics, industrial hygiene,
metallurgy, chemistry, instrumentation, computer programming, and maintenance
forces contributed significantly to the MSRE.

After the completion of the nuclear startup tests, the shift crews
were each reduced to one engineer and three technicians. Later, after the
operation became more routine, each crew was further reduced to 3 tech=-
nicians. On day shifts or when special experiments were being run this

3-man crew was augmented by an engineer and 1 or 2 technicians.
 

 

 

OPERATIONS DEPT, HEAD

 

 

 

 

 

COORDINATOR J_-—FDESIGN LIAISON ENGR. ]

 

ANALYSIS CHIEF
3 ENGINEERS
1 TECHNICIAN

 

MAINT, COORDINATOR

 

 

 

CHEMIST

2 TECHNICIANS

 

 

 

 

 

 

METALLURGIST

CHEM. PROCESSING ENGR.

 

 

 

 

 

 

 

Fig. 4. Original MSRE Operations Organization Chart.

 

ORNL—DWG 67 - 6806

 

 

OPERATIONS CHIEF
ASST. OPER. CHF,
2 TECHNICIANS

 

 

 

SHIFT SUPERVISOR
ASST. SHIFT SUR
3 TECHNICIANS

 

 

 

SHIFT SUPERVISOR
ASST. SHIFT SUP.
3 TECHNICIANS

 

 

 

SHIFT SUPERVISOR
ASST. SHIFT SUP.
3 TECHNICIANS

 

 

 

 

 

 

SHIFT SUPERVISOR
ASST. SHIFT SUPR.
3 TECHNICIANS

 

 
10

Turnover in personnel was moderate; between July 1964 and December 1969
3 engineers and 5 technicians joined the operations group and 10 engineers
and 4 technicians left the group.

The duties of the original operating staff were as follows.

The Operations Chief was responsible for organizing and training the
operations staff., He participated in long-range planning, general policy
decisions, and safety reviews of the reactor. He (or the Assistant Opera-
tions Chief) was responsible for the execution of the daily experimental
program and decided on the course of action to be taken in case difficul-
ties prevented carrying out the planned program.

The Assistant Operations Chief, in addition to substituting for or as~
sisting the Operations Chief in the above, investigated special operational
problems encountered, reported on operational activities at planning meet~-
ings, and directly supervised the activities of the day-shift technicians.
The Operations Chief and Assistant Operations Chief were on call at all
times. There were frequent communications between them and the evening
and midnight shifts and all shifts over the weekends.

The day-shift technicians helped in special experiments, helped take
routine samples, maintained records of chemical analyses, filed other data,
maintained operating supplies including log forms, check lists, etc., and
provided vacation relief for the regular operators.

Each Shift Supervisor was responsible for coordinating and overseeing
all operations and maintenance on his shift., It was his duty to carry out
all shift instructions, to maintain an overview of the plant with special
alertness for anomalous conditions, and to keep all members of his crew in-
formed of changes in equipment, plans, or operating procedures. He was re-
quired to be especially familiar with all equipment and operations so as to
be able to react promptly and effectively to any emergency.

The Assistant Shift Supervisor assisted in the foregoing duties and
handled special tasks such as the numerous tests required during the start-
up phase of the MSRE.

The three technicians on each shift rotated through three different

assignments; building-log, control~room, and utility. The building-log
11

technician was responsible for taking all routine data except the control-
room log. During each 8-hour shift, he made 2 complete rounds of the re~
actor site and was responsible for noting any abnormalities or departures
from the recommended maximum and minimum limits shown on his log sheets.
The control-room technician took the control-room log, calculated and plot-
ted important data, and was responsible for the operation of the reactor
from the control room. He kept a written record in the console log of all
occurrences during his shift. The utility technician was responsible for
sampling and miscellaneous tests and duties as requested. Because the
three assignments were rotated, each operator had a good understanding of
the overall operation and a familiarity with each job.

In the final organization, one of the three technicians on each crew
was designated Chief Operator and assumed most of the duties of the previ-
ous shift supervisor. Candidates for Chief Operator were given additional
training and examination before being given responsibility for the shift.
Six technicians were certified as Chief Operators so that vacation relief
could be handled. An engineer was assigned as shift supervisor on the day
shift to handle the extra work assoclated with coordinating special tests
and scheduled maintenance.

The educational background and previous experience of the original
staff plus those who were later added are summarized in Fig. 5.

All of the supervisors had BS degrees or equivalent. Two were reactor
school graduates. Five had majored in chemical engineering, 2 in nuclear
engineering, 2 in mechanical engineering, and one each in electrical engi-
neering, marine engineering, electronic engineering, and chemistry. Two
of the shift supervisors came to MSRE as new hires, having just received
BS degrees in nuclear engineering. The others had a minimum of 7 years of
prior experience in design and/or development. The average period of prior
employment by Union Carbide was 13 years (including the new hires).

Most of the technicians had at least 2 years of college (average 2.6)
and a minimum of 4 years design or operating experience (average, 12) be-
fore coming to the MSRE. The average period of prior employment by Union

Carbide was 12 years.
ORNL-DWG 733087
ENGINEERS TECHNICIANS

1 2 3 4 5 5 7 8 9 10 M 12 13 1 2 3 4 5 6 7 8 a 10 11 12 13 14 15 16 17 18

7L COLLEGE ETC.
REACTOR SCHOOL

DESIGN AND CONSTRUCTION
(OTHER THAN MOLTEN SALT)

MOLTEN SALT DESIGN AND
CONSTRUCTION

DEVELOPMENT AND OPERATION
(OTHER THAN MOLTEN SALT)

MOLTEN SALT DEVELOPMENT
OPERATION OF REACTOR LOOPS
OPERATION OF OTHER REACTORS

PREPARATION FOR MSRE TRAINING
OR OQPERATION

   
   
   
   
   
   
   
   
   
   
   

]

 

 

 

 

 

21
20
19
18
17
16
15
14
13

1

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1

YEARS EDUCATION OR EXPER{ENCE

O = 0w b o N

   
    
  
   
 
      

     

  

 

 

  
 
 
      

 

 

 

  

 

 

 

 

 

 

 

 

 

  

 

      
 

     
  
   

 

 
  
  

  

 

  
  

  
 

 
 
    

 

 

 

 

   

  

 

 

 

 

 

 

 

o ’ :‘..‘
7 o
7 ol
o Py
Vo oy
v e
s T o
S g e
I o 0l
Ll
};;; s
s
S
S
s
e
prs
Ay
. [ //
-~ -~
~ s, z e - o
Pyl i e s~ s crod oo L
o Sk sqs e -1
s P ~ // /f e o Ak // 3
o Pard P A vl e /,v/’/;/ A A -

 

 

 

  

 

 

 

 

 

 

 

 

 

Fig. 5. Educational Background and Previous Experience of MSRE Operators.
(as of date of hire by MSRE Operations Department)
13

5. TRAINING

Although most of the MSRE operating staff had had considerable previ-
ous experience and several had an intimate familiarity with some parts of
the plant, all were required to participate in a training program leading
to examination and certification. Because of the unique nature of the
MSRE, the content of the training program and examinations was determined
by Molten-Salt Reactor Program personnel.*

The goals of the training program were to instill in each operator
and supervisor the proper attitudes, a good basic knowledge of the entire
plant, an adequate understanding of every phase of the operation, and a
familiarity with the procedures that he would be following. The approach
used at the MSRE, chosen on the basis of past experience, was a combina-
tion of formal training, self-study, and on-the-job training. Maximum ad-
vantage was taken of the special knowledge of members of the staff, by us-
ing them as instructors for the subjects in which they were experts. Other
instruction was given by members of ORNL's Health, Health Physics, Instru-
mentation and Control, Reactor Chemistry, and Chemical Technology Divisions.

The initial period of formal training began when the staff was first
assembled in July 1964. The second formal training period (called pre-
critical training) followed the initial flush-salt operation. After the
zero—power experiments, the reactor was again shut down and in the fall
of 1965 we had the third formal training period (called prepower training).
In 1967, before technicians were certified as Chief Operators, they were
given special training. Prior to the beginning of operation with 233U fuel
in 1968, extra time was allocated to all operators for self-study but there
were no formal training sessions.

As indicated in Table 1, a total of approximately 350 hours was spent

in formal training on various topics. In addition, considerable time was

 

*Although the details of training and examinations were not subject to
outside approval, the entire operation of the MSRE, including operator
training, procedures, etc., was reviewed periodically by the ORNL Reactor
Operations Review Committee. There was also an annual safety review by an
AEC~0ORO committee and the Operating Safety Limits required AEC approval,
Table 1.

Hours Spent in Formal Training Sessions

 

 

 

 

 

Initial Training Pre~Critical Training Pre-Power Training
Ass't
Subject Shift Shift Shift Supr. &

Supr. Supr. Tech. Ass't Shift Supr. Tech. All

Design and Operation 99 107 107 44 36 30
Instrumentation 33 33 33 18 12 5
Reactor Physics 10 10 10 9 0 0
Reactor Chemistry and Metallurgy 4 4 1 0 0
Health Physics and Ind. Hygiene 2 2 2 7 7 0
Other Reactors 24 8 8 0
On-Line Computer 8 2 3
Flow Sheet Checkout 61 71 71 0 0 0
Total 233 227 227 95 65 38

 

w1
15

allocated for self-study. Details of each training period and the train-

ing of replacements are described in the remainder of this chapter.

Initial Training

 

The initial training period lasted several months. Most of the time
during the first few weeks was spent in classroom sessions with some self-
study periods and tours. Later, less time was allocated to classroom in-
struction and more time was spent in checking out flowsheets, labeling
equipment, lines, valves, and switches, shakedown of equipment, and self
study.

Assignments had been made at least a month in advance and time had
been allocated for the instructors to prepare for the training sessions.
As this was the first step in the development of the operating crew, the
teachers were instructed to cover all information which might be needed
by all operating personnel, but to present it so that it could be under-
stood by newly-hired operators who knew nothing about the reactor. All
sessions were held in a conference room at the reactor site where a black-
board, slide projector, and opaque projector were available. Future oper-
ators and supervisors attended all sessions; other project personnel at-
tended those sessions pertaining to their future responsibilities at the
MSRE.

The schedule followed for the first 4 weeks and the instructors for
the lecture periods are given in Appendix A. The first session was an in-
formation and orientation meeting cofiering such items as organization
charts, schedules, and plans. This was followed by a general description
of the reactor emphasizing the functional interrelationships, and omitting
nonessential detail. Drawings of the building and general equipment lay-
out were then discussed, followed by guided tours. Instruction then turned
to the design of the various systems and equipment, interspersed with ses-
sions on the basic subjects of molten-salt chemistry, reactor physics,
health physics, and industrial hygiene. Beginning on July 15, emphasis
was switched from details of design to methods of operation. The startup,

normal operation, shutdown, and unusual operation of each auxiliary system
16

was covered. This was followed by instruction on the actual startup of
the reactor, including heatup, addition of fuel salt, filling the reactor,
and going to power.

During the period from July 21 to July 28, classroom training was held
only in the mornings with self-study and equipment checkout in the after-
noons. In the morning sessions, training on control circuits and instru-
mentation was started and subjects not completed in the first 2 weeks were
covered. In the afternoons, the staff was divided into four groups and
each group was assigned certain systems to check out flowsheets. Phase 1
of the flowsheet checkout involved a thorough check to assure that the
flowsheets agreed exactly with what had been installed. When a group had
finished checkout of a system, they were considered experts and in Phase 2
provided detailed guided tours of their systems to all other operators,
(Instructions given to the operators are reproduced in Appendix B.)

During the following 3 days, the assistant operations chief and the
4 shift supervisors spent full time at the Oak Ridge Research Repctor ob-
serving a startup. The remaining crew had classroom discussion, equipment
study in the field, and visited moltenw=salt test loops.

On August 3, 1964, all trainees were given their first test, containing
153 questions with 2-1/2 hours allocated to complete it., This first test,
made up mainly from questions noted during the training sessions, was not
intended as a qualifying examination but rather as a guide to each indi-
vidual as to his relative progress. On this, as well as later tests, it
was felt that there should be definite answers to all questions. Therefore,
multiple choice, true/false, matching, and filling blanks were used where
possible. A test of this type minimizes misunderstanding of the questions
and allows coverage of a wide variety of subjects in a reasonable time.

No intentional trick or catch questions were used. Where conflicting in-
formation was given in different sources or where the exact answer was not
important, a range of answers was accepted. In the assignment of credit
for each question, an attempt was made to weigh the difficulty and/or im-
portance of the question. The time allocated was considered to be adequate
for most people to finish. The distribution of grades on the test for the

23 engineers and technicians who took it are shown in Fig. 6.
TEST SCORE (%)

100

80

60

40

20

17

ORNL-DWG 73—-3088

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

O TECHNICIANS
® ENGINEERS

 

 

 

 

 

 

 

 

 

 

 

 

1

2 3 4 5 6 7 8 9 10 1112131415 16 17 18 19 20 21 22 23

Fig. 6.

PERSON NUMBER

Initial Training Test Grades

 
18

The papers were graded promptly and were returned for self-evaluation
by each individual and further study. Although this was the primary pur-
pose, comparisons were made of average grades in various subject categories
in an attempt to determine which areas needed the most additional training.
The results, listed below, showed that the trainees had learned the physi-
cal layout rather well, and showed little difference in proficiency among

the other subjects.

 

Subject Average Grade, 7
Design 56
Physical Layout 79
Line and Valve Numbers 35
Operation (General) 44
Chemistry 55
Nuclear 44
Instrumentation 47
Engineering Conversions 55
General 48

From August 4 to August 18, 1964, 2 to 4 hours each day were spent
on training sessions emphasizing control circuits. The rest of the time
was used for self-study and operational duties. Self-study consisted of
studying available documentation as well as tracing out lines, studying
equipment in the field, and becoming familiar with the area. Operating
duties included such items as checkout, calibrating and operating equip-
ment, making up log forms, procuring supplies, and labeling panel boards.

All operating personnel attended all sessions of a special MSRP In-
formation Meeting on August 18 and 19, 1964. The meeting recognized the
completion of MSRE construction as an important milepost and the papers
presented there gave a comprehensive picture of the development of molten-
salt reactor technology up that point. (The papers were issued as an ORNL

report, reference 9.)
19

The period between August 20 and late September 1964, when rotating-
shift work started, was spent largely on operating duties. Formal train-
ing consisted of some additional control circuit instruction and a thorough
review of the fuel, coolant, and all auxiliary systems. Each system was
assigned to an operations engineer and sufficient classroom time was allo-
cated to cover all information pertinent to operation of the reactor. Top-
ics covered included normal startup, normal operation, unusual or special
operations, normal shutdown, operating limits, associated instrumentation
and operation prior to criticality. A short quiz was given after each ses-
sion to determine progress.

During the prenuclear operation and testing, nearly all of the reactor
equipment was operated except for the heat-removal equipment, containment,
and control rods. This valuable on-the-job training was supplemented by
informal training by shift supervisors and day persomnel. To stimulate
self~study, each person was asked to submit a list of questions and an-
swers which he felt were important. These were reviewed to eliminate dupli-
cation, other questions were added to fill in gaps and the final compila-
tion of 569 questions and answers plus a separate group of 42 questions and

answers on nuclear characteristics were issued for study.

Precritical Training and Certification

In March 1965, when the reactor was shut down for completion of prep-
arations for zero-power nuclear tests, the operating personnel were taken
off shift for intensive training, examination, and certification. All re-
ceived two weeks of formal training in which radiation safety and nuclear
instrumentation and control were stressed. The engineers were given an
additional week of formal training, with emphasis on nuclear aspects of
the MSRE and discussions of the zero-power physics experiments. The sched-
ule for this session is given in Appendix C.

Two steps were taken to develop some familiarity with and proficiency
in nuclear testing and operation. First, each person participated in a
criticality experiment at the ORNL Pool Critical Assembly (PCA). Second,

a realistic simulation of the MSRE was set up and was used extensively.
20

The MSRE simulation made use of a TR-10 analog computer set up in the
MSRE control room, with input signals from the control-rod position indi-
cators. Outputs representing log count rates, log power, period, and lin-
ear power were sent through the actual instrument and control channels,
including displays in the reactor control room.l® Thus it was possible
for each operator to practice taking the reactor critical, using the same
procedures and controls and reading the same instruments that would be used
later in the actual experiments. To further familiarize personnel with the
controls and response of the MSRE each crew was given a set of "experiments"
or problems relating to predicted criticality, rod worth, etc., that they
were required to carry out on the simulator.

On April 13, 1965, all operators were given a qualifying examination
containing 300 questions and lasting 7 hours. An additional test, con-
taining 57 questions and lasting 1 hour, was given to the shift super-
visors and assistant shift supervisors. An outline of the tests and sample
questions are given in Appendix D. Before the tests were graded, passing
grades for each subject were established by the MSRE Department Head and
the MSRE Operations Chief. A breakdown of the various categories and the
number of persons failing to meet the standard is given in Table 2. The
overall tests grades are plotted in Figs. 7 and 8 along with a weighted
average passing grade,

After the tests had been graded, each person was examined orally by
the department head and the operations chief to verify that his test scores
accurately reflected his understanding of each subject. In addition, vari-
ous situations were posed and questions asked to determine how the operator
would react under unusual circumstances. During the oral examinations,
subjects apparently needing additional study were noted and were pointed
out to the individual so that self-improvement could be made. If the exam-
iners judged the person to be qualified, he was formally certified. 1If
not, he was given further time for self-study and informal instruction.

He was then reexamined and in most cases showed sufficient improvement to
be certified.

Before startup at least one shift engineer and 2 technicians had been
certified on each shift. Uncertified persons were not permitted to operate
any of the controls relating to nuclear operation except in the presence

and under the supervision of a certified operator.
21

Table 2. Evaluation of Precritical Test

 

 

 

 

 

Time allowed Passing grade Number
Subject for test % not passing
(minutes) Engr. Tech. Engr. Tech.

Instrumentation 100 70 60 7 14
Flow Sheets and Operating

Parameters 100 75 70 0 7
Design l 50
Health Physics and

Industrial Hygiene J 75 75 4 14
Miscellaneous 70 60 0 1
Layout : 75 75 3 2
Calculations 60 80 65 3 9
Electrical ’ 75 65 4 0
Emergencies 80 70 2 1
Processing or Sampling 75 70 4 7
Physics and Nuclear ’ 25 75 60 2 8

ENGINEER'S EXAM
General ) 75 ~— 7 —_—
Computer 65 — 2 -
Abnormal Operating 60

Procedures ' 75 - 6 -
Physics and Nuclear 75 - 8 -

 

213 engineers and 18 technicians took the test. Of these, 1 engineer
and 5 technicians had not been in the initial training program.
TEST SCORE (%)

100

80

60

40

20

ORNL-DWG 73-3089

 

I

 

 

— —+ -t 00— % 4 | -

 

 

 

 

 

PASSING GRADE (ENGINEERS)

 

o—o—0—2-

O

O
O
O

 

PASS

o ®

e O0O© bl
|
|

—

T
ING GRADE {TECHNICIANS)

 

—P — — —

 

 

 

 

 

O
®

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

TECHNICIANS
ENGINEERS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

T 2 3 4 6 6 7 8 9 10111213 141516 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31

Fig. 7.

