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Exam Rept 50-325/OL-87-02 on 870824-27 for Units 1 & 2.Exam Results:Two of Four Senior Reactor Operator & Eight of Nine Reactor Operator Candidates Passed Exam
	ML20235Y932
Person / Time
Site:	Brunswick
Issue date:	10/09/1987
From:	Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20235Y923	List: ML20235Y920; ML20235Y932; ML20235Z199;
References
	50-325-OL-87-02, 50-325-OL-87-2, NUDOCS 8710200533
	Download: ML20235Y932 (141)
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[9; o' as 5 ENCLOSURE-1 I b ,i I ^ si EXAMINATION REPORT 9 50-325/0L-87-02 e 3 fa Facility l Licensee: Carolina Power and Light Company P. O. Box 1551 .e Raleigh, NC 27602' r Facility Name:. Brunswick Steam E c'ric Plant-9,, M t 7 s . Facility Docket Nos.: 50-325:and'50-324 J w' ' Written and operating examinations' were administered at the Brunswick Steam' Electric Plant ~ near. Southport, North Carolina. Chief ' Examiner: / M './o/9/F7 -J.'F. Munro-- Da'te 61gned e < Approved by:. [ / o/9[$7 J. F. Munro,"Section Chief Date Signed Summary: J.. Examinations'were administered on August'24-27, 1987; ^ Writteh and operating examinations were administ6 red to four SR0 and nine'RO. candidates. Two'SR0s and eight.R0s.pessed the written examination. Four SR0s and nine R0s passed the~ operating examination. Based.on the.results described above, two of four SR0s and eight of nine R0s passed the overall examination. Of the 17 technical corrections-made to both written examinations, seven (41%) 'were due to' inaccurate / incomplete materials.provided to,the Camission for examination preparation. The facility,is encouraged to ensurq'the accuracy and completeness of facility reference material. ] e (W L gO2OO533872016 l ADOCK 05000325i y PDR E__. __

3, - REPORT DETAILS 1. Facility Employees Contacted: P. W. Howe,' Vice President Brunswick Nuclear Project C. R. Dietz, General Manager S. Strickland, Operations J. Moyer, Manager Training R. Poulk, Jr., Regulatory Compliance G. Barnes, Training J. Keith, Administrative Assistant

Attanded Exit Meeting 2.

Examiners: J. F. Munro, Region II s D. C. Payne, Region II M. O. Bishop, EG&G J. H. McGnee, EG&G J. Hanek, EG&G

Chief Examiner 3.

Examination Review Meeting: At the conclusion of the written examination, the examiners provided your training staff with a copy of the written examination and answer key for l review. The NRC resolutions to comments made by the facility reviewers are listed below. These comments did not conform to the format prescribed in of our letter to Mr. E. E. Utley dated June 29, 1987. a. SR0 Exam (1) Question 5.01b Comment accepted. The training material providad does not refer to I MAPRAT in this form but the Process Computer printout P-1 does. This answer will be added to the answer key as an alternate correct response. The utility is encouraged to ensure that all training I materials are both complete and accurate. Point value remains unchanged. .(2) Question 5.09c Comment accepted. The answer key was obtained directly from the training material provided for generation of the written exam. Huwever, since BSEP does in fact not have automatic flow control, this portion of the question will be deleted from the exam. The utility is encouraged to ensure that all training materials are both complete and accurate. Section and total points will be adjusted accordingly. ) i )

(3) Question 6.02 Comment accepted. Credit will be given for an answer of " greater than minimum flow". However, no credit will be given for an answer of "high flow".since this value would have no reference or definition. Point value remains unchanged. (4) Question 6.03b Comment accepted. The reference provided, 50-14, conflicts with the training material provided for the generation of the written exam but is a controlled system description. The utility is encouraged to ensure that all training materials are both complete and accurate. Point value remains unchanged. (5) Question 6.14 Comment accepted. Each candidate's response will be evaluated to determine his interpretation of the question. Listing a specific condition which, while not a reactor safety trip initiator, would l result in a half-scram condition will be accepted. No change to the answer key is required. (6) Question 7.01 Comment accepted. This will be added to the answer key as a possible correct response. Four responses will be required for full credit. Point value remains unchanged. (7) Question 7.16 Comment accepted. This will be added to the answer key as a possible correct response. It should be noted that the Fuel Handling Procedure does not specify which methods to utilize for visual verification of fuel orientation. Point value remains unchanged. (8) Question 8.01 Comment accepted. Answer key will be modified such that the order is not a required part of the answer. Point value will remain the same but will be redistributed. (9) Question 8.02c Comment accepted. The material supplied for the written exam was Rev. 16. The utility is encouraged to ensure that all training materials are both complete and accurate. This section of the question will be deleted from the exam. Section and total points will be adjusted accordingly.

u.f y (10)' Question'8.06b Comment accepted. For safety-related systems, the answer will be. a modified by adding " Approved by SF with the. permission of" to the front of.the sentence.. Point value remains unchanged. '(11).-Question 8.08 Comment. partially accepted.-.The "24 hours a day" statement is a' procedure. requirement'only and is not a determination of whether an Event Evaluation Check Sheet.is required. :The:"Found to be -inoperable for a reason other than actions required by the test" is valid and will'be added as a required part of the answer. Point-value will-remain the same but will be redistributed to accommodate the. additional answer. Examiner. Changes During Grading (1) ' Question 7.06. Added " Increase in 0FF GAS Rad.. Levels".as an acceptable answer. This would be an indication even though the procedure does'not --specify it.' Also accepted either of the " PROCESS OFF GAS RAD HI"= annunciators'since'A0P 5.3 lists these as symptoms of a possible resin intrusion. The facility should' evaluate the apparent inconsistency between the symptoms provided in A0P 5.3 and /0P 26.0 for a. resin intrusion. b. R0 Exam (1) Question 1.16 Comment noted. The number of SRV's is in parentheses and is not a required part of the correct response. No change to the answer key is required. (2) Question 2.09 Comment noted. The actual interlock is sensed from the position of F041 and F042; however, the question specifically asked for l CONDITIONS or PARAMETERS which will not allow CST to CST testing. 1 I Low CST level is a condition which will not allow this type of testing. Likewise, any conditions emanating from CST low level 1 (e.g., CST suction valves closed) will be credited. (3) Question 2.10 I Comment accepted. Head spray will be deleted from the answer key. Because this was listed as a part of Shutdown Cooling, no change in point totals required. i i

