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 | 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%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 / 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