PERSON NUMBER

Precritical Training Test Grades

cZ
100

80

60

TEST SCORE (%)

40

20

23

ORNL—-DWG 73—-3090

 

 

 

 

. PASSING GRADE

 

 

 

 

 

® ENGINEERS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1 2 3

Fig. 8.

4 5 6 7 8 9 1011 1213
PERSON NUMBER

Precritical Training Test Grades (Engineer's Exam)

 
24

Criticality was attained on June 1, 1965. During the zero-power runs
very little time was available for any training other than that gained from

carrying out the detailed experimental procedures.lls12

Prepower Training

In September and October 1965, during the shutdown prior to approach
to power, each operator was given a week of formal instruction. (Since it
was not possible to leave the reactor unattended, two training sessions
were provided, with half of the operators in each.) The subjects covered
and time spent on each are given in Appendix E. The session included some
review coupled with detailed instruction on the operation of equipment and
systems not in use during the previous operational period.

Provisions were also made to allow each operator to practice simulated
power operation, using two TR-10 analog computers connected to the reactor
controls and instrumentation.!® Inputs to the computers were signals indi-
cating the actual positions of the rods and radiator doors and the actual
pressure drop of the cooling air across the radiator. The outputs indi-
cated on the reactor instrumentation were linear power, log power, log
count rate, period, fission chamber position, reactor inlet temperature,
reactor outlet temperature, radiator salt outlet temperature, radiator salt
AT, and radiator heat power (flow times AT). The operators practiced
taking the reactor critical and raising and lowering the power using manual
or automatic load control. Simulator problems were solved such as deter-
mining the power level with various load system configurations, determining
the heat capacity of the loop, and determining the temperature and power
coefficients of reactivity.

Prior to startup for power operation, a 2-hour written examination
was given to all operators and an additional one-hour examination was given
to the engineers. The distributions of grades for these examinations are
given in Figs. 9 and 10. Sample portions of the exafiinations are given in
Appendix F, As was done on the pre-critical certifying tests, each person
was examined orally before he was certified for operating the reactor at

power,
TEST SCORE (%)

100

80

60

40

20

ORNL—-DWG 73—3091

 

l

 

 

o ©O

 

 

PASSING GRADE (EN

GINEERS)

 

 

 

O
ce© 1 1 | L1
PASSING GRADE (TECHNICIANS)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

® ENGINEERS

 

 

 

 

 

 

 

 

 

 

O TECHNICIANS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1

2 3 45 6 7 8 9 101112131415 16 17 18 19 20 21 22 23 2425 26 27 28 29 30

Fig. 9.

PERSON NUMBER

Prepower Training Test Grades

4
TEST SCORE (%)

100

80

60

40

20

26

ORNL—-DWG 73—-3092

 

 

 

PASSING GRADE

 

 

 

 

 

 

 

 

 

 

 

 

® ENGINEERS

 

 

 

 

 

 

 

 

 

 

 

 

 

1

 

2 3 4 5 6 7 8 91011 12 13

PERSON NUMBER

Fig. 10. Prepower Training Test Grades (Engineer's Exam)

 
27

Training of Chief Operators

The reduction of the operating crews to three men each had been care-
fully considered from the standpoints of safety and effectiveness and was
judged to be feasible. (By mid-1967, equipment difficulties encountered
during early operation had been overcome and the test program was well into
its sustained high-power operation phase, so that unusual demands on the
operating crews on other than the day shift had been minimized.) It was
recognized, however, that the technicians who were to take over the super-
vision of the crews would need some special training for their new duties.
Six candidates for Chief Operator were chosen on the basis of their past
experience and test grades, and the opinions of the shift supervisors, as-
sistant operations chief, operations chief, and operations department head.
These technicians were given the additional instruction outlined in Appen-
dix G. Training was on an individual basis and to some extent was tailored
to the individual's need. Emphasis was placed on duties previously handled
by the shift supervisors, such as trouble-shooting, tracing out control
circuits, etc. After instruction, each candidate was given charge of a
shift for one or two weeks with an original shift supervisor on duty but
acting only as adviser. They were then given a 2-1/2-hour closed-book
written examination and a 1-1/2-hour open-book written examination. Sample
portions of these examinations are given in Appendix H. All 6 candidates
passed and after oral examinations, all were certified as qualified Chief
Operators.

It appeared that considerable benefit was gained from this training.
Therefore, it was subsequently given to ali the other operators. No exami-

nation was given to the others, however.

Training Before 233y Startup

The 5-month shutdown that followed the end of extended operation with
2357 fuel and preceded the 233y startup experiments was recognized as an
appropriate time for refresher training of the operators. Instead of train-
ing sessions attended by the entire staff, however, a program of self-study

and small-group study was initiated. The schedule was rather flexible to
28

permit other jobs to be carried on, but a date was announced on which all
members of the operating staff were to be reexamined. The nuclear aspects
section of the Operating Procedures!! was revised to show the calculated
reactivity coefficients and other characteristics with 233y fuel and a
memorandum summarizing the changes was issued. As an aid in review of all
systems, a 74-page programmed instruction manual was prepared. Prior to
startup with 233y, a comprehensive review test was given to all operators,
The distribution of grades on this test is shown in Fig. 1l1. The results
of the tests were discussed with each operator and each was required to do

more studying of subjects in which they appeared weak.

Other Training

From the standpoint of keeping in practice, it would have been bene-
ficial if each operator could have filled and started up the reactor, shut
down and drained the reactor, and done all other operations periodically.
However, with the MSRE, which was on line most of the time, this was not
possible. The various job categories were rotated among the operators to
help distribute the experience and, whenever there was opportunity, each
operator was required to take the reactor critical and subcritical and to
raise and lower the power as well as fiost of the other routine operations.
Occasionally in slack periods, problems were assigned for each crew or each
individual to solve. Typically these involved radiation and health physics,
reactivity changes and other topics that were important but were seldom en-
countered in routine operation. Information regarding changes in plans or
equipment were usually communicated to operating personnel by means of
Shift Instructions and changes in the Operating Procedures. (See Chapter
8.) Each Shift Supervisor or Chief Operator was responsible for assuring
that all persons on his shift were familiar with all changes and that they

remained adequately trained to run the reactor.
TEST SCORE (%)

100

80

60

40

20

29

ORNL-DWG 733093

 

 

 

 

 

® ®
o @
0
oo @ PASSING GRADE (ENGINEERS)
o O PASSING GRADE (CHIEF OPERATORS)
= b - - - - -
o ©
o

 

 

 

 

 

 

 

 

 

 

 

© TECHNICIANS
® ENGINEERS

anil

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1

2 3 45 6 7 8 9 1011121314 1516 17 18 19 2021 22 23
PERSON NUMBER

Fig. 11. Pre~233 Training Test Grades
30

Training of Replacements

During the five years of reactor operation, 3 engineers and 5 tech-
nicians were assigned to the MSRE for training and use as operators. Their
training was on a person-to-person basis, with instruction by members of
the MSRE staff. The program outlines which were followed for group train-
ing were used as guides. Considerable time was provided for self-study of
the various sections of the MSRE Design and Operations Report, semiannual
reports, flow sheets, control circuits, etc. When training was complete,
each of these operators was given the same series of tests and oral reviews
that had been used with the main group. The elapsed time between arrival
at the MSRE and final testing varied from 3 to 8 months depending upon need
for replacemént, availability of instructors, background of the individual,

and the extent of his other duties.

Equivalence to AEC Examination

In their annual safety review of the MSRE in September 1968, the USAEC-
ORO committee suggested that the MSRE operator examinations be reviewed to
assure equivalence with the requirements for non-AEC reactor operators (ref.
13). At their suggestion, an ORNL employee who administers tests for the
AEC's Division of Reactor Licensing reviewed one of the three major tests
that had been used at the MSRE. It was his opinion, based on this one test,
that the MSRE operator testing was adequate and equivalenfi. However, he
suggested that if other operators were to be certified, new tests should be
prepared conforming more directly to the AEC guidelines. A test was made
up using 10-~CFR~5>5, the AEC Licensing Guide,!l" and suggestions from the
examiner. This test, designated "™SRE Operator's Examination — May 1969,"
is given in Appendix I. One engineer, who was in training at the time, was
given the MSRE Precritical, Prepower, and Pre-233y examinations, and then
took the May 1969 examination and an oral operational test before certifi-
cation. Twelve hours were allocated for taking the operator's sections
(A-~G) of the May 1969 examination and 9-1/2 hours for taking the shift
supervisor's section (H-L). He passed all tests, but certification was
based, as with past operators, on the more inclusive MSRE tests, informal

review, and oral questioning.
31

6. OPERATION OF THE REACTOR

Although the goal of the training program at the MSRE was to make all
operators capable of operating the reactor without instructions or proce-
dures if need be, the practice was to follow detailed written instructions
to the maximum practicable extent. The intelligence, interest, and capa-
bility of the operators was recognized, however, and in all instructions
an attempt was made to give the reason for as well as the details of the
required actions. In return, it was each operator's duty to provide feed-
back. For example, he was expected to provide an explanation, if known,
for all log notations which were not self-explanatory and for any devia-
tions from normal or expected events. Suggestions, including unproved but

credible theories to explain observed events, were encouraged.

Planning

Operation of the MSRE followed a published test program,l®,16 and the
objectives of each run were worked out in advance among the engineers, chem-
ists, physicists and directors of the Molten~Salt Reactor Program. Most of
the day-to-day, detailed planning was done in planning meetings held at the
reactor site. Originally held each day, these meetings later came to be
held at intervals of 1 to 3 times per week, depending upon the need. All
technical personnel on duty at the reactor plus representatives of sup-
porting groups attended these meetings. (The MSR Program Director was usu-
ally present also.) A member of the operations group (usually the Assist-
ant Operations Chief) began the meeting by reporting on the previous peri-
od's activities. Others added to this any other pertinent events, analyses
made, maintenance status, etc. Tentative plans were then presented and
discussed. Often a smaller group remained to work out details or to dis~
cuss problem areas. In cases where there was uncertainty or conflicts (be-
tween experimenters, or between maintenance and operation, for example) the
MSRE Department Head usually made the final decision. The meetings started
at 11:00 a.m., which allowed ample time before to digest most of the previ-

ous period's work and sufficient time afterward to prepare detailed shift
32

instructions, obtain materials needed and make the necessary plans to carry

out the program for the following period before the end of the day shift.

Instructions to and Communications with Operators

Shift Instructions were prepared each weekday by a member of the ope-
rating group (usually the Assistant Operations Chief or the Operations
Chief). These were based mainly on decisions reached in the planning meet-
ings. A general information section outlined the plans and provided other
information of interest to the shift personnel. The instructions sometimes
were very specific, such as a detailed check list; at other times they re-
ferred to a separate procedure such as a section of the Operating Proce-
dures!! or a test memo. If the instructions were to be followed for longer
periods, instead of being repeated each day in subsequent shift instructions,
they were usually put in the Permanent Shift Instructions. The Permanent
Shift Instructions were reviewed periodically and appropriate sections were
incorporated into the Operating Procedures, The shift instructions could
override the operating procedures or test memos. If difficulties prevented
following the prescribed program, the Chief Operator, Shift Supervisor, As-
sistant Operations Chief, Operations Chief, and/or Operations Department
Head decided what alternatives to follow. Typical examples of Shift In-
structions, Operating Procedures, and Test Memos are given in Appendix J,

A master up~to-date copy of the Operating Procedures was maintained
in the operations office. To avoid confusion, changes in this master copy
were entered only by the Operations Chief or Assistant Operations Chief.

All personnel were required to be familiar with these changes and records
were maintained indicating this. The master copy or a copy corrected to
agree with the master was used for all operations. Control-room log and
building log forms were a part of the Operating Procedures. Current limits
were indicated on these for each variable as well as the frequency for re-
cording variables. The first few pages of the control-room and building
log forms are shown in Appendix K. Before starting a log (or check list)
the operator was required to correct his copy of the form to agree with

the master. If variables were found out of limits, they were corrected

or reported to the Shift Supervisor or Chief Operator. He decided on the
33

course of action to be taken either independently or after contacting ap-
propriate day personnel. During transient conditions, such as at shutdown
or startup, when the appropriate log limits were changing rapidly, shift
personnel were expected to decide on what limits were proper and what equip-
ment should be operating at the time.

Test Memos were detailed procedures which were written for more in-
volved speclal tests. These required the approval of the MSRE Operations
Department Head, the Operations Chief, and others as designated by the Ope-
rations Department Head.

There was, of course, an oral exchange of information and instructions
at shift change between the crew leaving and the one taking over.* Every-
one was required to read the shift instructions and console log as soon as

possible after reporting for duty.

Data Taking and Record Keeping

One of the primary records of the MSRE operation was the Console Log,
which was essentially a chronological narrative of events. The time was
recorded at the start of each entry and the initials of the person making
the entry were put at the end. A status and summary entry was made at the
end of each shift. Practically all operations events were noted, but only
the most important maintenance items. (Other maintenance was reported on
the completed work requests or by memos.) Copies were made of the console
log sheets and distributed daily to interested personnel. The originals
(in bound books) were kept in the operations files at the site.

As mentioned previously, almost all variables were recorded on the
building log or controel room log forms. These were arranged by areas to

facilitate recording of data. Completed copies of these were filed in the

 

*
Crew members were required to be at the control room, ready for duty,
6 minutes before the nominal shift change. (They received 0.1 hour overtime
pay for this.)
34

operations files along with completed check lists, operating procedures,
and test memos. Recorder charts were marked at least once each shift and
were removed and filed daily.

The bulk of the data were logged by an on~line digital computer
(Bunker-Ramo Model 340),l7 which monitored all of the important variables
and recorded these on magnetic tape every 5 minutes. In addition, it typed
out an hourly log, an 8-hour log, and a daily report. 1In case a variable
went out of limits, an annunciation occurred and the value of the variable
was typed out by a typewriter located in the main control room. When any
of certain important variables went out of limits, it initiated a fast scan
which increased the frequency of recording of a selected group of variables
to 4 times per second. A fast scan could also be initiated by the operator.
Variables could be trend-logged or plotted if desired. Data typed out by
the computer was removed and filed each day. Magnetic tapes were removed
daily and then transferred to record tapes which held several days' data.

These were filed at the reactor site.

7. APPROVAL OF MAINTENANCE AND DESIGN CHANGES

The forms used in connection with maintenance and design changes are
included in Appendix L. Anyone knowing the need of maintenance could initi-
ate action by filling out a Punch List form. If the job was small, the
work was completed without any additional paper work. No formal record
was kept of work done in this manner. If the job was more involved, a Work
Request was issued. This had to be approved by the Operations Chief or
Assistant Operations Chief. 1In either case the Shift Supervisor's approval
was necessary before work was started. He was responsible for making neces-
sary changes in the system and tagging out necessary valves or switches to
assure safety of the reactor and personnel. After maintenance waswcomplete,
he checked out the job before removing the tags and putting the equipment
back in service. The completed work requests were retained as a record of
what was done.

Proposed modifications which could produce significantly different

characteristics or functions in any component or system (such as piping,
35

instrumentation, switch settings, etc.) were initiated using a "Change
Request' form. The MSRE Operations Department head reviewed all change

requests and determined what other reviews were necessary.
8. INFORMATION AVAILABLE TO OPERATORS

Documents are listed in this section which were useful in the training
of operators or were used by them in the operation of the MSRE. A brief
description of their content and how they were used is given where appli-
cable. The approximate date of issuance is also given when this is thought
to have been pertinent to the training program or operation of the reactor.

-

Design and Operations Report

The need for documentation of reactor design and preparations for ope-
rations, was recognized very early by the MSR Program Director, who issued
a memorandum in July, 1962 assigning responsibilities for writing parts of
a Design and Operations Report. Table 3 lists the subjects and the dates
when the wvarious parts were eventually published.

Part T -'Design” described the mechanical and electrical portions of
the reactor. Flowsheets, electrical one-line drawings and equipment details
were given. Bases for design or design calculations were not included. A
rough draft of Part I was available for the initial operator training and
proved to be very helpful. The report was issed at the time of the zero-
power experiments and although it was not kept up to date, it was neverthe-
less referred to frequently during subsequent operation.

Parts IT-A and II-B — Nuclear and Process Instrumentation!®;1° de-
scribe the instrumentation layout, individual instruments or systems, con-
trol circuitry and other details. Unfortunately, these were not available
for use in training and it was necessary for the instructors to search
through drawings and unpublished information to prepare for the training
sessions. The students had to take very good notes and had to search for
information if needed later. Section II-A was used for reference after its

publication near the end of 2335y operation. Section II-B was not published

until after the conclusion of MSRE operation.
36

Table 3. MSRE Design and Operations Reports

 

 

Part Title ORNL-TM  ‘ublication
Date
I Description of Reactor Design 728 6/65
IIA Nuclear and Process Instrumentation 729 2/68
IIB Nuclear and Process Instrumentation 729 9/72
III Nuclear Analysis 730 2/64
IV Chemistry and Materials 731 -
v Reactor Safety Analysis 732 8/64
V-A Safety Analysis of Operation with 233y 2111 2/68
VI Operating Safety Lifiits 733 " 4/65%
VII Fuel Handling and Processing Plant 907 5/65b
VIII Operating Procedures 980 12/65°
IX Safety Procedures and Emergency Plans 909 6/65
X Maintenance Equipment and Procedures 910 6/68
X1 Test Program 911 12/66
XI-A Test Program for 233U Operation 2304 9/68
XII Lists - -

 

%Three revisions of ORNL-TM-733 were issued in 8/65, 9/66, and 7/69.
bA substantially revised version of ORNL-TM-907 was issued in 12/67.

®Loose-leaf copies were issued to members of the operating staff in
12/65; revised pages were issued as necessary at later dates.
37

Part IITI — Nuclear Analysi520 described the predicted nuclear charac-
teristics, rod worth, coefficients of reactivity, poisoning effects of fis=-
sion products, kinetics of operation and adequacy of biological shielding.
It was the first part published and was available for all of the training
periods. Although it was not kept up to date, it was used occasionally for
reference during operation.

Part IV — Chemistry and Materials was expected to include in a form
convenient for the operators: (a) a summary of salt chemistry (phase dia-
grams, effects of moisture, oxygen, etc.), (b) a description of the chemis-
try of water systems (acceptable limits, suggested analysis, and effect of
activation), (c¢) a listing of the properties of lubricating oil for the cir-
culating pumps (analysis required, acceptable limits, etc.), (d) informa-
tion on the metallurgy of INOR-8 and stainless steel, (e) a discussion of
graphite, (f) explanation for keeping the oxygen and water content of helium
cover gas low, and (g) a discussion of fission products. This part of the
Design and Operations Report was never prepared.

Part V — Safety Analysis?l! described some of the more important sys-
tems, controls, and instrumentation. Plant layout, site features, con-
struction, startup and operating plans were covered. Various possible ac-
cidents were discussed and the consequences delineated. This was available
for most of the training and was extremely valuable. The operations copy
was updated periodically and used regularly during operation. Prior to
operation with 233y fuel, another safety analysis report22 was published,
which provided a basis for operator training and was a useful reference.