v 3g 1 l_- .. t i. '(4): Question 2.11 Comment'acce'pted. Part "a" of the answe.r' key will.be changed to~ allow,- but not require, "in conjunction with a LOCA' signal. Point' =value remains unchanged. - (5)fQuestion2.14- . Comment accepted'. If a candidate lists a response of Generator

Fault, Differential Overcurrent, Phase Overcurrent, Loss of-Excitation, or Reverse Power, credit will be given for response "3"

.of.the required trips. Because'the question was specific in asking for ENGINE. TRIPS'with SETP0INTS, parts "1" and "2"'will also be required for full credit. The utility is encouraged to. ensure that all training materials:are both complete and accurate. (6). Question 3.03- .-Comment accepted. Sign convention will be changed.to that used in Technical Specifications as.follows:

a. > 3 CPS.

b.,_i 1 x 10E05 'No change to point. values required. (7) Question 3.11 Comment accepted. Sign convention will be changed to that used in Technical Specifications as follows: < 90% open and < 20% rated flow No change to point values required. (8) Question.3.15 Comment accepted. " Turbine bypass valves open" will be deleted from the answer key. Section and total points will be adjusted accordingly. (9) Question 4.05-Comment partially accepted. Radiation exposure limits are the same for either 10 CFR 20 or CP&L requirements given the information in the question. The requirements for yearly limits will oe' deleted from the exam. Section and total points will be adjusted accordingly. The reference section will be changed to include 10 CFR 20. L L 3 l l I

4' 1 l-(10.) Question 4.07a Comment accepted. ' Answer key will be changed to TRUE and the reference changed to Standing Instruction 87-026, 4/21/87. Point

value remains unchanged.

Examiner Changes During the Exam (1) -Question 3.07a Question was modified as follows (change is in parentheses): "What automatic actions occur when the scram discharge volume high level scram bypass switch is placed in BYPASS (and the SCRAM is RESET)?" 4. Exit Meeting At the conclusion of the site visit, the examiners met with representatives of the plant staff to discuss the examination. All candidates were very strong in administrative requirements and Technical Specifications; however, several generic weaknesses were noted and are listed below: Candidate Weaknesses 1. The SR0 candidates were unfamiliar with the new LCO computer tracking system. Candidates must be able to demonstrate proficiency with all administrative requirements and systems during the exam process, 2. The operators, in general, had difficulty in finding many of the administrative and procedural documents associated with normal operations in the Control Room complex. This was due to the new Control Room furniture that was recently installed and the subsequent relocation of reference materials. Care needs to be taken to ensure that all candidates are fully cognizant of procedure locations, and any other changes, prior to assuming normal operating duties. The status of an alarm on Reactor Building Component Cooling Water (RBCCW) was brought to the attention of plant management. This was necessitated when several candidates were asked the cause of the alarm and how long it had been locked in. None knew why, despite the fact it has been in locked in for several years. This situation raises a concern on the formality of the training given to candidates as they stand control reem snifts under instruction. Management indicated that they would check intc the training aspects of the problem as well as pursue a resolution to the locked in RBCCW alarm. The cooperation given to the examiners and the effort to ensure an atmosphere in the Contol Room conducive to the oral examinations were noted and appreciated. The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.

Q f gp _., l F4 1 b U.. S. NUCLEAR REGUL'ATORY COMMISSION O. REACTOR. OPERATOR LICENSE EXAMINATION +g ,O, FACILITY: BBUNSWICK 1 & 2 j REACTOR TYPE: BWB-GE4: DATE ADMINISTERED: 87/08/24 EXAMINER: BISHOP. M. .' CANDIDATE llHSTRUGIl0NS TO' CANDIDATE: m ,-.Use.. separate. paper for the answers. ' Write answers on one side only. ' Staple question sheet. on top-of the answer sheets.- Points for;each . question.are indicated in parentheses after the question. The passing. grade requires'at least 70% in each category 'and a final' grade of at least.80%, Examination papers will be picked up six (6): hours after the! examination, starts. ^ % OF . CATEGORY ' % - OF CANDIDATE'S CATEGORY ,__YALME_ _IDIAL SCOBE___ _Y6h!JE__ CAIEGORY AM 04 3_25.50__ e+-7-7 1. PRINCIPLES OF NUCLEAR POWER i PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND' FLUID FLOW 2/.00 2. PLANT DESIGN INCLUDING SAFETY - RLhD AND EMERGENCY SYSTEMS

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JM Jo .-EL;5 - _;L-4t 3. INSTRUMENTS AND CONTROLS' ,,2 M 0 0 ' JJ. IV _2673 _fJL GG 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL /0F. 77 itT:44 Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. l Candidate's Signature 1 1 -l 1 L l VASTB COPY s

c. ,A 1 1 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS i 1 ) During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties. ) 4 I 2. Restroom trips are to be limited and only one candidate at a time may I leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. 3. Use black ink or dark pencil only to facilitate legible reproductions. 4. Print your name in the blank provided on the cover sheet of the examination, 5. Fill in the date on the cover sheet of the examination (if necessary). 6. Use only the paper provided for answers, i 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet. 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write " Lact Page" on the last answer sheet. 9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face d."n on your desk or table.

12. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the quection or not.

15. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16. If parts of the examination are not clear as to intent, ask questions of the examiner only.

17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.

This must be done after the examination has been completed.

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18. When'you complete'your examination, you shall:

i a. Assemble'your examination as follows: (1) Exam questions on top. i (2)- Exam aids - figures,-tables,'etc. (3) Answer pages including figures which are part of the answer. b, . Turn in your copy of the examination and all pages used to answer the examination questions. c. Tr.rn in all scrap paper and the balance of the paper that you did not use for answering the questions. d. Leave the examination area, as defined by the examiner.~ If after leaving, you are found in this area while the examination is still in: progress, your license.may be denied or revoked. J i ) 1 f 1

e ,j' ts Yi< e 3D :EBINQlELES'0F NUCLEAB POWEB_ELANT OPEEATION. Page 4 IHEBM9EXH&MLQSa_ HEAT TRANSFER AND FLUID Fh0_H s [ ' QUESTION! 1.01 (1.00) Tlus reactor trips from full power, equilibrium XENON conditions.. Twenty-four hours later the reactor is brought critical and power level is main. i' 'tained on RANGE 5 of the IRMs for several hours. Which ONE of the 'following statements is CORRECT.concerning control rod motion? a. -Rods will have.to be withdrawn due to XENON build-in. b. Rods will have to be. rapidly inserted.since the critical I reactor will cause a high rate of XENON burnout. c.' Rods will'have to be inserted since XENON will closely follow its. normal decay rate. d; Rods will remain'approximately "as is"'as the XENON establishes its equilibrium value for this power level. i QUESTION' 1.02 (0.50) Which ONE of the following radiation doses would have the WORST biological effect-on a human? a. 1 Rad. of GAMMA b. 1 Rad. of BETA c. 1 Rad. of NEUTRON d. 1 Rad. of ALPHA (Interna)) i i (***+* CATEGORY 1 CONTINUED ON NEXT PAGE ****+)

e c i 1. ERINGLELES__QF NUCLE 6R POWEB PLAMI_QEERATIQHt Page 5 T.UEBMODYUatllCSuBEAT..TRANSFEB_AND FLUID FLQH QUESTION 1.03 (2.50) For each of the pairs of conditions listed below, state WHICH condition would have the GREATER rod' worth and briefly, EXPLAIN WHY. a. Reactor moderator temperature of 150 deg F or 500 deg F. b. For an inserted rod next to a fully withdrawn control rod or next to a-fully inserted control rod. [ Assume average core flux is constant] c. For a rod at position 10 or position 40 of a core operating at 100% power, l l QUESTION 1.04 (1.00) l A reactor heat balance was performed (by hand) during the 00-08 shift due to the Process Computer being 000. The GAF's were computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE. Which ONE of the following statements is TRUE concerning reactor power? a. If the feedwater flow rate used in the heat balance calculation was LOWER than the actual feedwater flow rate, then the actual power is HIGHER than the currently calculated power, b. If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED, then the actual power is HIGHER than the currently calculated power. c. If the steam flow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is HIGHER than the currently calculated power. d. If the RWCU return temperature used in the heat balance calculation was LOWER than the actual RWCU return temperature, then the actual power is HIGHER than the currently calculated power. l (*+*** CATEGORY 1 CONTINUED ON NEXT PAGE 4****) l l l

I 1. EBINCIELES_OF NUCLEAR POWEB PLANT OEEB& TION. Page 6 ISEBUQDXH6MI E _ HEAT TRANSEEB AND FLUID FLQH QUEETION 1.05 (1.50) STATE whether the following thermodynamic properties INCREASE, DECREASE, or REMAIN THE SAME as steam travels from the inlet to the outlet of the BSEP H.P. Turbine. i ) a. Enthalpy b. Entropy c. Quality QUESTION 1.06 (2.60) Following an AUTO INITIATION of HPCI at a reactor pressure of 800 PSIG, reactor pressure decreases to 400 PSIG. HOW are EACH of the following parameters affected (INCREASES, DECREASES, REMAINS CONSTANT) by the change in reactor pressure? BRIEFLY EXPLAIN your choice. ASSUME the HPCI System is operating in automatic as designed. a. HPCI flow to the reactor. b. HPCI pump discharge head (assuming NPSH remains constant). c. HPCI turbine RPM. (***+4 CATEGORY 1 CONTINUED ON NEXT PAGE *****)

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ilca;EBINCIPLES OF NUCLEABiEQWER PL&HT OPERATIONz Page ~ 7.. I IHEBURDXH&MIRS. UEaI; TRANSFER =AND FLUID FLOW-N l 2 0 ' QUESTION 1.07' (2.00) 'During your Shift, an SRV inadvertently opens from 100% power and:1000-1 psia.- Use a1Mollier Diagram or.the Steam' Tables to answer..EACH of;the -l 'following: j j (ASSUME A SATURATED SYSTEM AND INSTANTANEOUS' HEAT-TRANSFER) { l a.. STATE the tailpipe temperature, as'suming atmospheric pressure in'the Suppression-Pool and No Reactor Depressurizatio'n. b. If the Suppression Pool Pressure 1were to INCREASE, STATE whether the iTailpipe Temperature would INCREASE,. DECREASE, or REMAIN THE i SAME. c. If the reactor' starts to depressurize when the SRV. opens, 1 STATE whethercthe Tailpipe Temperature will initially INCREASE, DECREASE, or REMAIN THE SAME in relation to what it would have 1 -done if the pressure had not decreased. .d. STATE-theLReactor. Pressure at which.the Tailpipe Temperature would be at11ts MAXIMUM value.(during the.depressurization). l 1 I LQUESTION 1.08 (0.50) Without core orificing, the coolant flow through a high power bundle i will be less than the flow through a low power bundle because a. The channel quality increases. I b. The two phase flow friction multiplier decreases. c. The fuel rode expand due to thermal effects. d. The bypass flow increases. \\ I i 1 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l 1