Part VI — QOperating Safety Limits2326 tabulated the absolute limits
within which the reactor had tobe operated. This was available before
criticality and was used for instruction. Three revisions were issued.,

Part VII — Fuel Handling and Processing Plant covered all aspects of
the fuel processing plant, (design, hazard analysis, operating procedures,
etc.) and was useful in training operators and for running the chemical
processing plant. The original version?? was issued at the time of the
hydrofluorination of the flush salt in 1965 and a substantially revised

version’ was issued in preparation for the processing in 1968.
Part VIIT — Operating Procedures!! provided most of the information

necessary for operation. A simplified description of basic nuclear facts
38

and nuclear characteristics of the MSRE was given. Methods of startup,
operation, and shutdown of all systems were covered with detailed check
lists. Special and unusual operations were described and attempts were
made to anticipate any troubles which might have developed and suggest cor-
rective action. For instance, each annunciétor was listed. Things which
could cause the annunciation, control actions which would occur and sug-
gested operator actions were tabulated. Data sheets, logs, and check lists
were included and the methods.of taking data and its storage was covered in
detail. The methods used for getting maintenance done safely and for making
modifications to the system or approved documents was described. A rough
draft of this was available for the initial training. This report was is-
sued in a bound edition (two volumes) to the recipients who did not desire
or need to be notified of changes. Others, inciuding all operators, were
issued loose-leaf editions and they were sent copies of revised sections

as they appeared. The master copy was updated as changes occurred.

Part IX — Safety Procedures and Emergency Plans?8 provided procedures
and background information to assist personnel in anticipating, preventing,
and handling emergencies such as fires, release of beryllium, or increases
in radiocactivity, This procedure was availlable before criticality and was
used for training operators. Operators' copies were updated periodically
but a formal revision was not issued.

Part X — Maintenance Equipment and Procedures?? gave general proced-
ures for doing remote maintenance. The reactor operators were not normally
involved in maintenance except to prepare the system and assure that ade-
quate safety precautions were followed.

Part XI — Test Programl5 described the program for the shakedown of
equipment, approach to criticality, and power operation. Various special
tests and experiments were described. Details of experiments were not
given in the report itself, but test memos, numbered to correspond to the
sections in the test report, did give details. These were written and ap-
proved before the experiment or test was started. Operation with 233y
fuel was conceived and planned after the original program was underway and

a separate document!® was issued to outline this phase of the program.
39

Part XII — Lists was to give a ready reference or index to all reports,
drawings, and other information. Such things as a cross reference index to
drawings were to be included. It would have been very valuable during de-

sign, construction, and all stages of operation, but it was never prepared.

Other Reports

Progress reports of the Molten Salt Reactor Program were issued semi-
annually. Informal reports'covering progress in most areas of the program
were also issued monthly. Copies of these, as well as topical reports on
molten—-salt work done at the reactor or elsewhere, were maintained at the
site. Copies of other pertinent reports such as ORNL and Reactor Division
Safety Procedures, Health Physics Procedures and ORNL Standard Practice
Procedures were also available. A Daily Report was issued covering all
known events of importance at the MSRE and a tabulation of important items
such as number of hours critical, total power produced, operating condi-

tions, etc. (See Appendix M.)

Drawings

A stick file was maintained with the latest revisions of flowsheets,
instrument application drawings, block diagrams, control circuit schema-
tics, annunciator schematics, heater schematics, heater and thermocouple
location drawings, electrical on-line drawings, and a few layout drawings.
These were updated whenever a change was made in the system. The trans-
parencies were revised periodically and new revisions put on the stick
file. Microfilm negatives of these drawings and all other MSRE drawings
were maintained at the site. A reader-copier was available for quick view-

ing or obtaining copies of any drawing.
Miscellaneous

(1) Calibration curves and other operational information on equip-
ment or systems was maintained in a loose~leaf notebook in the operations

office. These were updated as new information was accumulated.
40

(2) Copies of manufacturers' bulletins and instructions for all
equipment were maintained in the maintenance office.

(3) Switch tabulations giving the setpoints of all switches were kept
up to date.

(4) Two thermocouple tabulation books were maintained which gave the
readout location of each thermocouple. One was indexed according to thermo-
couple number and the other according to readout instrument number. Ac-
curate records on this were necessary because there were approximately 1000
thermocouples which were routed through a patch panel to the readout devices.

(5) The latest copies of the instrument tabulations, instrument speci-
fications, line schedules, and design data sheets were available.

(6) Up-to-date information regarding the computer including the "‘Com-
puter Manual for MSRE Operators”30 was kept in the computer room located
next to the main control room.,

(7) The main control board was a graphic panel and was very helpful
to the operators. Most of the control circuits were depicted graphically
on a jumper board. Indicator lights provided the operator with information
on each contact in the control circuits. Plug-type jumpers could be used
to bypass some of the interlocks. The jumper board reduced the need for
behind~the-board jumpers. All jumpers at the MSRE were used only under

strict administrative control.

9. RECOMMENDATIONS

The following recommendations are based on the author's experience
gained during operator training, startup and operation of the MSRE. The
intent is not simply to criticize the way things were done at the MSRE but
to try to share recommended techniques with other operators and to point
out possible pitfalls. In general, only those items which directly af-
fected training, preparation for operation, or actual operation are in-
cluded.

A. The MSRE Design and Operations Report was intended to cover all
phases of the reactor and to avoid duplication as much as possible. There-
fore the Operating Procedures, Part VIII, did not cover descriptions of the
systems, equipment, or instrumentation. To learn what was needed for ope-

ration, the operators had to read through much information which was not
41

pertinent to operation. All other parts of the Design and Operations Re-
port should have been abstracted and the information needed for operation
should have been repeated in the Operating Procedures or in a Training
Manual.

B. A special set of drawings needs to be made for use by operators.
Since these would be repeatedly referred to, more than normal effort should
be spent in proper layout, cross-referencing, etc. Most of these need to
be reduced to 1ll- x 17-in. size as were the flow sheets at the MSRE. Some
specific examples are given below.

(1) TFlow sheets should be similar to those issued for the MSRE.
They should include line numbers and names, equipment num-—
bers and names, simplified instrument numbers, locations of
lines and equipment, direction of flow, normal flow rates,
normal temperatures, normal pressures, relative elevations

when important, and normal and fail position of valves.

To avoid unnecessary confusion, most data on components,
cross~references, and approval signatures could be on a
separate sheet rather than on the drawings. Such things
as line sizes, the number of thermocouples or heaters on a
line or equipment, flanges and leak detectors, etc., would

not need to be included.

Considerable effort should be taken to include all items

needed in routine operation but to make these as simple as

possible,

The MSRE flowsheets were made so that lines extending be-
yond the edge of one drawing matched the next drawing. In
other words it was possible to fasten the flowsheets to-
gether and end up with a large drawing of the entire plant.

This is desirable as it facilitates tracing lines.

(2) Although flowsheets, as described above, supplied most of
the day-to-day needs of the operators, there was often a
need to know more detailed flowsheet type information.

Drawings showing this should be made using the simplified
(3)

(4)

(5)

42

flowsheets as a starting point so that the orientation on
the drawings is the same. Details of instrumentation loops,
like given on the instrument application diagrams, should be
added along with any instrument valving and valve numbers.
Control and annunciator circuit numbers should be given from
all sensing elements as well as to all valves and equipment
such as pumps. Line sizes and flanges should be shown along
with the flange leak detector lines. Approximate location
of thermocouples should be indicated. These drawings would
be the starting point for most trouble-shooting.

Four types of control circuit drawings were provided for the
MSRE. Examples of the three used most by operating personnel
are shown in Fig. 12. No one type was entirely satisfactory.
The block diagrams were probably the easiest to use when
learning the circuits but they were not sufficient for ope-
rating the reactor since they did not readily show all the
control actions that occur, the switches that initiate the
action, relay numbers, jumper board lights, etc. The engi-
neering elementaries supplied most of the information neces-
sary but were difficult to use. The jumper board drawings
lacked detail and cross-referencing. These three if com-
bined as shown in Fig. 13 should be much more useful to the
reactor operators. Switch setpoints should not be shown
but should be tabulated similar to the MSRE Switch Tabula-
tion3! with provisions for keeping a master copy up to date.
Simplified electrical distribution drawings were needed
showing all switches, bfeakers, indicators, recorders, etc.
Control power for the breakers should be shown. Control
circuits and switch tabulation similar to that described
above would be helpful.

Combination heater and thermocouple drawings as used at the
MSRE were very helpful. These showed the approximate loca-
tion of the heaters and the thermocouples. Isometric draw-

ings could have been used to advantage in some places.
 

 

 

 

 

 

 

Reactor Temp. Fuel Pump Request

TR >1 3000F Pressure Overflow Tank Emergency
2outof3 >10 psig Level Not High Dump
Channels

 

 

 

 

 

 

 

 

 

 

 

 

 

Emergency
Dump
Demand

 

BLOCK DIAGRAM

Fig. 12,

 

 

 

181
S5A
=~ S5B
. ;F{ PS 380A OPEN WHEN
s’ FP PRESSURE
A >10 psig

 

K1C K2C

1 1 Im
[ 1

\: - K2D K3D
”

 

 

.~ = K33
o h__]'xunA
K181
181, 30, 127
137, 1088

EMERGENCY DUMP
WHEN DEENERGIZED

ENGINEERING ELEMENTARY

Control Circuit Drawings Used at MSRE.

ORNL-—-DWG 73—-3094
181

l NO EMERGENCY
DUMP REQUIRED

#- FP PRESSURE
a

Reactor Temp a
Not Hi

Channel 1 L 21

 

 
   

 

 OFT LEVEL
NORMAL

EMERGENCY
DUMP

DUMP WHEN
DEENERGIZED

JUMPER BOARD

£y
44

ORNL—DWG 73—3095

181

1

ISSA} EMER. DUMP SW.
=558 ON CONSOLE

¥ PS 380A OPEN WHEN FP PRESSURE IS HIGH

_@a
L L K2C

:E K1X I
E K1iD E K3C OPEN WHEN R TEMPATURE IS HIGH
K2D —— K3D

T _ __ T
@b

- K33
-
: c
K181

a 7

 

——d
—

OPEN WHEN OFT LEVEL IS LOW

 

 

 

 

(i), - JUMPER

 

 

30, 127
137, 1088
= SAFETY JUMPER
EMERGENCY DUMP CONTACT
ECC ACTION WHEN K181 IS DEENERGIZED
30 FILL RESTRICTION
127 THAWS FV 298
137 CLOSES HCV 355
1088 ANNUNCIATES EMERGENCY DUMP
MB8-1

 

 

 

Fig. 13. Suggested Type Drawings for Control Circuits.
45

C. Early in construction certain drawings and documents such as
flowsheets and most sections of the Design and Operations Report should
be assigned to personnel who are responsible for keeping these up to date.
Most of this should probably be assigned to future operating personnel.

D. During preparation for operations, a table was made listing all
known variables in alpha-numerical order. Columns were provided for indi-
cating the readout instrument number and location, the function of the
variable, the normal value, the alarm and control setpoints and proposed
logging frequency. Members of the design and development groups were con-
tacted to determine recommended initial conditions. This provided a very
good starting point for making up the various logs and aided in writing
operating procedures. Design and development personnel should be responsi-
ble for advising operations of any recommended additions or changes in
these.

E. Flowsheets and circuit elementary drawings should always precede
piping and wiring drawings. These should be approved by operations person-
nel. Operations personnel should follow construction or construction per-
sonnel should be thoroughly familiar with operating criteria. All lines,
valves, wires, and equipment should be labeled as they are installed.

F. During design of the MSRE, it was thought that extreme flexi-
bility would be needed for thermocouples and their readout instruments.
During operation it was obvious that the flexibility provided by the patch
panel was not necessary for a good part of the thermocouples. To keep
track of the over 1000 thermocouples at the MSRE presented a difficult prob-
lem but the system of two log books described earlier adequately solved the
problem.

G. The operating crew should be at the site long before the cells
are closed so that they can become familiar with the physical layout. A
model of in-cell equipment is valuable for operation and maintenance.

H. The desire of non-operating personnel to know the status of the
reactor often led to confusion in the control room especially at the start
of the work day. This was alleviated by posting a short report on the
control-room window giving the major happenings for the last 24 hours, the
important problems encountered and the present plans. Copies were also

made of the console log sheets each morning and placed in convenient locations.
46

I. The use of simulators for operator training was found to be very
useful. The value of this was enhanced by having the operator sit at the
reactor console, use the actual control switches, and refer to the regular
reactor instrumentation. Perhaps additional training could be done by
simulating failure of some of the control and/or safety instrumentation.

By running tests where the system is maloperated, the simulator would very
dramatically illustrate the consequences. |

J. Often it is difficult for the shift supervisor to find time to
thoroughly review the building log. To aid in this, a cover sheet was pro-
vided. The person taking the log listed items out of limits or items re-
quiring attention. (See Appendix K.)

K. Limiting the number of people who are authorized to change ope-
rating procedures is recommended. Keeping everyone informed of the changes
was cumbersome. These were entered in an operating procedures change book
and at first it was the responsibility of the shift supervisor to keep all
operators on his shift informed. Vacations and other leave made it diffi-
cult to be sure that all had been notified. Places were then provided in
the Operating Procedures Change Book for each operator to sign when he had
read and understood the change. This worked fine except that often there
was too much lag time. In reviewing the difficulty, it.was decided that
many of the changes were on check lists or for some other reason did not
need to be known by everyone. Therefore, only important changes were en-
tered in the Operating Procedure Change Book. Reading of these was strict-
ly enforced.

L. Graphic panels showing important flowsheets with valve position
and equipment status and jumper boards showing important control circuits
and their status were very helpful to the operators.

M. The tagging system used at the MSRE worked very well. All wvalves
. and switches, which if inadvertently operated could cause harm to personnel
or to the reactor, were tagged. The tags used are shown in Fig. 1l4. The
valves or switches were not operated without removing the tag which required
the approval of the shift supervisor on duty. Most of these were attached
as specified in the startup check lists. 1In general, we did not tag valves

or switches which were operated regularly.
Red Cardboard

Green Cardboard

'

 

 

 

©

 

KEEP VALVE CLOSED

 

 

OR SWITCH OFF

 

This is the normal position for this valve or switch
during operation. It should not be operated
without permission of the shift supervisor,

ltem No.: Date

Signed

UCN — 5923
3 7-64)

 

©

 

KEEP VALVE CLOSED

 

 

OR SWITCH OFF

 

This is the normal position for this valve or switch
during operation. It should not be operated
without permission of the shift supervisor.

Item No.: Date

Signed

 

UCN -~ 3924
8 74

 

 

 

 

Fig. 14.

Operations Tags.

LY
48

"Do Not Operate'' tags were used for tagging out valves and switches
for maintenance and special procedures. As indicated in Fig. 15, space
was provided on these for additional information. The white paper was re-
moved and attached to the maintenance work order or special procedure.
This aided in assuring that all tags were removed when the work was com-
plete. In general, the requirement to remove the tag was to have the shift
supervisor's permission. It was his responsibility to see that the work
was complete and that it was safe to operate the switch or valve.

N. The system used in selecting instrument, heater, and valve num-
bers at the MSRE was an aid to training and operation. Each system was
assigned a series of numbers (i.e., fuel system is 100 to 200). Any valves
in the line, instruments attached to the line or heaters installed on it
were numbered accordingly. (i.e., HV-522A was a hand valve in Line 522,
PS~-575C was a pressure switch attached to line 575, H-101-2 was a heater
on line 101, and TE-103~1 was a thermocouple on line 103.)

0. The training instructors were engineers and scientists who, in
most cases, were very familiar with their subject. However, some did not
understand the needs and the limitations of the operators and most had no
experience as teachers. It would be better if a few skilled instructors
were used, with the experts present to answer questions that the instructor
could not handle, These Instructors should be given some prior training
on such things as theory of learning, lesson planning, and methods of
presentation,

P. A training manual should be written. This would have been help-
ful in the early training and would have saved considerable time in train-
ing replacements. This manual should be updated as changes are made.

Since much of the nuclear, health physics, and general engineering
training needed is common to most reactors, avallable informatlon such as
the programmed instruction manuals published by the ORNL Operations Di-
vision32,33,3%,35,36 ghould be put to use. However, this should be care-
fully reviewed and the appropriate sections made a part of the training
manual for the particular reactor. The reactor design, instruments in-
stalled, and methods of operation should be covered from the operator's
viewpoint, To aid in learning and in handling emergencies and unusual oc-

currences, the reason for the various features should be stressed.
49

 

 

> RED CARDBOARD

 

 

UCN - 5928
(3 764

 

A

 

rltem No. ___ _  Work Order Ne.

 

Description _
Location
Reason for Tagging

 

 

 

 

 

WHITE PAPER WITH
RED CARDBOARD
BEHIND AND CARBON
Conditions Required to Remove Tag PAPER BETWEEN

 

 

 

 

 

 

 

 

 

 

 

Fig. 15. Maintenance Tags.
50

Instructions should be included for manipulative training and for
aiding the trainee in becoming familiar with the physical layout. Photo-
graphs of operating areas would be helpful.

Detailed outlines of tours and oral instruction should be provided.
Operating techniques such as changing chart paper, log-taking, etc.,
should be included.

The manual should be designed for the least educated operator and
should advance this person to the étage needed to operate the reactor.
Additional sections should be provided for the supervisors including the
art of supervision.

This manual would probably be more valuable if put into the form of
programmed instruction. Review tests should be included at the end of

each section.
1.

10.

11.

12.

13.

14,

15.

16.

51

REFERENCES

M. W. Rosenthal et al, The Development Status of Molten-Salt Breeder
Reactors, ORNL-4812 (August 1972).

P. N. Haubenreich and J. R. Engel, "Experience with the MSRE,"
Nuel. Appl. Tech. 8, 118 (1970).

M. W. Rosenthal, P. N. Haubenreich, H. E. McCoy, and L. E. McNeese,
"Current Progress in Molten-Salt Reactor Development," Atomic Energy
Review, IX, 601 (1971).

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

J. R. Tallackson, MSRE Design and Operations Report, Part IIA, Nuclear
and Process Instrumentation, ORNL-TM-729, Part IIA, (February 1968).

R. L. Moore, MSRE Design and Operations Report, Part IIB, Nuclear and
Process Instrumentation, ORNL-TM-729, Part IIB, (September 1972).

R. B. Lindauer, MSRE Design and Operations Report, Part VII, Fuel
Handling and Processing Plant, ORNL-TM-907 Revised, (December 1967).

B. H. Webster, Quality-Assurance Practices im Construction and
Maintenanee of the MSRE, ORNL-TM-2999 (April 1970).

MSR Progr. Semmiannu. Progr. Rep., July 31, 1964, ORNL-3708.

S. J. Ball, Simulators for Training Molten-Salt Reactor Experiment
Operators, ORNL-TM-1445, April 5, 1966.

R. H. Guymon, MSRE Design and Operations Report, Part VIII, Operating
Procedures, Vols. I and II, ORNL-TM-908, (Dec. 1965).

B. E. Prince, S. J. Ball, J. R. Engel, P. N. Haubenreich, T. W. Kerlin,
Zero-Power Physics Experiments on the MSRE, ORNL-4233 (Feb. 1968).

Operators' License, Code of Federal Regulations, Title 10, Part 55.

AEC Licensing Guide — Operators' Licenses — "A Guide for the Licensing
of Facility Operators, Including Senior Operators" published by the
Division of Reactor Licensing, USAEC.

R. H. Guymon, P. N. Haubenreich and J. R. Engel, MSRE Design and
Operations Report Part XI, Test Program, ORNL-TM-911, (Nov. 1966).