g',. 0 ,,1 y 1 RBIHQIELES OEJQGLE6R POWER PLANT _QEEB&TIOL Page 8 1 THE_BMODYHAMICS. HEAT TRANSFER AND-FLUID ELOJ l 1 QUESTION 11.09 (2.00) Briefly explain the response of. reactor. power to a scram from 100% power. Include;the three phases of response and applicable magnitudes, as well i as the primary'cause for EACH in your discussion.

QUESTION-1.10-(1.50) l: Unit 1 power is increased by control rod withdrawal. The void fraction 4

increases 1.5% and the fuel temperature increases 40 degrees as the-result of the rod withdrawal. What was the reactivity worth of the. portion of the control rod that'was withdrawn ? SHOW ALL WORK AND STATE ALL ASSUMPTIONS. \\ QUESTION l'11 (2.00)' l The reactor has been operating at 95% power for several days. The opermtor rapidly reduces reactor power to 60% by reducing the speed of the recirc pumps. During the next few minutes ( 2-3 minutes) the operator notices reactor power slowly increases approximately 3%. BRIEFLY EXPLAIN'the cause of this effect. i I I (+**** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

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J 1. ' PRIU92ELES__9E_HUCLEAR POWEB_EL8NT QEEBAHQB1 Page 9 1 IEEEMQDY:MMICS BEAT TB&HHEEB18HD FLUID FLOW .] -QUEGTION 1.12. J( 3. 00 )- a. List:THREE'(3) parameters which contribute to AVAILABLE NPSH l -(Net Positive Suction Head) for a recirculation pump. Limit your answer to'those parameters which are DIRECTLY indicated =i in the CONTROL: ROOM. (1 50) i b. Consider.TWO Reactor Plant conditions: Low Power and. Low Flow (<10%) OR High Power.and High Flow (>85%). 1. During which condition ~is the REQUIRED NPSH for a recirculation pump greater? (0.50) 2. During which condition is AVAILABLE NPSH for a recirculation pump greater and WHY is it greater? (1.00) l i QUESTION 1.13 (1.00) Using the steam tables, calculate a. reactor cooldown rate (F/hr) for a reactor pressure decrease from 1000 psig to 250 psig in one hour and forty i five minutes -(105 minutes total). SHOW ALL WORK. QUESTION-1.14 (0,50) i The void coefficient of reactivity becomes less negative for'which ONE of the following changes? ) a. Percent void changes from 30 to 40. i b., Average fuel temperature changes from 500 deg.F to 550 deg,F. .c. Core age changes from beginning of life to end of life. d Control rod density changes from 20% to 25%. j e. Reactor pressure changes from 980 psi to 1010 psi. (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

g 3 L._.iERIEDl2LES OF NUCLEAB POWEB_ELANT_QEEB&Tl0E Page 10 IBEBM9DYH8MICS, HEbI_.IBANSEEB_hED_ELMID_EL9W - i i LQUESTION 1.15 (2.50) a. Define the term BETA with regard to delayed neutrons? (1,0) b. When comparing.the individual BETA from thermal fission of U-235, LPu-239, and fast fission of U-236, which BETA is largest? '(0.5) c. .From beginning of life to the end of life does the core average beta INCREASE, DECREASE or REMAIN THE SAME7 EXPLAIN your answer. (1.0) i GUESTION 1.16 (2.50) f Referring to attached Figure 1, " Closure _of All MSIV with Valve Position Scram'"'for-Unit,1, answer EACH of the following;

a. EXPLAIN WHY neutron flux increases sharply at approximately.

two seconds into the event (Point A). (A complete response will include both the action which causes.the spike and the mechanism. .which causes the power increase.) (1.0) la. inly has vessel steam flow returned to approximately 90% of rated at Point B7 (0.5) ) c. WHY is reactor pressure increasing at Point C (eight seconds 5 into event)? (0.5)

d. WHY is reactor water level increasing at Point D even though a steam flow is much higher than feed flow?

(0.5) 1 ) (***** END OF CATEGORY 1 *****)

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'R. PLANT DEEIGN INCLUDING SAFETY AND EMERGENCY Page~ 11 EXEIEMS a n ',5 QUESTION, 12.01 (1.00) . Primary containment isolation' design criteria states, "All I nonessential systems, with the exception of RBCCW, isolate on one .l" or more' containment' isolation signals." LWHAT are TWO (2)'different cooling loads (or systems) supplied by RBCCW l .that. are beneficial :ba mitigating the ef f ects of. a LOCA inside containment thun1 requiring RBCCW remain in service? QUESTION 2.02 (1.00). RHR-injection valves E11-F015A & P will isolate.ONLY on reactor low 'l -waterLlevel'when in the' shutdown cooling mode, j WHAT are'the TWO-(2) parameters the RUR logic checks to confirm the system is in the shutdown cooling mode? (INCLUDE APPLICABLE SETPOINTS) o. . UESTION. '2.03 (2'.00) Q 1. WHAT.is the purpose of the check valve installed in the discharge 4 line cf'each Standby Liquid Control pump? 2. For WHAT specific component failure in the Standby Liquid Control system is it designed to provide protection? QUESTION 2.04 (2.00) The RBM gain change is done so the RBM output will be equal to or greater than'the reference APRM output. _WHAT are TWO (2) reasons for changing the. gain? (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l