J. R. Engel, MSRE Design and Operations Report, Part XI-4, Test
Program for 233y Operation, ORNL-TM-2304, (Sept. 1968).
17.

18.

19.

20.

21.

22.

23,

24,

25.

26'

27.

28,

29.

30.

31.

52

J. R. Engel and P. N. Haubenreich, Use of On-Line Computer in Analysis
of MSRE Operation, ORNL-CF-62-3-26, (March 1962),

J. R. Tallackson, MSRE Design and Operations Report — Part II-A:
Nueclear and Process Instrumentation, ORNL-TM-729, Part II-A, (Feb. 1968).

R. L. Moore, MSRE Design and Operations Report — Part II-B: Nuclear
and Process Instrumentation, ORNL-TM-729, Part II-B, (Sept. 1972).

P. N. Haubenreich, et al., MSRE Design and Operations Report, Part III,
Nuclear Analysis, ORNL-TM-730, (Feb. 1964).

S. E. Beall, et al., MSRE Design and Operations Report, Part V, Reactor
Safety Analysis Report, ORNL-TM-732, (Aug. 1964).

P. N. Haubenreich et al., MSRE Desi%n and Operations Report, Part V-4,
Safety Analysis of Operation with 233U, ORNL-TM-2111, (Feb. 1968).

S. E. Beall and R. H. Guymon, MSRE Design and Operations Report: Part VI,
Operating Safety Limits for the Molten-Salt Experiment, ORNL-TM-733,
(April 1965).

S. E. Beall and R. H. Guymon, MSRE Design and Operations Report,
Part VI, Rev., Operating Safety Limits for the Molten-Salt Experiment,
ORNL-TM-733, (Aug. 1965).

S. E. Beall and R. H. Guymon, MSRE Design and Operations Report,
Part VI, 2nd Rev., Operating Safety Limits for the Molten-Salt
Experiment, ORNL-TM-733, (Sept. 1966).

R. H. Guymon and P. N. Haubenreich, MSRE Design and Operations Report,
Part VI, 3rd Rev., Operating Safety Limits for the Molten-Salt
Experiment, ORNL-TM-733, (July 1969).

R. B. Lindauer, MSRE Design and Operations Report, Part VII, Fuel
Handling and Processing Plant, ORNL-TM-907, (May 1965).

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

R. Blumberg and E. C. Hise, MSRE Design and Operations Report, Part X:
Maintenance Equipment and Procedures, ORNL-TM-910, (June 1968).

G. H. Burger, J. R. Engel, and C. D. Martin, Computer Manual for MSEE
Operators, ORNL-CF-67-1-28, (Jan. 1967).

R. L. Moore, Molten Salt Reactor Experiment Process Instrument Switch
Tabulation, ORNL-CF-65-6-5,
53

APPENDIX A

SCHEDULE FOR INITIAL TRAINING
(July 1964)

 

 

Day Time Subject Instructor
Monday
July 6 0800-1000 Information Meeting. Operations Dept. Head and
Organization chart, schedule, Operations Chief
general comments, questions,
and answers)
1015-1200 General Description of all Systems Operations Chief
1300-1400
1415-1500 Description of Area Shift Supervisor
1500-1630 Guided Tour Shift Supervisor
Tuesday
July 7 0800-0930 Guided Tour Shift Supervisor
0930-1015 Questions on Guided Tour Shift Supervisor
1030-1200 Reactor Chemistry Reactor Chemistry Division
Director and Chemist
1300-1500 Design of Fuel System Asst. Operations Chief
1500-1630 Self~Study
Wednesday
July 8 0800~0930  Reactor Physics Operations Dept. Head
Problems and Group Solving
0930~1130 Design of Coolant System Shift Supervisor
1130-1200 Self-~Study
1300-1400 Reactor Chemistry Chemist
1415-1600 Design of Fuel and Coolant Drain
Tank Systems Shift Supervisor
1600-1630 Self-Study
Thursday
July 9 0800-1130 Reactor Physics Operations Dept. Head
1130-1200 Self-Study
1300-1400 Design of Cover and Off-gas System Cover and Off-gas System
Design Engineer
1415-1630 Design of Fuel Pump and Oil System Pump and 0il System

Design Engineer
54

 

 

Day Time Subject Instructor
Friday
July 10 0800-0930 Reactor Physics Operations Dept. Head
0930-1030 Design of LKD System Leak Detector Design Engr.
1030-1200 Design of Cooling Water System Shift Supervisor
1300-1430 Design of Component Coolant System Shift Supervisor
1445-1630 Review Operations Chief
Monday
July 13 0800-0900 Design and Operation of
Instrument Air System Asst. Shift Supervisor
0900-1000 Health Physics Health Physicist
1015-1130 Beryllium Hazards Head of ORNL Industrial
Hygiene Dept. and Asst.
Director of ORNL Medical
Division
1130-1200 Self-Study
1300-1500 Instrumentation Instrumentation Engr.
1515-1630 Instruments — Explanation of each
in the field Instrumentation Engr.
Tuesday
July 14 0800-0930 Design and Operation of Shield and
Containment Systems Shift Supervisor
0945~1200 Explanation of Control Schematics  Operations Chief
1300-1400 Design and Operation of
Electrical System Shift Supervisor
1415-1530 Design and Operation of
Ventilation System Shift Supervisor
1530-1630  Self-Study
Wednesday
July 15 0800-0900 Electrical Systems Shift Supervisor
0900-0920 Instrument Air System Asst. Shift Supervisor
0920-0940 Water Systems Shift Supervisor
0940~1020 Component Cooling Systems Shift Supervisor
1030-1100 Shield and Containment Systems Shift Supervisor
1100-1130 Ventilation Systems Shift Supervisor
1130-1200 Leak Detector Systems Shift Supervisor
1300-1330 Cover and Off-gas Systems Shift Supervisor
1330-1400 0il Systems Asst. Operations Chief

-y
55

 

 

Day Time Subject Instructor
Wednesday
July 15
(continued)
1400-1500 Self-Study
1500-1530 Auxiliary System Startup
Check Lists Operations Chief
1530-1600  Purging Systems Asst. Operations Chief
1600-1630 Startup and Operation of Cover
and Off-gas Systems Shift Supervisor
Thursday
July 16 0800-0820 Heatup of Drain Tanks Shift Supervisor
0820-0850  Addition of Salt Asst. Operations Chief
0850-0920  Startup and Operation of
Lube 0il System Asst. Operations Chief
0920-0940 Heatup of Fuel and Coolant Systems Shift Supervisor
1000-1030 Prepare for Reactor Startup,
including Pressure Test Shift Supervisor
1030-1130 Filling Fuel and Coolant Systems Asst. Operations Chief
1130-1200 Self-Study
1300-1330 Starting Power Operation Operations Chief
1330-1430 Normal Operating Conditions Operations Chief
1445-1615 Nuclear Instrumentation Nuclear Instrumentation
Design Engineer
1615-1630 Self-Study
Friday
July 17 0800~0900  Unusual Operations Operations Chief
0900-1030 Heat Balance Analysis Chief
1045-1200 Control Rods Control Rod Design Engr.
1300-1400 Remote Maintenance Remote Maintenance
Design Engineer
1400-1530 General Operating Practices,
Plans, etc. Operations Chief
1530-1630 Self-Study
Monday
July 20 0800-0900 Discuss Precritical Testing Operations Chief
0900-1200 Self-Study, get clothing and lockers
56

 

 

Day Time Subject Instructor

Monday

July 20

(continued)
1300-1430 Reactor Physics Operations Dept. Head
1430-1630 Operational Checkout

Tuesday

July 21 0800-1000 Unusual Operating Conditions

and Control Circuits Operations Chief

1000~1100  Self-Study
1100-1200 Melton Valley Facilities Operations Chief
1300-1630 Operational Checkout

Wednesday

July 22 0800-1000 Reactor Physics Operations Dept. Head
1000-1100  Self-Study
1100-1200 Thermocouple Scanner Asst. Shift Supervisor
1300~1630 Operational Checkout

Thursday

July 23 0800-1000  Samplers Sampler Design Engineer
1000-1100 Self-Study
1100~-1200 Metallurgy Metallurgist
1300~1630 Operational Checkout

Friday

July 24 0800-1000 Control Schematics Operations Chief
1000-1100 Self-Study
1100~1200 Discussion Operations Chief
1300-1630  Operational Checkout

Monday

July 27 0800-0930  Chemical Processing of the Fuel

Salt Chemical Processing Engr.

0930-1030 Control Shematics Operations Chief
1030-1200  Fuel Salt Production Salt Production Engr.
1300-1630 Operational Checkout

Tuesday

July 28 0800-0930 Electrical System Shift Supervisor
0930-1030 Self<Study
1030-~1220 Waste System Shift Supervisor
1300~1630 Operational Checkout
57

APPENDIX B

INSTRUCTIONS FOR FLOWSHEETS CHECKOUT

First Phase

Using a copy of all flowsheets and all instrument application drawings

needed for the system being checked, proceed as follows:

1.

2.

3.

4.

NOTE:
(a)
(b)
(c)

(d)
(e)
(£)

Trace all lines and equipment including instrumentation at least
to the primary sensing element.

Mark flowsheets and instrument application drawings with yellow
pencil as lines are traced. Do not miss anything.

Change flowsheets and instrument application drawings to agree
with the actual installation. Use green pencil for deletions
and red pencil for additions,
Tag all lines, instruments and equipment at locations where perma-
nent tags are needed. (Use round tags and green pencil. Tags
should be exactly like desired permanent label. Line numbers
should not be preceded by any symbol. Hand valves should be
labeled V-501, V-944, etc. Control valves should be HCV-516,
PCV-522, etc, Instruments should be PI-500, FI-933, etc.)
Some special things to look for are given below:
Location of tees, sample valves and capped lines.

Double lines and containment extensions.

Instrument valves not shown or numbered, hand valves and
check wvalves.

Leak detector lines.
Areas where items are located.

Be on the lookout for any faulty construction, or errors in
design or construction which might cause mal~operation.
58

Second Phase

Using a copy of the small size flowsheets:

1. Trace out lines under the direction of the "system flowsheet
experts."

2. When completed, return to conference room for self-study until

the other group has returned. Then continue to next flowsheet
checkout.

NOTE: It will not be necessary to trace every line to check for tees,
sample valves, etc., however, each person should be familiar
with each system when the checkout is finished.
59

APPENDIX C

SCHEDULE FOR PRECRITICAL TRAINING

(March 1965)

All operating personnel participated in the sessions from March 15 through

March 26 unless otherwise noted,

from March 29 through April 2 unless otherwise noted.

Only the engineers participated in the sessions

 

 

Day Time Subject Instructor
Monday
March 15 0830-1000 Discussion of Status and Plans Operations Dept. Head &
Operations Chief
1030-1630 System Discussions (Brief Reviews)
Fuel System Asst. Operations Chief
Coolant System Shift Supervisor
Fuel Drain Tank System Shift Supervisor
Cover and Off-gas System Asst. Shift Supervisor
0il System Asst. Shift Supervisor
Component Coolant System Shift Supervisor
Water System Shift Supervisor
Instrument Air System Asst. Shift Supervisor
Shield and Containment System Shift Supervisor
Ventilation System Asst., Shift Supervisor
Waste System Shift Supervisor
Tuesday
March 16 0830-1000 Radiation Safety Policy Head of ORNL Radiation
Safety
1000-~1130 Annunciators Instrumentation Engineer
1230-1330 Process Monitors (and High Level Process Monitor
Gamma) Design Engineer
1330-1400 Ventilation System Asst. Shift Supervisor
1500-1630 Instrument Tour (Group 1) Instrumentation Engineer
Portable Radiation Instruments
(Group 2) Health Physicist
Wednesday
March 17 0830-1000 Radiation Safety Health Physicist
1030-1200 Personnel Monitors Personnel Monitor
Design Engineer
1300-1500 Electrical Systems Shift Supervisor
60

 

 

Day Time Subject Instructor
Wednesday
March 17
(continued)
1500-1630 Instrument Tour (Group 2) Instrumentation Engineer
Portable Radiation Instruments
(Group 1) Health Physicist
Thursday
March 18 0815-1015 Control Circuits Operations Chief
1030-1220 Computer Analysis Chief
1300-1430 Freeze Valves Freeze Valve Design Engr.
1500-1630 Tours:
Chemical Processing (Group 1) Chem. Processing Engr.
Instruments (Group 2) Instrumentation Engr.
Friday
March 19 0830-0930 Safety Circuit Testing (Process) Safety Instrumentation
Design Engineers
0930-1030 Discussion of Pool Criticality Pool Criticality
Assembly Experiment Assembly Engineer
1100-1200  Jumper Board Asst. Operations Chief
1300-~1400 Safety Circuit Testing (Nuclear) Safety Instrumentation
Design Engineers
1500-1630 Tours:
Chemical Processing (Group 2) Chem. Processing Engr.
Instruments (Group 1) Instrumentation Engr.
Monday
March 22  0800-1200 PCA Experiment (Group A) Pool Criticality
Assembly Engineer
Chemical Processing (Group B) Chem. Processing Engr.
0830-1200 PCA Experiment (Group F) Pool Criticality
Assembly Engineer
1230-1630 PCA Experiment (Group C) Pool Criticality
Assembly Engineer
Sampler (Group D) Sampler Design Engineer
Tuesday
March 23  1230-1630 PCA Experiment (Group F) Pool Criticality
Assembly Engineer
Chemical Processing (Group B) Chem. Processing Engr.
NOTE: Crews will spend spare time Monday and Tuesday on group or individual study.
61

 

 

Day Time Subject Instructor
Wednesday
March 24 0830-0930 PCA Discussion Shift Supervisor
0930-1200 Start-up Check Lists (Brief
Summaries)
4A Electrical Shift Supervisor
4B Instrument Air Asst. Shift Supervisor
4C Water Shift Supervisor
4D Component Cooling . Shift Supervisor
4E Shield and Containment Shift Supervisor
4F Ventilation Shift Supervisor
4G Leak Detector Asst. Shift Supervisor
4H Instrumentation Asst. Shift Supervisor
41 Freeze Valves Asst. Shift Supervisor
1300-1330 5A Purging Systems Asst. Shift Supervisor
1330-1400 5B Start up of Cover Gas Asst. Shift Supervisor
1400-1445 5C Heatup of Drain Tanks Asst. Operations Chief
1500-~1630 Tours:
- Instrument System (Group 1) Instrumentation Engr.
Electrical System (Group 2) Shift Supervisor
Thursday
March 25 0830-0845 5E Startup of Lube 0il System Asst. Shift Supervisor
0845-0930 5F Heatup of Fuel and Coolant
System Asst. Operations Chief
0930-1000 5G Prepare DT's for Startup Shift Supervisor
1030~1100 5H Routine Pressure Test Asst. Shift Supervisor
1100-~1200 5T Filling Fuel and Coolant System Asst. Operations Chief
1300-~1400 Criticality and Power Operations Dept. Head
1400-1500 Criticality and Power Shift Supervisor
1500-1630 Tours:
Instrument System (Group 2) Instrumentation Engr.
Electrical System (Group 1) Shift Supervisor
Friday
March 26 0830-1000 Normal Operation Shift Supervisor
1030-1200 Shutdown and Transfers Shift Supervisor
1300-1400 Sampling and Analysis Asst. Shift Supervisor
1400~1500 Data Taking, Operating Techniques, |
etc. Asst. Operations Chief
1530-1630 Plans Operations Chief
62

 

 

Day Time Subject Instructor
Monday
March 29 0830-1030 Nuclear Aspects of MSRE Operations Dept. Head
1100-1200 Control Rods Control Rod Design
Engineers
1300-1400  Computer Analysis Chief
1530-1630 Reactor Chemistry Discussion Chemist
Tuesday
March 30 0830-1030 Nuclear Aspects of MSRE Operations Dept. Head
1100-1200 Core Specimen Arrays Remote Maintenance Engrs.
1300-1500  Computer Analysis Chief
1530-1630 Maintenance Discussion Remote Maintenance Engr.
Wednesday
March 31 0830-1200 Nuclear Instrumentation Nuclear Instrument
Design Engineer
1300-1500 Nuclear Simulation Instrument Engineer
1500-1630 Computer Analysis Chief
Thursday
April 1 0830-1030 Computer Analysis Chief
1100~1200 Critical Experiments Analysis Chief
1300-1400 Critical Experiments Analysis Chief
1430~1630 Nuclear Instrumentation Nuclear Imstrument
Design Engineer
Friday
April 2 0830-1030 Nuclear Aspects Operations Dept. Head
“ 1100-1200  Operating Limits Operations Chief
1300-1400 Nuclear Aspects Operations Dept Head
1400-1500 Emergency Plans Operations Chief
1530-1630 Plans Operations Chief
63

Appendix D
PRECRITICAL EXAMINATIONS

In April, 1965, six weeks before the beginning of nuclear operation,
written examinations were administered to all MSRE operators. This ap-
pendix reproduces the instructions and the schedule issued just prior to
the examinations. Some representative questions and problems are also

included. (The complete examination covered 43 pages.)
Instructions

This examination is to be given in five periods plus one extra period
for the engineers. Various type questions are asked throughout. True-
false questions are worth one point (+1) if right, zero (0) points if not
answered and minus one (—1) if answered wrong. All other types have dif-
ferent values as noted to the left of each question number. A summary of
the exams and time allocated for each period is given in the attached
table.

It should be noted that the value of the question or the number of
blank lines left for answers does not necessarily indicate how many correct
answers there are. For instance if there are 3 things which turn off the
control-room air conditioner and 5 lines are left, the question may still
be worth 1 or 2 points.

There are some questions which are marked extra credit. Skip these
until all other questions are answered. Then if time permits go back and
complete these. It is more desirable to answer the required questions right
than to answer the optional ones.

Be sure to read all instructions throughout the test.

In general, it is felt that the meaning of each question is clear. Put
down the best answer that you know, based on your understanding of it. Do
not ask for assistance. If you feel it is necessary to explain your answer,
do so. Credit will be given if a question is not clear, however, no credit
will be given if it appears that the question was clear and the intent was
changed by your explanation even though your answer to your explanation is

correct, i,e., you cannot make up your own questions.
64

When finished be sure your name and date are on the first sheet and
your initials are on the bottom of each sheet. Do all calculations, ex-
planations etc. in the margin or on the back of the examihation sheets.
Turn in your papers promptly when requested. The next test period will
start promptly at the time indicated in the tables. If you are late,

you will not have as much time for that test period.
65

Schedule

(All operators take tests in periods 1—5. Only engineers take tests

in period 6.)

 

 

Required
Period Duration Time Subjects Covered Questions
(min) No. Value
0815-0955 .
1 100 (1615-1755) Instrumentation 94 165
1005-1145 .
2 100 (1805-1945) Flow Sheets & Operating Parameters 55 176
3 50 1215-1305 Design 40 50
(2015-2105) Health Physics & Industrial Hygiene 27 37
Nuclear 6 6
Chemistry 8 14
General 0 0
Chemical Processing 3 3
4 60 1315-~1415 Lay Out 6 27
(2115-2215) Engineering Calculations 7 12
Electrical 8 14
Emergency Procedures 9 9
Chemical Processing 10 13
or Samplers 8 13
1425~1450 . s
5 25 (2225-2250) Nuclear Characteristics 27 40
%
Total of above 298 566
General 20 49
1500-1600 Computer 10 10
6 60 (2300-2400) Abnormal Operating Procedures 7 10
Nuclear Characteristics 20 28
Total of this part 57 97

 

%
Either Chemical Processing or Sampling Section of the test is to be taken.
66

Samples of Questions

 

 

Instrumentation
1. High reactor outlet temperature (f1300°F) causes a rod
scram. True False
5. Reactor cell activity will cause a fuel drain. True False
High reactor outlet temperature will cause a load scram. True False
16. Reactor cell air activity closes all air line block valves.
True False
31. All three fill valves (104, 105, and 106) must be thawed
to be in the operate mode. True False
39. If the power level is changing circuits will not permit
going into the run mode. (Period <30 sec) True False
54. Automatic load control will not raise the radiator doors
above the lower intermediate limit unless the reactor
is in the run mode. True False
62. When in servo the regulating rod limit switches can be
raised but not inserted using the manual rod actuator
switch (S5-19) on the console. True False
78. Stack fan #2 will start automatically only if the start
button has been pushed after starging #1 stack fan. True False
84. There are a number of annunciators at the MSRE. Check
which of the following will cause an annunciation.
(A list of 36 conditions followed. One point was scored
for each correctly marked.)
85. Assuming that FV-103 is frozen on automatic, fuel drain
is caused by:
(Five blanks. Question worth 10 points.)
93. 1If the reactor is at full power the following may stop

the fuel pump.

 

 
67

Flow Sheets and Operating Parameters
4. The various systems have series of numbers assigned to
various lines. Place the number 100, 200, 300, after
each of the following.
(a) Fuel System
(b) Coolant System
(c) Cover Gas
(d) Containment Ventilation
(e) 0il System
(f) TFuel Processing
(g) Waste
(h) Leak Detector
(i) Water
10. The drain tank condensers are cooled by circulating water through
the tubes. Place the proper letter in the following blanks.
Normal cooling water - (a) Potable
Emergency cooling water (b) Process
(¢) Cooling Tower
(d) Treated
(e) Condensate
28. At fuel pump levels above approximately 7 there is entrainment
or carryover to the overflow tank.
41. Check which of the following are monitored by process radiation
monitors.
(a) Cell air on line 565.
(b) Cell air in the cell by high level gamma meter.
{(c) Coolant pump off-gas.,
(d) Helium supply to bubblers.
(e) Off~gas from charcoal beds.
(f) Off-gas upstream of the charcoal beds.
(g) Fuel o0il supply tank.
(h) Coolant system oil supply tank.
(i) Cooling tower water.

(j) Reactor and drain tank sump jet discharge.
46.

50.

68

There are two levels of annunciation on the monitron. These are
normally set at mr/hr and mr/hr.
Assume that the reactor has been filled from FD-2 and is operating
at a few watts after the criticality tests. Check the items below
which are considered to be normal. (Watch out for units.)
(a) TFuel salt temperature “1200°F.
(b) Coolant salt temperature Vv1200°F.
(¢) Fuel salt flow 1200 gpm.
(d) Fuel salt indicated by FT-201.
(e) F. P. power —25 kW.
(f) Both FP bubblers in operation.
(g) Flow indicators in supply to bubblers is 25 psig.
(h) Main flow to FP (FIC-516) = 4.2 &/min.
(i) Gas flow to OCT-1 would be ~50% of MCB.
(j) Water flow to FP = 5 gpm.
(k) FOP-1 in operation, FOP-2 in standby.
(1) 0il flow to lubricate FP bearings = 3~5 &/m.
(m) CCP 1 or 2 on.
(n) One RC cooler on, other in standby.
(o) Both drain tank cell coolers on.
(p) Component coolant air on the CP shroud.
(q) Pd4IC 960 set at 8 psi AP.
(r) CP power reading approximately 38 W.
{s) CP speed is approximately 1750 rpm.
(t) Main purge to CP (FIC-512) reads 0.37 2/min.
{u) FV-103 frozen.
(v) TFV-106 deep frozen.
(w) FV-105 thawed.
(x) FV-108 thawed.
{y) FV-107 and 109 deep frozen.
(z) FV-204 thawed.
(aa) FV-206 frozen.
(bb) FI-201 reads 407%.
(ce) Equalizer 545 open between FD-2 and FP.

{dd) Other equalizer valves closed.
(ee)
(££)
(88)
(hh)
(ii)
(33)
(kk)
(11)
(mm)
(nn)
(00)
(pp)
(qq)

(rr)

(ss)
(tt)
{uu)

(vv)

(ww)

(xx)

(yy)
(zz)
(aaa)
(bbb)

(cee)
(ddd)
(eee)
(f££)
(ggg)
(hhh)

69

FFT vent (577) open.

All drain tank helium supply valves open (572, 574, 576).

CDT equalizer (527) open.

Approximately 5000# of salt in FFT.

FD-2 practically empty.

Feedwater tanks full.

Steam dome level indicators at greater than 70%.

FP vent open (533).

AP across main charcoal bed 4 in. of water.

Valves in line 566 open.

High bay at a slight negative pressure.

Stack fan No. 2 on.

Stack fan No. 1 in standby.

Helium being fed from helium trailer whose pressure is above
500 psig.

Helium dryers at 800°F.

Oxygen removal units at 1200°F.

FCV-500J indicating less than 10 2/m.

Water content of cover gas less than 1 ppm, oxygen content
less than 6 ppm.

Lube o0il system at 5 psig.

Lube o0il supply tank level (0T-1) setpoint set at V17 above level
indicated in the tank.

Reactor and drain tank cell sumps empty.

Three coolant cell coolers on.

No. 1. or No. 2 cooling tower pump on.

No. 1 or No. 2 treated water pumps on and the other set for
automatic startup.

Process water flowing through the drain tank condensers.

All "A" valves to the flanges from the leak detector system open.

All "B" valves to the flanges closed.

LKD pressure greater than 90 psig.

FD-1 may be at 1200°F or may be cooled if desired.

Fuel system overpressure will be maintained at approximately

5 psig.
(iii)

(333)
(kkk)

(111)

(romm)
(nnn)

(000)

(pPp)
(qqq)

(rrr)

(sss)

(ttt)
(uuu)
(vvv)
(www)
(xxx)

(zzz2)

Design

3.

70

Coolant system will be at a slightly higher overpressure to
assure that any leakage at the heat exchanger will be from the
coolant system to the fuel system.

All three containment air filters will be on stream.

0il level in the o0il catch tank should be just above the
intersection between the smaller and larger sections.

Auxiliary air compressor will be in operation with CCP #3 in
standby for cooling FV 204 and 206.

Either instrument air compressor can be in operation.

Vapor condensing system will be vented to the stack filters.

The bypass valve around the vapor condensing system rupture
discs should be closed.

MG sets 1 and 4 will be in operation and either MG 2 or 3.

DG-5 should be in operatiom.

The computer will be in operation at all times. If trouble
develops with it, the reactor will be shut down.

All process and personnel radiation monitors will be in
operation.

An engineer must be in the control room.

The waste tank blower will be on.

HFIR feeder line will be in parallel with the normal TVA feeder.

Radiator heaters will be on,

Reactor cell will be at —2 in. of water.

Helium surge tank will be on stream.

Fuel pump flow is approximately

(a)
(b)
(c)
(d)

850 gpm

1200 gpm
1800 gpm
3000 gpm

Volume of reactor plus drain tank cells is

(a)
(b)
(c)
(d)

12000 cubic ft
18000 cubic ft.
24000 cubic ft.
1200 cubic ft.
17.
33.

43.

71

(e) 1800 cubic ft.

(£f) 2400 cubic ft.

Normal motor driven speed of the control rods is in./sec.
What is provided to indicate a plugged radiator tube during power

operation?

 

At 10 MW the largest item in the reactor heat balance is determined by
(a) Measuring the heater amperes and converting to power.
(b) Determining the AT across the reactor and converting this to power
using the fuel salt flow rate.
(¢) Calculating the heat loss at the radiator based on salt AT and
and coolant salt flow.
f{d) Calculating the heat loss at the radiator based on the air AT

and the air flow up the coolant stack.

Health Physics and Industrial Hygiene

 

7. The maximum recommended weekly radiation dose is mrems.
9. How much concrete (TLV = 11 inches) would be required to reduce
radiation from lr/hr to 10 mr/hr?

16, Readings are taken of a radiation field and we find that it is 300 rads/hr
for fast neutrons which have a RBE of 10 and 2 rads per hour for Gamma
which have a RBE of 1. If we stay in this field for 6 minutes, how much
dose would we receive? Be sure to include proper units in answer.

20. Approximately 1/2 of the people who receive a one-time dose of rem
will die.

26. The maximum permissible concentration for acute exposures of beryllium
is ug/m®. (micrograms per cubic meter)

Chemistry
1. The fuel salt to be used initially in the MSRE has the following

composition. Fill in proper letter.

BeF, (a) 5 mole %
LiF (b) 0.9 mole %
ZrF., (c¢) 65 mole %

UF, (d) 29.1 mole %
6. Zr0O; starts to precipitate at

72

8. Fuel salt liquidus temperature is approximately

(a) 100 ppm
(b) 200 ppm
(¢) 300 ppm
(d) 400 ppm
(e) 500 ppm
(a) 650°F
(b) 750°F
(e) 850°F

Laxout

6. Fill in the blanks with the proper letters. Not all letters need be

used and any letter can be used more than once.

OCT Level indicators
CT Pumps Start Button

Lube o0il pumps

Off-gas radiation monitor

(RE-557)

Scanner Switches

Fuel Processing Heater

Controls

Sampler Enricher

FP pressure test switch

Bubbler Test Switch

HS 519

HS 523

Test lo helium safety
switches

CCP No. 3

CCP No. 1

Vapor Condensing Rupture Disc

CW rotameter for lube oil

0il system

Emergency Air usage

Rotameter

MG No. 1

—

(a)
(b)
(c)
(d)
(e)
(£)
(g)
(h)
(1)
(3)
(k)
(L)
(m)
(n)
)

(p)
(q)

Main Control Room

Auxiliary Control Room
Transmitter Room

North Electrical Service Area
South Electrical Service Area
High Bay

Stack Area

Vent house

Water Room

Service Room or Tunnel

Blower House

Heater Panels

Diesel House

Switch House

840 level other than transmitter
room and heater panels

Special Equipment Room

Coolant Drain Tank Cell
73

MG No. 2
250v battery bank
PCV-528
PCV-522

Reactor Physics and MSRE Nuclear Characteristics

7.

13.

14.

29.

t/T

If the flux is on a stable period, it varies as e . The value of e
is . (3.14,Y10, 2.72, 1.414)

In the foregoing situation, if the flux doubles every 21 seconds,
what is the value of T?

1 keff
what was k before the change?

eff

When the fuel pump is started, keff decreases because of

 

changes b6 0.01 and the subcritical multiplication doubles,

 

and
The reason why the fuel salt is critical in the core, but safely sub-

critical in the drain tank is

 
75

APPENDIX E

SCHEDULE FOR PRE~POWER TRAINING SESSTON®

 

 

 

Day Time Subjectb
Mon. 0800-1200 OP2 (Nuclear Aspects of Operation)
1230-1630 OP3A-0P3J (Opeeration of Auxiliary Systems)
Tu. 0800-1200  OP3K-0P5H (Operation and Startup of Auxiliary Systems,
Reactor Startup)
1230-1630 OP5I-0P6 (Reactor Startup, Sampling and Additions)
Wed. 0800-1000 OP7-0P9D (Heat Balance, Instrument Calibrations,
Unusual Operating Conditions)
1015~1200 Instrumentation
1230-1630 Nuclear Power Operation
Th. 0800~1000 On-Line Computer and Data Logger
1000-~1200 Self-~Study
1230-1330 Reactor Simulator
1330~1430 OP9D-0P9H (Unusual Operating Conditions)
1430~1630 On-Line Computer
Fri. 08001630 Shutdown Procedures, Plans for Power Ascension, Miscel-
laneous Questions on Procedures and Plans
“Attended by all MSRE operations personnel. Instruction by MSRE
engineers.

bOP refers to sections of Operating Procedures (ORNL-TM-908).
77

APPENDIX F
PREPOWER EXAMINATION

In October 1965, prior to power operation, additional written exami-

nations were given to all operators. Some representative questions are
given below.

All Operators

2.

l6.
21.

49,

55.

What variables affect reactivity in core? (Name 4.)

 

If you increase concentration of *2°U by 1%, how much will reactivity
increase?

What stream does the Be monitor in the vent house sample?

Since fission chamber confidence will not be established

before filling the reactor, interlocks are provided which

prevent withdrawing the rods unless you have safety

chamber confidence. True or False
Automatic rod scram is caused by:

(a)
(b)
(e)
(d)

According to the operating safety limits, underline greater than or

 

 

 

 

less than and £fill in blanks.

(a) Cover-gas pressure must be greater than or less than .

(b) The maximum cell leakage should be greater than or less than
calculated for condition of maximum credible accident.

(c) The steady state power should be greater than or less than

(d) The safety temperature level rod scram should be greater than or

less than

Engineers only

2.

If the regulating rod is near its maximum sensitivity, how much dif-
ference would there be between critical position circulating and not
circulating (assume shim rods do not move). inches. The

rod would be inserted further when salt is circulating or not circu-

lating?

 
11.

15,

78

The reactor is operating at 2 MW with the reactor inlet temperature
of 1210°F. Calculate the reactivity change required to compensate
for temperature effects in going to 8 MW and outlet temperature of
1200°F. (Use predicted values of temperature coefficients.
Sk/k.
How far will the regulating rod have to be moved and in which direction?

(Assume it has maximum value.)

 

Describe the system of getting repairs or minor changes done at the
MSRE starting with the punch list and ending with a completed and

checked-out job.

 

According to the "Operating Safety Limits'" how many safety channels

must be in servide during reactor operation?
79

APPENDIX G

CHIEF OPERATOR TRAINING SCHEDULE
(October and November, 1967)

 

Approximate
Time Subject and Items Covered
Allocated

Instructor

 

1 hour Discussion: Justification for removing
engineers from shifts, training schedule,
suggested general subjects for review.

4 hours Unusual Operating Conditions and Alarms:
Review of Section 9 and important annunci-
ators. Cover power outage, action to be
taken, and consequences.

3 hours Information Available and Use of It: Cover
stick file, switch tabulation, line
schedule, calibration curves, etc. Give
problems to be worked.

3 hours Instruments and Controls: Cover important
circuits and reasons for these. Include
fill restrict, drain, rod scram, load
scram, modes, FP, and CP circuits.

2 hours Nuclear Instruments: Cover types installed,
what each is used for and how trips are
tested. Explain flux servo and temperature
servo.

 

2 hours Health Physics and Emergency Plans: Cover
definition of terms, monitoring, use of
instruments, permissible exposures,
shielding, evacuation matrix, stack moni-
toring, and handling emergencies.

2 hours Nuclear Aspects: Cover neutron balance, de-
layed neutrons, reactivity, criticality
sources and limits of deviation from calcu-
lated reactivity. Explain reason for and
values of wvarious coefficients of reac-
tivity.

Operations Chief

Shift Supervisor

Asst Operations Chief

Asst Operations Chief

Operations Chief

Operations Chief and
Health Physicist

Operations Dept Head
or Analysis Chief
80

 

 

 

 

 

Approximate
Time Subject and Items Covered Instructor

Allocated

2 hours Computer: Cover special operation and Analysis Chief
operator duties as well as failures and
what to do.

2 hours Normal Operating Conditions: Review 5K and Operations Chief
the logs. Tie in all systems and emphasize
the interrelationship.

1 hour Operating Safety Limits: Review these and Shift Supervisor
answer questions.

1 hour Shutting Down the Reactor: Point out dif- Asst Operations Chief
ficulties which might be encountered.
Suggest places where each could use more
self-study.

1 hour Miscellaneous: Review previous tests and Operations Chief

suggest areas for self-study. Answer
questions.
81

APPENDIX H

CHIEF OPERATOR REVIEW TEST

In the fall of 1967 prior to certifying selected technicians as chief
operators, a 2-1/2 hour closed book and a 1-1/2 hour open book written ex-
amination was given to each. Sample questions from these are given below.
Closed Book

1. Put the proper letter or letters in the blanks.
A Compensated ion chamber
B BF3 chamber
C Fission chamber
D

Uncompensated ion chamber

(a) Safety channel

(b) Linear channel

(¢) Log P

(d) Used for flux servo

(e) Provide sag bypass

(f) Used for temperature servo

(g) Used for inverse count rate plat
(h) We have three of these

(i) Has automatic position control

(i) Connected with range.seal

(k) Called wide-range counting channel
(1) Gives a 5-sec rod reverse

(m) Gives a 9-MW rod reverse

(n) Confidence required for early fill
(o) 500-kW run permit

3. An unshielded gamma source is placed at the southeast corner of the

 

drain tank cell at the 852' elevation and it does not give a high
level alarm on the Chem Plant monitron. What is the approximate
maximum radiation reading at the north end of the drain tank cell
(852 ft elevation)?

Show assumptions and calculations.
82

10. The following vent to the (A) MCB (B) ACB or (C) direct to the stack.

Put proper letter following each.

FP off-gas (522) L F. Drain tanks .
CP off-gas (528) L FP o0il leakage (524)
C. Drain Tanks L CP oil leakage (526)

17. 9 mg/mf = ppm. |

33. What is the full-load operating life of the 250-V battery after the
loss of TVA power?

 

47. Name four variables that have important effects on reactivity or k.

State briefly how each affects neutrons and hence k.

 

Open Book
Use any information available in the Operations Office but do not

consult with anyone. Include information as to how you arrived at the
answer. (That is, remembering circuits is not sufficient.) Include
drawing numbers, circuit numbers, etc.
1. A requirement of the run mode circuit is that the FP be running.
In this circuit this requirement is met by which of these:
(a) Both of two normal speed signals.
(b) Either of two pump motor current normal.
(¢) Either of two normal speed signals.
(d) Two of three normal speed signals.
5. According to the building log, the normal level in the feedwater tanks
(FWT) is 46%. What quantity of water is this in gallons?
9. XPR-201A records radiator power.

(a) Where is this located?

 

(b) What is the chart speed?

 

(c) What does it read full scale?

 

(d) This recorder should be how accurate? * MW
10. Determine the following on Line 991.

Size

Schedule

Material
83

APPENDIX I

MSRE OPERATOR EXAMINATION — May 1969

Point
Value

A, Principles of Reactor Operation
(8) 1. Define the followings:

a, Criticality

b. Chain Reaction
¢, Subcritical

d., Moderator

e, Fission products
f. K

eff
g. Reactivity
h. Period
(2) 2, Explain the reactivity balance calculated by the computer.

What items are used in the calculation and what is the mean-
“ing of the final value (i.e., KNET)°

(2) 3. Explain why the count rates on the counting channels increase
as the reactor is filled. How can we predict that we will
not go critical before the reactor is completely filled?

(6) 4, What is the reactivity effect of each of the following (de-
crease or increase). Explain each.

a, Fuel burnup

b. Decrease in temperature
¢, Removal of bubbles from the core

d. Buildup of fission products
e. Starting the fuel pump
f. Insertion of control rod

(2) 5, Explain the importance of delayed neutrons to the control
of the reactor.

(2) 6. What is meant by ""the reactor is on a stable period?" Ex-
plain how values for stable periods can be calculated.