v. l 2. ELANT DESIGN INCLUDING SAFETY AHD._EMERGEU_QY Page 12 AYSIEUS Kl' l 3 i V L l l-QUESTION -2'.05 (l'.50). DESCRIBE the operation-and function of.the AUXILIARY TIMER in the 1 reactor manual control system. (INCLUDE' APPLICABLE SETPOINTS AND _ PROTECTIVE FUNCTIONS.IN YOUR DISCUSSION) 'l 1 ~ I l I -QUESTION 2.06 (3.00) l ' Answer EACH~of the following concerning the ADS gas accumulators: a. WHAT is the normal gas supply? b '. WHAT is the' backup gas supply? c. . WHAT.are'TWO (2) design criteria which are considered in determining I the accumulator size? d. WHAT are TWO.(2) of.the three conditions or signals that will i . automatically line up the backup supply system? (Setpoints not Lrequired) f i ' QUESTION 2. 0'7 (2.00)' Answer EACH of the following TRUE or FALSE: a. None of the ADS valves have controls on the remote shutdown panel, b. Safety relief valve position indication is provided by acoustic monitors and valve limit switches. c. Safety relief valve tail pipe thermocouple provide a common VALVE LEAKING annunciator in the control room. ( d. Position indication for the vacuum breakers located on the safety relief valve downcomer piping is provided by an amber light on the RTGB., i ? (***** CATEGORY 2 CONTINUED ON NEXT PAGE 4****) i

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j [id,,y" g 1 %dMt!I_QESIGN INCLUDING 2 SAEEIX AND EMEEGENCY Page'13 ] LSYSIEMS. F '"l QUESTION l 2.13 8 (1 00)! l -1 4 .The HPCI system minimum flow valve, E41-F012, is interlocked to close if' ) ~ .thefHPCILateam supply 1 valve, E41-F001, or HPCI. turbine stop. valve, H E41 V8, areofully closed.lWHATradverse'effect(s) could occur if this L interlock' failed.to actuate?- { V i QUESTIONI 2'.09 (2.00) ,3 WHnT.FOUR (4){different' conditions or parameters, when sensed by the-HPCI logic,lwill not allow full flow (CST to CST valve lineup) .l 1 . surveillance testing 7.(Specific values or setpoints not required) -QUESTION-2.10. (1.50)- WHAT are the SIX (6), basic modes of operation for the RBR system?' L QUESTION 2.11 (2.50) .In regards to the RHR loop injection valves, F015A/B and F017A/B: a. WHAT interlock must be satisfied to open BOTH injection valves in a loop? (0.5) b. WHAT is the-purpose of this open3ng. interlock? (0. 5') c. Following an automatic initiation, WHAT INTERLOCK ('s) must be satisfied to close each of these valves? (1.0) d. WHAT function is allowed after satisfying the close inhibit interlock (s)? (0.5) lc {. (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l.

(?* ; lv 7( c o +, m,1 .ELAMI' DESIGN INCLUDINGLSAF_EIY_6HD_EMEBGEEGY- .Page 14 EX&IEMS: [ QUESTION' 2.12[ (l'.50)

Concerning the Core Spray sparger line break. detection DP

' instrumentation: a. WHAT.are.the,high. side and low side pressure sensing points? (1.0) 'b. .Is theLnormal-(sparger intact) indication at rated power ~ positive 1or negative with respect to aero? (0.25. c; Is'it;normalL'for.the d' loop and B loop indications to have different values when operating at. rated power? (0.25) IQUESTION-2.1'3 '(1.00) Answer EACH: of ' the following. TRUE or. FALSE: i a. If the' control' switches'.for the, Core Spray. inboard and outboard l injection valves, E21-F004A/B and F005A/B,'are placed'in CLOSE'with l an automatic initiation signal present, the valves will CLOSE and a white " CLOSED SIGNAL SEALED"-light will illuminate. .i b. 'If-a Core Spray pump control switch is placed in STOP with a white l " INITIATION SIGNAL SEALED IN" light illuminated, the pump will STOP. ~1 QUESTION 2.14 -(2.00) LIST THREE (3) automatic engine trips provided when the Diesel Generators are running in the emergency mode? (SETPOINTS REQUIRED) QUESTIONL 2.15. (1.00) WHY does the SBGT system operating procedure, OP-10, not leave the SBGT control room switch in the " STANDBY" position during plant operation? (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i l )

___ __ l, j 4 m . c- .< 2 PLANT DESIGN INCLUDlHG SAFETY'AND EMERGENCY.. Page.15; ,)

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{ 1 1 0 4 I 1 QUESTION. 2.16: (2.50) l .WHAT.is the function ofithe gland exhauster. system associated-a. with the'RCIC system? (0.5) b '. : WHAT FOUR'(4) major components exclusive of piping add valves

make up the RCIC sland exhauster system?

(2.0) t b 1 (***** END OF CATEGORY 2 *****) I

n,; '~ s' XhbTRUMENTS"ANDCONTROLS: ' Paged 16' 3; s y l \\ iQUESTION' L3.01' (1.50)l

The.SBGT:'aystem:has; temperature switches, TS-3, 4,.5 and 6, located'in.

the carbon' filter banks which operate at 210 F.'

L4 LIST 1THREEi(3) automatic functions that occur when these switches I operate. QUESTION' 3.02 (0.50) , Answer the FOLLOWING=TRUE or FALSE: 'If an automatic TRANSFER TRIP" occurs on the Brunswick-Fayetteville1 230 KV line,'a " GUARD FAILURE ALARM" annunciator will be received in the~

control room.