(22)
Point
Value

(3)

(2)

(2)

)

84

Operating Characteristics and Features of Design

1.

2.

Two modes of servo rod control were provided at the MSRE.
Explain each, when used, and the need for the two types.

Describe the reactor vessel and reason for various features,

Describe the heat removal system of the MSRE and the func-
tions of the components. (Start at the coolant salt lines
outside the reactor cell.)

Ventilation and off-gas from various systems must be properly
handled to avoid undersirable releases of activity. Connect
up the lines on Figure B-4 attached and put arrows to show
direction of flow. Indicate important line numbers, valves,
and interconnecting lines.
 

 

 

 

- MA N >
| CHAR oA B

OFF &Gag
SAarnPL £ ’

 

 

 

 

P SHMPLER »
EvmiIcHER |

 

 

 

 

I-/: e

R"\)-\V\

 

 

 

 

 

 

 

 

 

 

 

 

e
SouTwi
‘5'- o
. ST AKX

FAN>

 

Figure B-4

<8
86

Point
Value

C. General Operating Characteristics

(3) 1. Assuming that the system is heated and coolant salt is cir-
culating, describe the steps taken to fill the reactor and
start fuel salt circulating. At each step include normal
response of pertinent variables and precautions.

(3) 2., Describe and explain briefly the response of the following
to a complete power outage (i.e., loss of TVA). Assume that
the reactor is operating at full power and the diesels can-
not be started.

a. Load and nuclear power

b. Fuel system temperatures

¢, Temperatures at FV-103, 105, 204, and 206
d. O0il flow to FP shield and bearings

(3) 3. Describe what happens to the reactor inlet and outlet tempera-
ture and to the nuclear power for each of the following
changes. (Assume no operator action except as noted.)

a. In servo rod control with one blower on, both doors open
and the bypass damper open, what happens when the other
blower is started?

b. In manual rod control with both blowers on, both doors
open, and bypass damper nearly closed, what happens when
one blower is stopped?

c. In servo rod control at v10 kW with both blowers off,
both doors closed, and bypass damper opne, what happens
when both doors are opened about 50%7?

(2) 4. Assume that the fuel pump is off, system is filled with fuel
salt and at temperature, The operator attempts to take the
reactor critical and to 100 kW. He proceeds as follows:

a, Inserts all rods

b. Starts FP

c. Switches to manual rod control

d. Raises Rods 2 and 3 to 40 in.

e, Raises Rod no. 1 slowly to take the reactor critical and
to a few watts

f. Raises Rod No. 1 to get on a 100-sec period and levels
the power at 100 kW,

What interlocks prevent him from doing this and what is the
purpose of this interlock?

(11)
Point
Value

(4)

(2)
(2)
(3)
(2)

(3)

(16)

87

Instrumentation and Control

lfl

Interlocks require that certain nuclear instruments are in
service during a fill of the reactor with fuel salt.

a, Explain which are required and why.
b. Does the fill section of the operating procedures (5-I)

require any other chambers to be in service?

Interlocks require what nuclear instruments to be in service
at 8 MW? Explain.

Why was the coincidence type safety system chosen for use
at MSRE?

What signals will automatically cause a rod scram? Explain
the reason for each.

What provisions are made to remove afterheat. What instru-
mentation assures this will function.

Describe the emergency fuel drain circuit and explain the
reason for each interlock and the matrix used (i.e., 2 out
of 3, 2 0of 2, 1 of 2, 1 of 1, etc,).
Point
Value

(3)

(3)
(2)
(3)

(2)

(13)

88

Safety and Emergency System

1.

Double containment is required at the MSRE. During normal
power operation (no sampling being done) describe the primary
and secondary containment for the FP and all connecting
lines.

Describe the emergency electrical systems,
Describe the normal and emergency instrument air system.

Briefly describe the building evacuation procedure for the
MSRE.

What provisions are made to assure that we always have a
stack fan in service? Describe power supply, interlocks,
dampers (at stack fan), etc.
Point
Value

(2)

(3)

(3)

(5)

(2)

(15)

89

Standard and Emergency Operating Procedures

1.

2.

Describe the procedure for burping the OFT and precautions
in case 522 is plugged.

During full~power operation, you receive a low radiator
outlet temperature alarm,

a. What action should you take to prevent it continuing to
to decrease?

b, If no operator action is taken and temperatures continue
to decrease, what control action will occur?

c. Describe waht will happen as a result of this and what
things the operator should do.

What areas cannot be entered due to high radiation when we
are at full power? What is the main source of the radiation
in each of these areas?

While operating at full power, a 5-min TVA power outage occurs,
Describe what must be done in the approximate sequence that
it should be done.

Describe sampling of the fuel salt in general terms, manpower
required, and precautions.
Point

Value

(1)

(6)

(4)

(11)

GO

90

Radiation Safety and Control

1,

2,

Explain what is meant by RBE (Relative Biological Effective-
ness) and how this is used in determining an exposure dose.

Give your reaction to and the basis for your reaction to each
of the following:

a. Using a G-M survey meter to measure the dose rate near
a vy source of unknown, but possible large, strength,.

b. Using a G-M survey meter without using the earphones.

c. Making a contamination surxrvey of personnel leaving a
contamination zone in a direct radiation field of
10 mr/hr.

d. Using a cutie pie to make a contamination survey.

e. Using a cutie pie to measure the dose rate in a y field
for the purpose of calculating working time.

f. Calculating the working time for a job at one of the
beam holes on the basis of dose rates determined using
a cutie pie.

An operator is standing 15 feet from a radiocactive point
source, He reads 1000 mrem/hr on a portable instrument., If
he must operate a valve located 5 feet from the source, how
long can he linger at the valve before obtaining the ORNL
quarterly whole body radiation exposure limit? Whose approval
is required for the job? Explain.
Point
Value

Hfl

(5)

(5)

(5)

(6)

 

91

Reactor Theory

1.

A reactor has a count rate of 30 cps when it has a K
eff
of 0,95,

a, Is 40 cps the proper count rate when Keff is 0.9757
Show how you arrived at your answer.

b, If the regulating rod position was 16 and 18 inches when

Keff = 0.94 and 0.975, what position would you estimate

for criticality? Show your calculations.

Discuss the following reactivity coefficients as to + or -
as applied to MSRE.

a. Reactor outlet temperature
b. Power
c. Uranium
d. Bubblers
e. ZXenon
Diagr?m the processes and regions involved in the production
of '3%Xe in the MSRE, its transport and its eventual fate.
What percentage of the *23U in the fuel loop is consumed in
a full-power day? %, How much would the regulating
rod have to be withdrawn to compensate for this effect?
in. (Use the following data and show yvour calculations.

1l g - mole - 6.02 x 10 atoms 1 day = 86,400 sec.
1 fission yields 197 Mev 1 j =1 watt-sec = 6,25 x 10**Mev
Full power = 8 MW ce/of = 0.11 for *3®°y

94% of all fissions in MSRE are now in 233U

There is 29 kg of *°?U in the fuel loop.

Q

233y concentration coefficient, (Ak/k)/(Ac/c) = 0.37
Regulating rod sensitivity = 0.07% 8k/k per inch.
Point
Value

(4)
(3)

(2)
(5)

(5)

(19)

92

Radiocactive materials Handling, Disposal, and Hazards

 

1.

2-

A reading is taken and found to be 2000 mr/hr; 6 hours
later 300 mr/hr. What is the half-life of the material?

What are the requirements for entry, work in, and exit from
a contamination zone? Radiation zone?

What is the difference between radiation and contamination?

A small fission-product source reads 80 mr/hr at 3 ft through
1.5 in. of Pb shielding. If the source is taken out of the
shield at what distance will the dose rate be 200 mr/hr?

(Use a tenth layer value for Pb of 1.9 in. and show your
work.,)

Assume that you are Shift Supervisor and the CAM and monitor
in the southeast corner of the highbay alarm. If you cannot
wait for a health physicist and you feel that you must remain
in the control room, what instructions and precautions would
you give your operators in determining the source of the
difficulty. (Assume that they are not familiar with handling
of activity.) How can they tell whether it is airborne
activity?
Point
Value

(5)

(5)

(3)

(5)

(5)

(23)

93

Specific Operating Characteristics

l.

Describe the modes and sub~modes of operation at the MSRE,
What operations are done in each and what are some of the
requirements to be in each?

Using the attached FP level curves (Fig. J-2) show calcu-
lations for filling reactor if we wish to operate at 55 -

60% at 1210°F and 60 cycles. Assume that average temperature
is 1180°F and the maximum operating level of salt is in the
OFT.

After filling the reactor there is a 3-hour delay before

starting the fuel pump. Explain the reason for this.

How can the nuclear power be changed without the control rods
moving? 1In such a case which remains more nearly constant:
the core outlet temperature or the average temperature of the
fuel in the core?

What process parameters are used in causing an instrumented
control rod scram (safety)? Explain the purpose of each of
these,
94

 

Figure J-2.
Point
Value

(3)

(3)

(2)
(3)

(11)

95

 

Describe the tests made and conditions which should exist

What is the wvalue of the overall temperature coefficient
of reactivity? If an enrighing capsule of ?2°U is added and
the control rods are not moved, how much will the reactor

Describe what happens to the reactivity immediately after

Fuel Handling and Core Parameters
1.

prior to filling the system with fuel salt.
2,

temperatures shift?
3.

the fuel pump is started and explain why.
4,

Describe the various neutron sources at the MSRE (external
and inherent).
Point

Value

(4)

(3)

(4)

(4)

(4)

(4)

(3)

(26)

96

Administration, Conditions, and Limitations

1.

Why are up~to-date procedures necessary to the efficient
operation of the MSRE? Who is responsible for the maintain-
ing of up-to-date procedures? How is this accomplished?

Describe the procedure necessary to make a major change in
the reactor. (For instance, bypassing the flow restrictor
in the supply line to the overflow tank.)

Describe the minimum personnel and classification of personnel
required during various stages of operation.

a., In the control room
b. At the reactor site

Describe the consequences of complete loss of both treated-~
water pumps while the reactor is at full power. Assume that
flow of treated water cannot be restarted for an extended
period. What operation action should be taken and why?

Rod scram is caused by safety circuits in a 2-out-of-3
matrix. Therefore, loss of one safety channel will not
scram the rods, If we are making an extended full-power
run and during 8D check 1list it is found that a high flux
signal will not initiate a scram, explain what you would do
and justify your actionms.

What are the Operating Safety Limits? In what ways are they
treated differently from the Operating Procedures?

Describe the reason for having jumpers available in the
jumper board? What are the two types? What approval is
necessary to use each type?
97

APPENDIX J

Examples of Instructions and Communication
99

SHIFT INSTRUCTTONS

December 13, 1968

GENERAL TINFORMATION

I

It seems that whatever is causing bubbles in the fuel system remains
with us. Note that WR~-2947 has raised OFT level alarm and action set-
points to 30 and 35%. We will operate the FP between 50 and 59% level
on the No. 1 bubbler (switch position No. 2). These new limits are
marked in the control room log. The lowering of FP level is for two
reasons — the most important being to keep the foam away from the off-
gas connection. The other being to lengthen the time interval between
OFT burps.

The most important disadvantage of the above changes comes when
we have a FP stoppage for some reason. The resulting FP level will be
much lower than necessary to restart the pump. The OFT must then be
partially emptied to get the FP level to ¢64%. With the FP on and
&8% worth of bubbles circulating, the FP level is out of limits (i.e.,
>59%). Care must be taken to avoid reheating the system too rapidly
and causing level to increase even more. The recovery could take

several shifts. See instructions in Permanent Shift Instructiomns.

SHIFT INSTRUCTIONS

I

IT

Continue with operation at 1 kW and 1210 = 2°F OAFOT until such time
the computer is repaired and we can go to 1 MW, Take temperature bias

data per Item IIT of 12/12 if not already done.

When computer is repaired (today or Monday) we should be ready to go

to 1 MW heat-balance power on temperature servo. This will be with no
blower and a jumper is approved for 139f to get into run mode. See
Item ILI for sample to take 1 hour after going to 1 MW, Dynamics tests
will be done while at 1 MW.
100

Shift Instructions 2 December 13, 1968

11T

IV

Isolate a FV salt sample one hour after going to 1-MW power. Isolate

per 6A3 except as follows:

(1) In Step 3.3.2.1, leave in salt for 10 minutes., Do not hesitate
on insertion.

(2) Add between 3.2.3.1 — 3.2.3.2. Hold 5 minutes at v6 ft - 5 in,
to allow for drippage.

Deliver to lab "special FV salt sample for Kirslis.," Hang a 10-gm

capsule on the latch,

To prevent the possibility of the FV capsule now on the latch from

losing its vacuum, keep Area 1C evacuated until ready to ‘use.

Please review Daily Report Sheets from computer 8-hr log (printed at
0745) and tabulate thermal cycles (heat, fill, and power cycle) on FP,
CP, FF-102, and FF-200. Daily from 0745 11/11/68 to 0800, 12/13/68.

Computer sheets are either in Watt's office or are filed downstairs,

J. L. Crowley
101

PERMANENT SHIFT INSTRUCTIONS

December 13, 1968

In the event an FP stoppage (and scram if at power) requires transfer

of salt from the OFT to restart, follow the general plan below.

1,
2,

Transfer only enough salt from the OFT to restart the FP per 9I.

While awaiting transfer, begin turning up system heaters (if load and
rod scram has occurred). Bring up to normal temperature.

Restart the FP and bring the reactor critical to 1 kW per 5J.

Add nuclear heat up to 100 kW as necessary along with electrical heat
to maintain FP level at 65% until system temperature reaches 1210°F,
This may take several shifts. At this time you can return to power or

whatever condition called for in shift instructions.
102

Approved by fiz:'3¢44“7 5J-1
W, 9/1k /65

2/2/68
7/2/69
5J CRITICALITY AND POWER OPERATION

The fuel and coolant salt will be circulated subcritically in the
loops until power operation is desired, at which time the control rods
will be withdrawn to obtain criticality.

During normal power operation, manual load control and servo rod
control will be used. However, manual rod control may be used if neces-
sary. Automatic load control will be used only for special tests for
which detailed instructions will be supplied.

The first section is written for use as a "Quickie" check list for
recovery of temperatures after a rod and load scram. After recovery is
made, the reactor will be taken subcritical and the normal pre-power
operation checks of Section 2 completed.

Any time the power is reduced below ~ 200 kw, the entire preparation
(Section 2) check should be completed.

This procedure is corganized as follows,

1. Quick recovery of temperature check list.
2. Preparation for power operation.

3. Starting power operaticn using rod servo,
L

. Starting power operation using manual rcd control.

1. QUICK TEMPERATURE RECOVERY CHECK LIST
This check list is to aid in recovering system temperature

using nuclear power following a rcod and load scram.

Tnit. Date/Time
1.1 If necessary, cocmplete Electrical Power Outage

Quick Check List — QAL,

 

1.2 Set FT heater variacs to 8% for low-power

operation.

 

1.3 Clear safety channels and ungallop safeties,

 

1.4 Check or put fission chambers in automatic.

 

1.5 Switech rod servo off.
1.6 Set flux demand at L0 - 60% of scale and select
1.5 x 10° range (150 kw).

 

 
103

Approved byyé‘_’w 2/&% 1AL / Ziég
9/14/65

NOTE «

NOTE -

2/2/68
7/2/69
Init. Date/Time

 

1.7 Group withdraw to ~ 30 in., Rod-3 lower-limit
light may not clear.

 

1.8 Switeh rod servo on. Reset the flux scram set-
points to Low Sensitivity (Red) range.

1.9 Insert No. 1 rod to 20 in.

 

 

1.10 Group withdraw rods until a 30-sec period is
obtained or until No. 2 and No. 3 are at
~ in. (See Control Room Log.) If 30-sec

periocd occurs, finish withdrawing No. 2 and

No. 3 individually.

 

Reactor will become critical below normal rod po-
sitions because of low-system temperature.

1.11 Withdraw No. 1 rod until critical and servo is
controlling rod near middle of regulating rod
limit switches.

1.12 Note whether Rod No. 3 lower limit cleared.

Yes

No ___, o

 

 

1.13 If Rod No. 3 lower-limit light did not clear,
raise Rod No. 3 to clear low-limit light and
then adjust Rod No. 3 to Rod No. 2 position.

Keep regulating rod on scale during this step.

 

If temperature 1s not increasing at a reasonable
rate, increase flux demand to 40% of .5 x 10° scale.

1.14 When 1175°F temperature (OAFOT) is reached, re-
duce power to 10 kw (66% of 1.5 x 10% range).

 

1.15 Reset load scram and use radiator doors to pre-
vent overheating (above 1210°F) because of

fission product decay.

 

1,16 Review 5J-2. As soon as convenient, take reactor
subcritical, run rod drop test, fiducial zero of
rods and any other tests or check lists which have

not been done within specified time limites.

 

1.17 Then return to power operation using 5J-3 or

5J-4 procedure.

 
104

. Al
Approved by fi' %@7’“”‘ 5J-3
r 9/1k /65
2/2/68
7/2/69

 

2 PREPARATION FOR POWER OPERATION
Prior to taking the reactor critical, the system should be
checked to assure that all pertinent equipment and instrumentation
are functioning properly. The more important items to be done are
listed below. Additional tests may be necessary depending upon
conditions. At times, such as during experiments, after short
periods of subcritical operation, etc., it may be desirable to omit

all tests, This is at the discretion of the shift supervisor,

Init, Date/Time
2.1 Check rod drop time for each rod to be less than
1l second. This can be omitted if it has been
done within approximately one month of this date.
2.1.1 Check that 2 rods are between 1 and
3 inches.
2.1.2 Raise other rod to 50 inches above the

rod position where the first lower limit

 

indicator light lights up.

 

2,1,3 Plug in the rod drop timer and set to

ZEero,

 

2.1.4 Actuate the rod scram switch and check
that all three rods drop.

2.1.5 Repeat with other two rods and record

 

results in table below.

 

 

 

 

Rod Drop Time Note That
Starting |Should Not Exceed | A1l Three Rods
Rod #{ Position 1.0 Sec, Dropped Tnitial | Date
1
2

 

 

 

 

 

 

 

 

 
 

Approved by A%/ @/M"i

Record rod position when maximum dp is obtained.

105

5J~4
9/1k /65

2/2/68
T7/2/69

Init.

Date/Time

 

2.2 Check fiducial zero of each rod if rods have
been scrammed since last FIDO determination.
Otherwise this can be omitted if it has been
done within approximately one month of this
date.

2.,2.1 Check that cell-air activity is not high:
Record RM-565B __ and RM-~565C
Also check radiation level in TR: Record
CAM __, and Monitron . Do not
proceed if cell-air activity is >5 mr/hr,

 

2.2.2 5Set valves as follows:
HV-985-A2 open
HV-98T7-A2 open
HV-987-A3 closed
HV-986A open

an

 

 

2.2.3 Open HV-986A, 987A, or 988A. Throttle
HV-985-A1 and HV-989A until ZI-98TA indicates
approximately 50%.

 

2.2.4 Establish communications between TR

and Control Room.

 

2.2.5 Insert control rod and determine control-

rod reference position.

 

indications for each rod being inserted and for each being withdrawn.