'Q'UESTION

3.03 (1.50)

LLIST THREE (3) rod blocks initiated by the SRM'S.that are a function.of-neutron count rate? '(SETPOINTS REQUIRED) ] . QUESTION 3.04 (2.00) 1 a. WHAT are TWO (2) interlocks associated with the bus tie breakers between emergency busses E7 and E87-I b. WHY are these interlocks necessary? I l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) L__ _. ___. _ - - _ _ _ _ _ _ _ _ _ - - - _ _ -

.;}. ~INSIBilMEN.TS_AHD_CONTBQLS Page 17 QUESTION 3.05 (1.75) WHAT is the EFFECT and the REASON it occurs, on EACH of the following loads supplied by the UPS electrical system, if the UPS power supply is lost? INCLUDE in your answer any automatic protective actions (scram or rod block) that would occur as a result of.this power loss. a. EHC (assume operating at 100% power) (0.5) b. Feedwater Control (assume operating at 100% power) (0.5) e, Reactor Manual Control (assume operating at 20% power) (0,5) d. Rod Worth Minimizer (assume operating at 20% power) (0.25) QUESTION 3.06 (2.00) TRUE or FALSE Answer EACH of the following TRUE or FALSE as they apply to the GE type 555 ("GEMAC") level detectors used at BSEP: l a. They are used to sense reactor level for NR level and feedwater control. b. The reference leg is compensated. c. The electrical power for one instrument is supplied from the UPS j

bus, d.

They provide RCIC, HPCI, Main and RFP turbine trips in a two out of three logic. l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) L___________

v, 3. _lESIEUMEHIQ_AHQ_COHIBOLS Page 18 QUESTION 3.07 (1.50) a. WHAT automatic actions occur when the scram discharge volume high level scram bypass switch is placed in BYPASS / And //r,5 c Adm f5 M56 7 b. WHAT position (s) must the reactor mode switch be in to allow BYPASS of the scram discharge volume high level scram function? c. WHAT additional protective function is inserted when the scram bypass switch is in the BYPASS position? QUESTION 3.08 (2.00) After a reactor scram, the CRD pumps will try to charge all 137 HCU accumulators at once, a. HOW are the CRD pumps protected from damage and trips during this condition? (0.5) b. WHAT combination of control signal and system design configuration is utilized, to direct the maximum CRD flow to the HCU accumulators, during this condition? (INCLUDE WHERE IN THE SYSTEM THE CONTROL SIGNAL IS SENSED) (1.5) QUESTION 3.00 (3.00) a. WHAT are FOUR (4) RWCU system isolation signals that will close BOTH the RWCU containment isolation valves, F001 and F0047 (setpoints not required) (2.0) b. There are TWO (2) pressure signals (LOW /HI) which will close the letdown flow control valve, FCV-F033. WHERE in the system are each of these TWO signals sensed? (Use a reference point between the two closest major system components) (1.0) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) t

2 .IUSIEUMENTS AUD_QQHIBQLS Page 19 QUESTION 3.10 (2.00) Answer EACH of the following TRUE or FALSE as they apply to the APRM'S at Brunswick: a. The reference signal to the upscale thermal trip (flow biased) unit is not affected by mode switch position and is always 0.66w + 54% from the slope and bias circuit. b. With the the mode switch not in run, the fixed reference signals of 15% for a rod block and 25% for a scram prevent exceeding the safety limit for core thermal power below 800 psig and 10% of rated core flow. c. With the mode switch in run, the upscale neutron trip reference signal becomes 120% with a time delay whose time constant approximates the fuel time constant. d. The APRM downscale with companion IRM Hi-Hi or INOP scram is only functional when the mode switch is in RUN. QUESTION 3.11 (3.00) There are THREE (3) speed limiters in the Recirc. flow control system which function to limit the maximum speed demand, a. WHAT is the maximum speed demand limit imposed by EACH limiter? (LIST EACH LIMITER Atm THE APPLICABLE SETPOINT. ) (1.0) b. During what plant conditions are the #1 & #2 limiters in control and imposing limitations on Recire. pump speed? (SETPOINTS REQUIRED) (2.0) (***** CATEGORY 3 CONTINUED ON NE.XT PAGE *****)

g- '&.s _. i .L-IUSTEDUENTS AND_CQUIBQLS Page-20 QUESTION' 3.22L .(1.00)- L L Concerning the EHC system valve closure sequence for a' turbine overspeed-condition, CHOOSE the MOST' correct: response from the.following' choices. Assuming 1100% steam flow and.100% load selector setting: l a. The' control-' valves will throttle to limitioverspeed during the' p' l -first 5%.(90. RPM), and'the intercept valves'will throttle-closed .from.105% to 109% turbine-speed. If turbine speed continues: L to' increase to 110% (1980 RPM) the stop valves will ehut, thus tripping the turbine, b; 'The' control-: valves wi11' throttle to limit'overspeed'during the first.5% (90: RPM), and the intercept valves.will throttle closed-from 105% to'107% turbine 1 speed. If turbine' speed continues to increase ~to-110% (1980 RPM) the stop valves;will shut, thus tripping the turbine. I c. The. control valves'will throttle to limit overspeed'during the first 10%'(180 RPM)', and the-intercept valves will close at'110% turbine speed. If turbine speed continues to increase'above 110% (1980 RPM) the~stop valvestwill shut,.thus tripping:the -turbine, ~ d. The control. valves will throttle to limit overspeed during the first 5% (90 RPM),- and the intercept. valves will throttle closed from 105% to 110%Jturbine speed. If turbine ~ speed continues to increase above 110% (1980 RPM) the stop valves will shut, thus tripping the turbine. I i . QUESTION 3.13 -(1.50) WHAT are THREE (3) methods of tripping the main turbine front standard \\ mechanical trip valve? i i e l I 1 (**4** CATEGORY 3 CONTINUED ON NEXT PAGE *****) j

3. INSTRUMENI.3_6ND CONIROLS Page 21 QUESTION 3.14 (1.00) The power factor of the main generator can be changed by adjusting the field current. For generators operating in parallel STATE if the following field current adjustments will INCREASE or DECREASE the listed parameters. a. Increasing the generator field will the KVAR supplied to the system. b. Decreasing the generator field will the power factor /J~ ' QUESTION 3.15 ( 2. &T; With regards to the APRM setdown feature utilized on Unit 2, provide the following information, a. WHAT is the fixed APRM scram setpoint when APRM setdown is in effect? (0.25) b. WHAT plant condition (s) (INCLUDING TIME DELAY) will place APRM(a,>>) setdown in effect? (1.20) c. Is the APRM setdown automatically reset when the annunciator alarm clears? (0.5) QUESTION 3.16 (1.00) Answer EACH of the following TRUE or FALSE: a. The containment atmosphere control (CAC) and dilution system consists of THREE (3) channels, 1260, 1261, and 1262 for monitoring containment radiation. One channel for particulate, one channel for gaseous, and one channel for iodine, b. An alarm on ANY CAC radiation monitoring channel will initiate a Group 6 isolation signal. (44*** END OF CATEGORY 3 *****) \\

m am -, 1

.s.

p. C._ESQQEDMBES - NORMAL 2._AHHORMAL, EMERGEHQY Page 22' . l

4HR_BADIOLOGlGAL_QQHIBQL i

a '-QUESTION 4.01J (3.00) ?

)

-WHAT are SIX ~(6) of the'seven. plant conditions resulting in an MSIV. Group I isolation? (Setpoints not required) . QUESTION 4.02 .(1.00) WHAT protective action will occur if 155 psig is exceeded during high pressure turbine warming, and WHY does this action occur? ' QUESTION 4.03 (1.00) Under WHAT circumstances are the service water system /RHR system cross-tie valves, E11-F073 and E11-F075, designated to be OPEN? I' QUESTION 4.04 (2.00) WHAT are FOUR-(4) separate visual indications used to confirm correct orientation of a fuel bundle placed in the core? o 7J' QUESTION 4.05 ( 1-St$ ) COMPLETE-the following table concerning the NRC radiation exposure limits for a radiation worker assuming that NRC form 4 is completed and on file, rem / quarter rc w /., n r Whole' Body Skin Extremities l' 1 (***** CATEGORi 4 CONTINUED ON NEXT PAGE *****) I

y- .e. ' % i;. SR/,.= r,,. ' i .:#x v o n /Page,29 i4. "PROCEDUBEG - HQBMAL; ABNORMAL, EMERGENCY' yf i M D_B6plghQG CAL QQHIROL J q l< l. .l .i 9,. f A!- 4 LQ'UESTION 4.06" (1.00) { Under.WHAT circumstances is the Control" Operator. authorized tol sign? .j 1 block 2.2, Safety' System Availability and Approval lfor;ClearancsLto be' R Lhingfon.th'efBSEP. Equipment Clearance Form?. j 'l ~ .1 c 1 .; y + i QUESTION' 4.07f L(2i OO): c 4 .,... e j Answer EACH of the following~TRUE or FALSE: l a. A tripped breaker or. overload.can not be resetLmore thantonce.-.. l b, LWhen preforming valve l lineups,. all normally open valves with the' exception off motor' operated: valvesfare backsbated. ) L x

c.. Stem travel Jis-an acceptable method for verifying vslve position.

d. Control switches:for motor operated throttle valves ope at'ed'from l i the RTGB,"are-held in the-closed position-for'an additjonal.10-seconds after full closed indication is received. i a e QUESTION 4.08 (1.00) LISTTWO-(2)confirmationcheckstheoperatormustmakebefore . securing or placing an ECCS or RCIC in~" Manual". sr j I -QUESTION 4.09 (2.00) j i ~ LIST THREE-(3) immediate operator actions required for a W Recirculation Flow Control Failure - Decreasing Flow. f l ~ i 1 QUESTION. 4.10 (1.00) WHY is it not advisable to operate with core flow less then 35 mlbs/hr and a control rod pattern greater than the 80 % rod liue? l l l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****). l )

y - -= y .[ ;.. ; ,,9 gy 3._ C Page'24 4;. 0PROCE2!!$E$wNORM61u_ ABNORMAL; EMERGE![CI CONTROL. H

AHD_RADIQLOGtQA(M

' T af tg(! y c p.. jf%, tr. ~- c _ ((,, ' [ f Wl L

QUESTION a4:11

,(O.50)' .v I +" .WATLis the maximum reactor thermal power. allowed.in steady. state ll SINGLE-loop ~ operation? I .I f i t I LQUESTION i > J4.12:

(1'.03)

'FollowingLa double recirculation pump trip, WHY does AOP 04.3. direct 1the.operatorEto reduce.CRD flow.to 30 gpm? < QUESTION 4.13 (2.50) 'Iri additien to " periodic test 13.1 completed unsatisfactorily", LIST FIVE (5).symptoisLindicating'fallure of-a jet pump. -QUESTION 4.14- .~(3.00) . Concerning an air system failure: 6[p7 a. WHAT automatic action will occur as each of the following parameters ' s reached? i m:

r-1.