Determine position

 

CONTROL ROD POSITION

 

 

 

 

 

 

 

 

 

 

Valve Valves Actual
Rod #| Open Closed Should be Inserting [Withdrawing |Init.| Date
1 | 986A | 987-Al] 988-A ]1.2 to 1.6
2 | 987-Al] 986-A | 988-A j1.4 to 1.8
3 1 988-A | 986-A | 987-A1|2.3 t02.65

 

 

 

 

 
106

fi@f?é
Approved by ' AN ) ijé5
9/14/65

7

Tnit. Date/Time

 

2.2.6 Open HV-987-A3.
2.2.7 Close HV-985-A1, 989A, 986A, 987-A1,g884A.

 

 

2.3 See “J-5A,
2.1 Check that two out of three safety channels will

 

scram all three rods. This can be omitted if it

has been done within approximately one week of

this date,

2.4.1 Inhibit fast scan.

2.4.2 With all rods above 1 inch, trip safety
channels 1 and 2 by pushing test buttons on
RSSNSC1AL and RSSNSC2AL and note that rods

 

 

 

scram.,
2.4,3 Reset Safety Channels 1 and 2.
2.L.4 Raise all rods to 1 to 3".

 

N

. 4,5 Trip Safety Channels 2 and 3 by pushing
test buttons on RSSNSC2AL and RSSNSC3AL4 and
note that all three rods scram.

2,46 Reset Safety Channels 2 and 3.

2.4.7 Raise all three rods 1 to 3 inches.

 

 

 

2,48 Trip Safety Channels 1 and 3 by pushing
test buttons on RSSNSC1AL and RSSNSC3AL and

rote that all three rods scram.

 

2.49 Reset Safety Channels 1 and 3 and re-
establish desired positions.

2.4%10 Cancel "Inhibit fast scan".

 

 

2.5 Check that thermocouples on the radiator outlet
tubes are plugged into Scanner D and E and that
the gain on both scanners is set at 150 so that
an alarm will occur at QOC°F or 1300°F. This
can be omitted if no heat is to be removed by

the radiator.

 
107

Approved by»u"ggy t{- ,flkhftz 5J-6
! ? 9/1k/65

2.6

2.7

2/2/68

7/2/69
Tnit, Date/Time

 

Complete the Neutron Instruments Check List,
(8A). This can be omitted if it has been done
within approximately one month of this date.

 

Complete the Calibration Check of Process
Radiation Monitors (8B). This can be omitted
if it has been done within approximately one
week of this date,

Complete the entire Safety Circuits Check List
(8D). This can be omitted if it has been done

 

within approximately one month of this date.

 

2.9 Take TE bias data if requested by Gabbard.
2.10 Take zero-power heat balance if not done within

approximately 1 week,

 

2.11 Check that annunciators are clear or approved
by the shift supervisor. ©Shift Supervisor's

approval

 

2.12 Check that entrance to all exclusion areas are

properly restricted and radiation signs posted.
2.12.1 Door No. 1 to radiator near north
annulus blower is locked.

2.12.2 Door No. 2 at foot of ramp to CDT cell

 

is locked.
2.12,3 Door No. 3 to blowers is locked.

 

 

2.12,4 Blocks are on reactor cell, drain-tank
cell, coolant cell, south electric service

area, and special equipment room.

 

3 STARTING POWER OPERATION USING ROD SERVO

The shim rods will be withdrawn manually while the servc con-

troller manipulates the regulating rod to attain criticality and controls

the power at the setpoint.

To increase the nuclear power, the flux demand will be increased.

This may cause the regulating rod to withdraw until the regulating-rod

1limit is reached. The limit can be changed by operating the regulating-rod
108

 
 
 
 

T

Approved by~ M 5J-=T

9/1L /65
2/2/68
7/2/69

3 (continued)
drive switch or the shim rods can be manmually withdrawn which will
cause the regulating rod to insert.

As the flux demand is increased, the nuclear power may cause the
system temperatures to rise necessitating removal of heat at the
radiator., Number one blower will be started and the resultant AP
across the radiator will cause the bypass damper to open if damper
controcl is on automatic.

Before pushing the run button, the power should be between 0.5

and 1 Mw; both range selectors should be in the 1.5-Mw range, and
the temperature demand setpoint should be slightly higher than the
outlet temperature. The regulating rod should be in the center
portion of its useful range and not at either the insert or withdraw
1imit. When the reactor is switched to the "run" mode, the range
selector will be sealed in the 15-Mw range, and the rod-control
circuitry will be changed frum flux servo to temperature servo.
Under these conditions, the regulating rod will be automatically in-
serted or withdrawn to maintain the reactor-outlet tempersture con-
stant. The temperature can be changed by adjusting the temperature-
demand setpoint.

Operation at full power will be reached with both doors open,
both blowers in operation and the bypass damper closed. If the regu-
lating rod reaches the withdraw limit, No., 2 and No. 3 rod will have
to be manually withdrawn or the limit will have to be raised. When
critical, Rod 2 and 2 should always be withdrawn the same amount
(within % inch of each other). Rod 1 should be kept at least L inches
below Rod 2 and 3 and within the range of 8 inches and 39 inches
withdrawn,

3.1 To take the reactor critical on servo rod control, prcceed as

follows:
109

Approved bYWY/V M 5J~8
o v

9/1L/65
2/2/68
7/2/69

Tnit. Date/Time

3.1.1 Put rod control on servo using servo-mode
selector switch (S-16). Note that lights
on console indicate that the flux servo con-

trol is on.

 

3.1.2 Set servo flux channel selector switch
(S-17) to No. 1 channel.

3.1.3 Set flux demand on selected channel to

 

10 kw using range-selector switch (RXNARC-A5),
and flux-demand knob (RXNARC-A6).

3,1.4 Set other channel range-selector switch

 

at lowest practical range.

 

3.1.5 Check that regulating rod is at the upper
regulating-rod limit. Regulating rod should

be at 12 inches or less.

 

3.1.6 Set fission-chamber selector switch
(8-15) as desired. DNo. 1 ,

No. 2 , Both

 

3,1.7 Set fission-chamber No. 1 mode selector
switeh (S-13) to automatic (pushed in).
3,1.8 Set fission-chamber No. 2 mode selector
switch (S-14) to automatic (pushed in).
3,1.9 Withdraw the shim rod No. 2 to ____ in.
Then withdraw shim rod No. 3 to _____ in.

 

 

 

3.1.10 Reset the flux scram setpoints to low
sensitivity (RED) range.

 

3.1.11 Switch regulating-rod actuator switch
(8-19) to withdraw. This will raise the
regulating-rod limit switch allowing the
flux servo to withdraw the regulating rod.
Continue regulating~rod withdrawal until
criticality is attained and desired flux is
reached., When critical the regulating rod
should be within the range of 8 to 39 inches

and should be at least 4" below the shim rods.

 

*
0

11
Approved by é_;_ zj\_fé e AN 5J-9
9/1L /65

> /2 /68
;757%9

NOTE: As flux increases, change the range on the alternate picoammeter
as required and observe the linear flux indicator. Also observe the
period meters on the console,

The withdrawal of the regulating rod can be made as fast as the
25-sec period control rod withdraw inhibit willi permit,.

When withdrawing the rods, criticality for operating purposes is de-
fined as being attained when the countrate is at least 500 times the
countrate prior to withdrawing the rods (all rods full inserted) and
RR-8100 indicates 1 kw or greater.

When taking the reactor critical after an unintentional shutdown,
unless special instructions are given, attain criticality as described
here except that RR-8100 shall indicate a power level not less than
1.1 kw for the critical rod positions of Step 3.1.12,

Init. Date/Time
3.1.12 When the flux reaches the flux demand
setpoint, the servo will level the power at
the setpoint. The regulating rod actuator
switch 5-19 should be released when the
regulating rod is near the center of the

regulating rod limit switches.

 

3.1.13 While maintaining the reactor at this
power, record rod positions (Rod No. 1
Rod No. 2 _ , Rod No. 3 ) and
CAFOT ___ ., Check the reactivity balance

ermrraremre?

% 8k/k. Net reactivity is normally
less than + 0,2%, Shift Supervisor's per-

mission to increase power .

 

3.2 To raise the power in rod servo and manual load

 

control

NOTE: The transition from start mode to run mode (i.e.
from 10 kw to 1 Mw) should be made guickly to avoid
over-heating the system. Review the remainder of
this portion of the check list so that the four
most important steps can te done with minimum delay,
These are Steps 3.2.7, 3.2.8, 3.2.9, and 3.2,10

and .
111

Approved by %49??;z2;9229/5uéfiA 5J-10
V 9/14 /65
2/2/68
7/2/69

Tnit. Date/Time

 

3.2.1 Check that fission chambers No. 1 and 2
are in automatic.

3.2.2 Start anmulus blowers (MB-2 and MB-k4)

3.2.3 Check that the bypass damper is set on

 

 

automatic and the bypass damper is fully

open,
3.2.4 Using $-18, adjust the temperature demand

 

setpoint slightly more than the indicated

outlet temperature.

 

3.2.5 Adjust the regulating rod limit switch so
that the regulating rod is about center of
its range.

3.2.6 Check that the blower damper switches,

 

on blowers which you expect to operate, are

set to "run" position (MB-5).

 

3.2.7 Increase nuclear power to about TOO kw by
increasing both picoammeter range selector
switches to their maximum range position and
setting flux demand knob to about L5%.

3.2.8 Start MB-1.

3.2.9 Push run button (S-11). This seals the

 

 

reactor in run mode and you should be in
temperature servo. Check the console to

confirm this.

 

3.2.10 Raise radiator doors. Begin by raising
outlet door half way, inlet door to its
minimum setting, outlet door to upper limit,

and the inlet door to upper limit.

 

3.2.11 Raise the heaters on coclant system flcw

transmitters to their power operation setting.

 

3.2.12 Adjust the temperature demand Lo give the
desired ocutlet temperature (normally

1210 + 2°F),

 
112

TSt
Approved b)/f; /5}“/1’// 5J-11
y’ 9/1k /65

2/2/68
7/2/69
Init. Date/Time

 

3.2.13 Close the bypass damper by raising the
radiator AP demand with the damper on auto-

matic control.

 

NOTE: So that the bypass will reopen upon starting of
2nd blower, raise the AP demand only enough to close
bypass damper — no more.

3.2.14 When the bypass damper is closed, start
the second main blower (MB-3). The damper

should open automatically.

 

3.2.15 Continue raising the radiator AP demand
until the bypass damper is completely closed

or until desired power is reached.

 

3.2.16 Take a reactivity balance, % sk/k,

 

3.2.17 When steady power 1s attained, set the
linear power switches to alarm if power varies
by = 5% of scale.

3.2.18 Check that the treated water surge tank

 

and degassing tank are being purged with air.

 

4 STARTING POWER OPERATION USING MANUAL ROD CONTROL

The reactor heat lcad and flux can be adjusted manually in a
number of different ways. The control rods can be manipulated
manually to attain criticality and adjust power using the individual
actuator switches. The regulating-rod limits have no function when
in marnual control. Group insertion is possible at all times and
group withdrawal can be done when in the start mode. When critical,
Rods 2 and 3 shculd always be withdrawn the same amount (within
% in, of each other). Rcd 1 should be kept at least 4 inches below
Rod 2 ana Rod 7 and within the rarge of 8 inches to 39 inches with-
drawn.

At very low powers, 1t may be necessary to adjust power removal
by adjusting the electrical heaters or it may be advantageocus to

raise one or bceth radiator doors with the blowers off. At higher
Approved by

113

5J-12
9/1k /65
2/2/68
7/2/69

power with one or both of the blowers on, fine adjustment may be

made by setting the doors at a fixed position and changing the

bypass-damper position. The damper control may be set on automatic

which will hold a constant radiator AP or on manual which will

maintain

a fixed damper position.
Init. Date/Time

 

4,1 To take the reactor critical manually, proceed

as follows:

4.1,

L.1.

L.1,

h.1.

L.1,

L.1.

L.1,

h.1.

L.1.

L.1.

1 Put rod control on manual using servo-
mode selector switch (5-16). DNote that
lights on console indicate that the flux

servo control is off.

 

2 Set servo flux channel selector switch
(S-17) to No. 1 channel.

3 Set both linear channel range-selector

 

switches at lowest practical range.

 

 

 

4 Check that regulating rod is at 12 in.
or less.

5 Set fission-chamber selector switch
(8-15) as desired. No. 1 )

No. 2 , Both

6 Set fission-chamber No. 1 mode selector

switch (8-13) to automatic (pushed in).

 

7 Set figsion-chamber No. 2 mode selector
switch (S-14) to automatic (pushed in).
8 Withdraw the shim rod No. 2 to in.

 

Then withdraw shim rod No. 3 to in.

 

9 Check that flux scram setpoints are to

low sensitivity (RED) range.

 

10 Switch regulating-rod actuator switch
(5-19) to withdraw. This will raise the
regulating-rod. Continue regulating-rod
withdrawal until criticality is attained and

desired flux is reached. When critical, the
114

 

T
Approved b¥4fi>7/{4%<vflfléf;¢//7l
) r

NOTE :

NOTE:

L.,2 To raise the power with manual rod control and

Init.

oJ-13
9/1L /65
2/2/68
T/2/69
Date/Time

 

4L,1,10 (continued)
regulating rod should be within the range of
8 to 39 in. and should be at least 4 in.

below the shim rods.

 

As flux increases, change the range on both the se-
lected and the alternate picoammeter as required and
observe the linear-flux indicator. Also observe the
period meters on the console,

4.1.11 TLevel the power at 10 kw using the
regulating rod.

 

4,1.12 While maintaining the reactor at this

power, record rod positions (Rod No. 1 |,
Rod No. 2 __, and Rod No. 3 ) and
OAFQT . Check the reactivity
balance % sk/k. Net reactivity is

normally less than * 0.2%. Shift super-

visor's permission to increase power

 

 

manual load control

4L.,2.1 Check that Fission Chambers No. 1 and

 

lNo. 2 are in automatic.

 

=
N

.2 Start annulus blowers (MB-2 and MB-4).

 

n

.3 Check that the bypass damper is set on

automatic and the bypass damper is fully

cpen,

 

=
no

.4 Check that the blower damper switches, on
blowers which you expect to operate, are set

to "run" position (MB-5).

 

The transition from start to run mode (i.e. from
10 kw to 1 Mw) should be doune carefully to avoid
over-heating the system. Review the remainder of
the check list so that the necessary action may be
taken with a minimum of delay.
115

'/
Approved b¥¢gfiézgiv4;;“%4fifl 5J-14
v

L.2,
L.2,

L.2,

L.2,

L.2,

L.2,

9/1L /65
2/2/68
7/2/69

Tnit. Date/Time

5 ©Start MB-1.
6 Adjust regulating rod to increase the

 

nuclear power. Increase the range on both
picoammeters as necessary to keep linear
power on scale, Continue to increase nuclear
power to about L0 to 50% of 1.5 Mw (maximum
range position of picocammeter selector

switches).

 

T Lift outlet door to minimum setting if
necessary at this time to maintain reactor
outlet temperature.

8 Push run mode button. Check that run

 

mode seals in (see jumper board).

 

9 Continue to increase both load and
nuclear power while maintaining reactor out-
let temperature fairly steady. Raise outlet
door half way, inlet to its minimum setting,
outlet door to upper limit, and then the

inlet door to upper limit.

 

.10 Raise the heaters on the coolant system

flow transmitters tc thelr power setting.

 

11 Close the bypass damper by raising the
radiator AP demand with the damper on auto=-

matic control.

NOTE: ©So that the bypass will re-open upon starting of

the second blower, raise the AP demand only enough

to close bypass damper -— no more.
u.gi

 

N

12 When the bypass damper is closed, start
the second blower (MB-3). The damper should
open automatically. Increase nuclear pcwer
as necessary to maintain reactor outlet

temperature,

 
116

 

Approved by/ G M
v

NOTE .

Init.

5J-15
9/1k4 /65
2/2/68
7/2/69
Date/Time

 

Very little adjustment of the regulating rod should
be necessary since you will get some increase in
nuclear power due to lowering of the average
temperature (negative temperature coefficient)
when more load is being removed.
4L.2,13 To reach maximum power, continue to
add load to the system by increasing the
radiator AP demand until the bypass damper

is closed.

 

L.2.14 Take a reactivity balance % dk/k.

 

L.2.15 When steady power is attained, set the
linear power switches to alarm if power

varies by * 5% of scale.

 

4,2.16 Check that the treated water surge tank

and degassing tank are being purged with air.

 
117

MSRE DYNAMIC TESTS
MSRE Test Memo 3.2.4.1

Transient Flowrate Measurements

Submitteflwfitwmwved Progeltirfos
Submittedd /. &,A ;bszflflpproved

Approved for use -

 

Introduction
Measurements will be made to determine the fuel and coolant salt flow
rates during startup and coastdown of the circulating pumps. The transient
behavior of the fuel void fractions as indicated by the gamma-ray densito-
meter will also be measured.
Purpose
Information on the transient salt flow rates and void fractions during
pump startup and coastdown will be used to interpret results of the tests
of flow on reactivity (see MSRE Test Memo 3.2.4.5).
Description
These tests will be run with the reactor subcritical and essentially
at thermal equilibrium. For the coolant loop tests, salt flow rate and
pump speed will be monitored, and for the primary loop tests, pump speed
and fuel density (from the gamma-ray densitometer) will be monitored.
Special Equipment and Preparations
Since the pump startup time is quite short, a multichannel oscillograph
recorder will be used to monitor the outputs of the coolant salt venturi
differential pressure cells directly, and both primary and coolant pump speed.
The data logger will be used in the continuous fast scan mode to also record
coolant flow rate, pump speeds, and fuel density, as well as certalin loop
temperatures for reference.
Procedures
1. Set input power to loop heaters so that a constant temperature
(nominally 1200°F) is maintained with both circulating pumps
running.
2. Jumper the control interlock so that the fuel pump may operate with
the coolant pump off.
118

Put the logger in the continuous fast scan mode, and monitor
coolant flow and pump speed on the oscillograph.

Shut off the cooclant pump.

After the flow transient has died out, restart the coolant pump.
Monitor primary pump speed on the oscillograph (also densitometer
signal if possible). Put logger in fast scan mode.

Shut off the fuel pump.

Restart fuel pump after flow trangient (i.e. pump speed and
densitometer signals) has settled out.

Repeat 7 and 8 after waiting various intervals between starts

to see if there is any burping effect detected on the densitometer.
119

MSRE Test Memo 3.2.k4,1
Apveudix I

Check List for Transient Flowrate Measurements

Submitted flgw 6165 Approved 4 N @g ©/22/65

Submitte K;_,é,, i [J Zh Approved

Approved for use

 

This check list is to be used in connection with the test procedures

described in MSRE Test Memo 3.2.k4.1,

Initial
Provide a control interlock jumper to permit
operation of the fuel circulating pump

with the coolant circuleting pump off.

Dete/time

 

(Connect jumper wire in circuit 147 from
terminal RILE1Q to terminal R18A7 in control
relay cabinet )

Permission to insert jumper,

Connect the densitometer output (1 sec.

T/C filter) to the data logger fast

scan sequence input 57 (replacing LT-599(OFT)).

 

Connect outputs of the coolant salt venturis'
d/p cells and cooclant circulated pump speed

transmitters to a multi-channel oscillograph.

 

Set input power to loop heaters so that a
constant tempersture (nominally 1200°F)
is maintained with both circulating pumps

running. (reactor sub-critical)

 

Put the data logger in the continuous

fast scan mode and start oscillograph.

 

Turn off the coolant pump.

 

After the flow transient has died out,

restart the coolant pump.

 

After flow transient has died out, take

logger out of fast scan.

 
9.

10,

11.

13.

1k,

15.

120

Initial Date/time

Connect outputs of the fuel circulating
pump speed transmitter and the fuel salt
densitometer to the multi~channel

oscillograph.

 

Put the date logger in the continuous

fast scan mode.

 

Turn off the fuel pump.

 

Re-start the fuel pump after the
transient has died out.

 

After flow transient has died out, take

logger out of fast scan.

 

Repeat steps 11 to 13 several times

at the discretion of the experimenters.

 

 

 

 

 

Remove Jjumper of step 1.

 
121

APPENDIX K

Examples of the

Control Room Log and the Building Log
 

123

" s i

Approved Ly iz A\, \f/.g/_f;/f:f{”\, 12A-2A
| 10/3/66
2/15/67
5/23/67
8/22/67
12-2A 2/7/68
CONTROL ROOM I0G L2k /68
(Computer in Operation) 12/26/68

COVER SHEET

In the spaces provided, list any variable not in limits or which
you judge needs attention.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ITEMS
Page No. 8 -4k h - 12 12 - 8
MCR
ACR

 

 

 

 

 

 

 

 

 

- -

 

 

 

 

 

 

 

 

 

, Supplement
Date Started Cover
124

 

 

 

 

 

 

     
          

     

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

r'y S
Approve! by M}/V/“Yfim 12A-2A
Y 8/6/65
1/17/66
3/7/66
6/20/66
10/3/66
- 2/15/67
12A-2A 5/23/67
CONTROL ROOM LOG 8422/6g
/7/6
(Computer in Operation) ), 700,
At
Log taken by:
8 -4 b - 12 12 - 8
CONSOLE
Record every U hours starting at 0830 unless otherwise indicated.
1
By-Pass
RADIATOR DOOR Duct
POSITION _ RADIATOR Damper Computer
OUTLET INIET AP Demand | Position Room Temp
2I-0D~A | ZI-ID-A PAIADRALl | 2I-AD-2 (East Door)
o ¥
5
T
et ey e s ses ! e —

*
If temperature exceeds 75°F due to icing, turn normal unit up to 80°F
and standby unit down to 70°F for several hours,

Date Started
125

L ’
Approved byfi;ZZ:ZZféf%fifizfl“"‘-w

12A-2A
8/6/65

L /2k /68
1/17/66 12/26/68

 

 

 

 

 

 

 

 

 

 

3/1/66
6/20/66
10/3/66
2/1
#5e]
Record every L hourc starting at 0830 unless otherwise indicated. 825$;6g
Reactivity
Balarice
When REACTOR OUTLET Particle
QACOT Critical Temperature Temperature Trap
(Pt. 37) (Pt 12) OAFOT* Demand TE-PT1-3
Time * * g ¥k TI 100- | (Pt 37) | XTIMARC A2 | (S-1Lb)
+0.,25 *¥¥ 1210°F
=32
llSO la-lo tO _0.25% i 2°F
XX
XX
XX

 

 

 

 

 

 

 

 

 

= ,
If computer is not functioning, calculate the official average fuel, coolant

temperature, and reactivity per Computer-Out Log 12-2B.

%
Record KNET every 4 hours.

FH

COMP. ION

Req. Dem}{ Reg. Dem

(Not necessary if

system is drained)

Req. Dem

Req. Dem

Seal |Functionilog.FS.1 Pt.|log Type II|log Alarms
Inhibit

t Agtive

XX

L

¥*

Inhibit

Stopped
l

*E

 

XX

*
This is fcr purpose of having record available in case computer needs

restarting.

*¥%
Review this ard reactivate all pcssible.

If >10, see Cookbook.

Date Started

 

Advise SS of those remaining.

Request
I/O
on

Line
-

Approved by 72

126

< B/6/65 /21 ¢
8/6/65 L /ok /68
1/17/66 lé>26/6€

 

Time

 

3/7/66
6/20/66
10/3?26
2/15/67
5/23/67
Record every U hours starting at 0830 unless otherwise indicated.8gfi%%g§
Flux
Demand
(Computed Servo
In Run Regulating | Servo Flux
Manual Rod Mode Channel
In Start! Position Light | Selector ==__#
25 - 75% #1 or #2

]

=============a=======T==========4===========================*

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

CONTROL ROD Load Safety L
POSTTTON* Scram Channel Exercise
Total Lights Lights BF 5 *¥ ission
Time #1 #2 #3 On On Chamber Chambers
2TVE6T] 36, | Hy |
woex | Anfl fn7| L &20n None CPS #1 |#2
24
XX xx
XX XX
XX XX
XX XX
i_ lxx=£====

 

 

 

 

 

 

 

 

 

 

Read coarse position indicator to nearest printed number below pointer. Add
fine reading to this and record the sum. Keep all three at 24 inches when
sub-critical.
*%¥BFx chamber will be withdrawn while circulating fuel salt.
¥¥Once per day exercise both fission chembers by selecting alternate, manu-
ally withdrawing (or inserting) several inches and returning to original
position.
*¥¥%At steady state and CAFOT 1210°F. If not at 1210°F or if FP is off

calculaste new limits.

Allow 1-1/L4 in. for every 10° of OAFOT and

1-1/4 in, for FP Off-Cn.
Date Started

 
127

Approved by ”52" e )
6/9/6() ~ ‘1/‘){5

8/25 /66 000 /68
2/6/67 L/7/69
L /26 /67
12A-3 BUILDING LOG 8/18/67
In the spaces provided, list any variable not in limits or which
you judge needs attention,

Page (Area) Items

8 - &

Electrical
and
Diesel House

Water Room
and
Cooling Tower

Vent House

Stack Panel
SER, Hi Bay
852t

 

Date Started Supp.ement

 
 

 

 

 

128

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Approved by A K/ S gmern }3?é3
Y N , 10/5/65
12A-3 BUILDING 1OG 1/18/56
Taken by: 2[2u/gg
8 - L run No. | 0/9/
8/25/66
v-12 2/6/07
/26 /e
l -
2= 8 8/18 /67
SERVICE TUNNEL 2/6/68
Record every L hours starting et 0830 unless otherwise indicated.
l
OT-2 Water
Personnel Monitors ¥ Temperature
Monltron CAM FI- FI- Qut In Flow
Time | RE-7017 RE-7005 753 | 754 | TI 822-1 | TI 823-1 FI-823-A
3-5 | 6-9 o ¥ on® | . .
< h <1000 ~ 85°F ~ 80°F > 7.5 gpm
3 mr/br P gpm | gpn 0 e
;>3( ~ X3 XX i XX XX XX XX |
1 1
Wi L AX XA XX E XX XX i XX XX
t‘j l
_ \k XX X X ]xx xx | oxx o boxx XX
*Colder if process water is in use.
Coolant 0il Pump Filter AP OT-1 Weter {
Pressure PI~-752-C Temperature i
#1 #2 Minus out In Fiow |
Time | PI-T51A | PI-752A | PT-753-C | TI 820-1 | TI 821-1 | FI 821-A
OP o o -
¥ K% N ~ ~ 7,
<5 psi 85°F 8C°F > 7.5 gpm
XX XX XX XX XX XX
XX
XX XX XX XX XX XX
XX
XX XX Xx XX XX XX {

 

 

 

 

 

 

 

 

 

.X.
%*Discharge pressure from the

¥
V=754 should be wide open.

obtaln desired shield plug.flow.

rate,

Date Started

pumps which

are on should be >55 psig.

Throttle V-765 (by-pass) only as required °

 

This will permit maximum recirculation

g2 fhd
L/7/6%
129

 

 

 

 

 

 

 

 

 

 

 

 

 

Approved b /W/G/ vy A, 12A-3
’ V 10/5/65 8/12/68
1/18/66 4 /7 /69
2/2L /66
6/9/66
8/25/26
. /67
Record every 4 hours starting at 0830 unless otherwlse 1ndicated.u 2676
522k
Fuel 01l Pump Filter AP
Pressure® PI-702-C "
#1 #0 Minus FI- | FI-
Time PI 701-A |PI 702-A | PI-7G3-C 703 | TOL
-3 % <5 pSi 3‘5_5 6—9
AP
XX XX XX XX
XX
XX XX XX XX
XX
XX XX XX

 

 

 

 

 

 

 

 

 

#*
Discharge pressure from the pumps which are on should be >55 psig.

e

V-704 should be wide open. Throttle V-715 (by-pass) only as required to
obtain desired shield plug flow. This will permit maximum recirculation
rate.

E E ARFA

Record every L hours starting at 0830 unless otherwise indicated.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

F. 0il Supply Tank |C. 0il Supply Tank Process Monitof***
Reading Reading
Tlme LIOT1A3 [Set Point JLIOT2AR |Set Point | RM OT-1 RM OT-2 _
>50% Re?ding >50 % Re%ding <5 mr/hr | <5 mr/hr
2% 2%
— e e
XX XX XX XX
XX XX XX XX

 

X
Set calibration at 0.25 to 0.35 mr/hr before reading.

Date Started

 
131

APPENDIX L

Forms Used in Making Changes
 

 

133

Form H-l. MSRE PUNCH LIST

 

 

To | Priority Date

 

Location ‘ _ Requested by:

 

Equipment, ILine No.; ete,

 

 

Description of Work to be Done:

 

 

 

 

Precautions:

 

 

Estimated Cost:

Approval to proceed:

 

 

 

 

Describe'work done if differenfiffofi abovefi

 

 

 

 

Job Cdmpleted and Accepted by: =~ - Date:

 

 

 

 
 

134

 

 

MSRE WORK REQUEST

 

 

WORK ORDER NUMBER

 

WORK REQUEST NUMBER

PRIORITY DATE

 

ISSUED TO

EQUIPMENT, SYSITEM, ETC.

 

 

DESCRIPTION OF WORK

 

 

 

 

 

 

APPROVED MAINTENANGE

]A--novto OPERATIONS

 

.PRECAUTIONS

 

 

 

STARTING DATE OR TIME

CRAFTSMAN

WORK COMPLETED

"WORK INSPECTED
WORK AFPROVED

AEMARKS

RVISOR"S

ENDING DATE OR TIME

AL TO PROCEED

14C-PLE REACTOR DIVISION
SUPERVISOR MAINTENANCE

SIGNED

REACTOR DIVISION
OPERATIONS

 

 

 

 

 

UCN-3824
1 =83

lomrl: comMPLETED

 

 

 
135

UCN-8820
3 5-88)

MSRE CHAN

 

 

SUBJECT

GE REQUEST

REQUEST NO.

 

TARGET DATES

 

TYPE CHANGE

[Jiec

[] ELeECTRICAL

(] PERMANENT
[0 temporary

FROM

APPROVAL TO PROCEED

 

DESIGN COMPLETED

 

PROCUREMENT COMPLETED

 

 

7] mecHaNICAL

TO

DOCUMENTS AFFECTED

 

(] oTHER

REQUESTED BY

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DATE REQUESTED REQUESTED COMPLETION DATE
PURPOSE AND DESCRIPTION
APPROVAL TO PROCEED
INIT. | DATE INIT. | DATE

MSRE DEPT. HEAD ] a&coesieN
[[] MsRE OPERATIONS [ rD DEVELOPMENT
[7] MSRE MAINTENANCE (] MsSR PROJECT DIRECTOR
O [(] reEAcTOR Div. DIRECTOR
COMMENTS OF REVIEWERS

STATUS

DATE DATE

 

[] peEsiGNn comPLETEDR

[[] work COMPLETED AND TESTED

 

] PROCUREMENT COMPLETED

 

] ORIGINAL DOCUMENTS REVISED

 

[[] work REQ. ISSUED: No.

[[] conTROL ROOM DOCUMENTS REVISED

 

 

 

REMARKS:

 

 

 
137

APPENDIX M

Example of

MSRE Daily Report
139
MSRE DAILY REPORT

period 0800  3/22/68
t

o
0800 3/25/68

The power operation of the reacztor for Run No. 1l was discontinued
at 0100, 3/25/68, by a rod and load scram as planned., After 8 minutes of
subcritical operation, the reactor was returned to reduced power operation
to recover the temperature lost during the scram transient, The power was
then reduced to 10 kw and is continuing at that level and at 1210°F.

Dynamics tests were completed at full power and at 2.5 Mw. The tests
at 2.5 Mw were run because the ternary sequence used in the previous tests
had not been correct. The power was reduced to 2.5 Mw for about 3 hours for
these tests; otherwise the reactor ran at full power until the scheduled
power reduction this morning. There was difficulty in returning to temperature
servo after the dynamics tests, The trouble, which was found to be faulty
contacts on Relay KA-1T70B, was corrected.

Testing continued on the offgas sampler thermal conductivity cells to
determine the effect of the flow rate through the absorber bed. The results
were too erratic to form any conclusions.

MG-3 was found running but off-line three times with no apparent
reason, Normal operation was restored each time without difficulty.

There were two computer outages during the weekend but the computer
was restarted after about 1/2 hour in both cases.

The equipment to be used in gamma scanning the heat exchanger arter
the shutdown was calibrated on a mockup of the heat exchanger and a 50-curie
i927nsource. The gamme scen is to be done with a Ge(Ii) diode to determine
the concentrations of the various fission products that are deposited on <The

heat exchanger tubes.

4 ,EauéQéanei&
C. H. Gabbard

al
140

CURRENT REACTOR STATUS

 

o85St fawrs

74

Date J-ZS-£f Initials
IO0P CONDITIONS
Fuel Loop Coolant ILoop
Contents » ¢ o » o + o o = o /L';c: (’0 /fi,uq/'
Circulating . . o o ¢« « o o & /65 /c‘f.‘;
Pressure (Psig) « « + +» o o o f, ¢ S.&
Maximum Temperature (°F) « « I704.2 /2OS. G
Minimum Temperature (°F) « - . /2LS. 4 /205, 2
Samples Taken « + « o o o o o o e m—
—

Enrichments ¢« o ¢ o ¢ ¢ o o o &
Overflow Tank Level (%) + +

REACTIVITY
Observed - Calculated

Nuclear Power , O Mw
Rod Control Mode /r/ug VO

% 8k/k
Rod 1374 7in.
Rod 2 % in.

235y in loop 6Zf5ka
2350 in salt @,Xz.s’kg

Rod 3 &in.

HEAT REMOVAL
Load-Control Mode Mdzwfd /
Blower 1 O Fr
Blower 2 3 OFF

ACCUMUIATED OPERATING. DATA

Inlet Door 0 in.
Qutlet Door 0 in.
Damper >/ 00 %

This Period This Run Total

 

 

 

 

 

 

 

 

 

 

 

 

Time Critical . . +» + « +» o o « o » o (hrs) 7/:37 44’(%47 ////7/.37
Integrated Power . . . . . . . (Mw-hrs) 4 73 /73 2452/ (SLEL
Fuel Loop Time Above 90CG°F , ., . . . .(hrs) 72 4499 75 2068S~
Fuel Pump Time Circulating Helium , , (hrs) O /1,25 3995
Fuel Pump Time Circulating Salt , . . (hrs) _ 74 4424, 5~ /9959.25
Coolant Loop Time above Q00°F , ., o+ . (hrs) 7,5 ‘747%125’ /25%514 25"
Coolant Pump Time Circulating Helium ., (hrs) 0 o SO08 2
Coolant Pump Time Circulating Salt . .(hrs) 7Z/ J497 7 /¢8/3
Heating Cycles (Fuel/Coolant) , . . . « %o /o 7/4

Fill Cycles (Fuei/Coolant) . . . . . . S/o /O 30/ 3
Power Cycles (Fuel/Coolant) 4 ¢ . o . . /1 10/10 65 /¢4-

Equivalent Full-Power Hours

455 3378.M4 TFoos.sH
141

AU‘I’ ILIARY SYST: STATUS

 

 

 

 

 

 

 

 

 

DRLIN TANKS Contents Temperature Pressure
1os Salt °F D3ig
Fuel Drain Tank - 1 2 G Hrel /Z .‘_1)/3. //, Z
Fuel Drain Tank - 2 Y2, e/ N3O 4,5
Fuel Flush Tank L) Elsh. //éf) f. o
Coolant Drain Tarnk / 1// / /44' 3.3
Mol Storage Tanlk — _,._£___ -_—
CHLMICAL PLANT STATUS:
HELTUM SUPPLY SYSTEM
-
Trailer Pressure /340 psig Moisture £, S5 opm
— -
oxygen _ 0. /S ppm Total Flov _2.4  4/min

 

OFT'CAS SYSTEM
Charcoal Beds in Service Zfi Pressure Drop _——— »psi

Containment Stack Flow 24 % ¢fn
Filter Bed Pressure Drop (in.-Hz0) Roughing fs O 3 Avsolute /.'g :Z

CONTATNVENT CONDITIONS

High Bay Coolant Cell Chem Cell Rezsctor Call

Containment Vacuum
(in.-#20) 0.9

Reactor Cell In-leakage

—_ —_— =7, .ZZé psig

 

4/min  Vapor Cond. Sys. Pressure CZ psig

REEZE VAIVES

Thewed /tfibj /6 Frozen /O3, 20¢ 206
Deep Frozen [JO# /07 /flg /fl? //0 L1, L

AUXTILIARY EQUIPMENT OPERATING

 

&
Computer On/ Helium Purification Unit ' /A
Fuel 0il Pamp 2/ Stack Fan =/
Coolant 0il Pund d’[ Cooling Tower Pump “/

Component Cooling Pumps_fé_ - Cooling Tower Fan __f_/__ -
Air Comnressors ¢/  £3 Treated Water Pump __L_____‘
Air Dryer & / Beryllium Pump ¢ Z

-oseus ¢ %7 %3

ther: /4&/4/“4{_5.; (3 lowers ,2 £4
40-42.
43-44,
45-46.
47 .
48.
49-65.
66.
67-68.

143

ORNL~-TM-3041

Internal Distribution

S. E. Beall 21, A. J. Miller

R. B. Briggs 22. R. L. Moore

W. B. Cottrell 23. A. M. Perry

J. L. Crowley 24, M. Richardson

J. R. Engel 25. M. W. Rosenthal

A. P. Fraas 26. Dunlap Scott

R. H. Guymon 27. M. R. Sheldon

P. N. Haubenreich 28. M. J. Skinner

H. W. Hoffman 29. 1. Spiewak

T. L. Hudson 30. D. A. Sundberg

P. R. Kasten 31. D. B. Trauger

A. T. Krakoviak (K-25) 32. G. D, Whitman

Kermit Laughon, AEC-0SR 33-34. Central Research Library

R. N. Lyon 35. Y-12 Document Reference Section
R. E. MacPherson 36-38. Laboratory Records Department
H. C. McCurdy 39. Laboratory Records (RC)

External Distribution

Director, Division of Reactor Licensing, USAEC, Washington, D.C. 20545
Director, Division of Reactor Standards, USAEC, Washington, D.C. 20545
N. Haberman, USAEC, Washington, D.C. 20545

D. ¥. Cope, AEC-0ORO
T.A. Nemzak, USAEC, Washington, D.C. 20545

Manager, Technical lnformation Center, AEC {For ACRS Members)
Research and Technical Support Division, AEC, ORO
Technical Information Center, AEC