Service air header pressure decreases to 105'psig. y ~ Q 2. Instrument-air header pressure decreases to.103 psig. !M 1* - .~ 3. Instrument. air header pressure decreases to 100 psig. l 4. Noninterruptible instrument air header decreases to 95 psig 5. Division I or II noninterruptible instrument wir header pressure falls below 95 psig, s b '. What operator. action,is required if noninterruptible instrument air pressure cannot be maintained above 95 psig ? j i (***** CATEGCRY 4 CONTINUED ON NEXT PAGE *****) v. _________o

N[r.,; - h,D4 PROCEQQBES_. _HQBMAL. ABNQBMAL, EMERGENCY Page.25 AHp__BADIOLOGICAE CONTROL p 1 n 'l -QUESTION 4 15 -(2.25) Concerning a. stuck open SRV: a. LIST'the IMMEDIATE operator actions taken in an attempt to close a stuck open'SRV7 b. .Under WHAT TWO'(2) conditions is the operatorfdirected to scram the t reactor in accordance with AOP 30.0, Safety Relief Valve Failure?- (SETPOINT REQUIRED) QUESTION 4.16 (1.00) ASSUME forced circulation from either reactor recirculation or RHR cannot be resumed within a 30 minute period with the plant in oper;tional condition 4. a WHAT operator action is suggested-in accordance with OP-17 Residual Heat Removal? (INCLUDE APPLICABLE SETPOINT/ LIMIT) b. WHY is this action necessary? l l l (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********) ._-_______________n

S ) )P)B I S (E P(S (PU 6 I N,8 1 j I hl 2 i \\ d E R. ET(O SBCL S t AF tS R (B R RE.UFM . R l SP SSDA 3 TOE 1 Ql E E O RE mEV T PN L S U IL ENE L / nim EM tE 6 SA B E SE RR ) ET U0N /o C VS T0C1

(

I234$ 2 6 3W Q TO j N* I 0C L ES F S ( E EM R 2 O T 4 5 \\ XU LF 6 XTA WE x 2 FH ,2 5 N O RA 6 1 T F U R NS VI ' O i2 D EU SR 0 M 5 E x 0 O 2 2 L 1 1 ^o a* g LN ) AO s P. T )S S p X REE FT r U I t EL KOH OI T NC)o EO F W S RT O P (Il* R N/ ,DN 5 HDDEO EA L L {L W o t) E TE F S / UE E NHW EO* S \\\\a F V V L (t X OM O EEEF U ELA RLL 5 TW L FE 2 L FtC 4 A T C fFRS 8 CEELL Nb INSS NiR E NF 3 F TL O (NNI USME R LEE ESSEE S TK V DS RI UAEE V RROR E EEVE ELWNCD NP AFV 2345 2345 I l 6 ) )C 3 C 4 5 E ES S ( (E E 4 MI s ,T T I 4 2 A d s \\ I 4 x \\ 5d ~ p

)t

- 7 Y,$ l / = 2 5d - = = ,= p g g O C 0 C 5 1 5 O 0 0 5 I 6#$ u-E g_ 4

i \\", s EQUATION SHEET f = ma y=sht }~ v = mg s = v,t + hat Cycle efficiency = "E ^ 2 E = mc a = (vf - v )/t 2 0 -At A = AN A=Ae KE = env. v i g = v, + at PE = mgh. m = e/t A = in 2/tq = 0.693/tg. .l - W = vaP. b(eff) = (t,;)(ts) . t AE = 9314m (fg+t) b - h = [nC 4T ~ I. I. IX p k=UAAT-I'. y,-UX ,~Pwr=W T -I = I,10 * ~ g SUR(t). TVL = 1.3/u y=p lo e /T HVL

0.693/u t

P=P o SUR = 26.06/T ~ T = 1.44 DT SCR = S/(1 - K,gg)- .I, A o

ff

. SUR = 26 CR = S/(1 - K,gf,) g, g 1( aff}1 " 2(I ~ Eeff)2 T = '(1*/p') + [( i

p ) / A,g g ]

~ =- p T,=-L*/ (o - Ty M = 1/(1 - K,gg) = CR /CR g 0 l - T = (3 'p)/ A p eff M = (1 - geff)0 (I ~ Keff)1 I 8"I -1)/Kaff

OEeff eff SM = 0 - K W

/K aff eff eff [1*/TK,'ff.] ~ + [I/(1 + A,gg )] 1* = 1 x 10" seconds p= T P =.I6V/(3 x 10 0) g ~ aff = 0.1 seconds A E = Na I;dy=Id22 WATER PARAMETERS I d '= I d g 2 2 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d g,,t,,,) l 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) I ft3 = 7.48 gal. MISCELLANEOUS CONVERSIONS 3 10 l Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps Density = 1 gm/cm I kg = 2.21 lbm 3 Heat of va;orization = 970 Etu/lbm I hp = 2.54 x 10 BTU /hr Heat of fusica = 144 Btu /lbm 1 Hw - 3.41 x 10 Btu /hr i 1 Atm = 14,7 psi = 29.9 in. I'g. I Btu = 778 ft-lbf 1 ft. H O = 0.4333 lbf /in 1~ inch = 2.54 cm 2 F = 9/S C + 32 'C = 5/9 (*r - 32) .-----a_--

7x-. Q Y;f l, 7 T." "EBlHQlELES QF NUCLEAR'POWEB PLANT OPEBAIIQBt Page.26: 2 w' IHERMODYH6MICSi' HEAT TRANSFER ~AED._EL.U_ID FLOW m

ANSWER'

1.01~

.(1.00); , c :. ~(1.0). REFERENCE BFNP: XENON'& SAMARIUM LP, P.4,'12;.RQ 85/03/05-JGONS:.LP OP-NP-514,cp.;5 R

BSEP: '02-OG-A,Mpp 57 - 60 1

BSEP Lesson: Plan 2A, Reactor; Theory;pp. 199 Sec. 15.2 Lesson.Objecti've .611&62 1 292006K112' c...(KA's) f ANSWER 1.02 (0.50) d. (0.50)' a 1 REFERENCE ~ BFNP: GET; Mitigating 1Rx Core Damage LP,p 4 GGNS: OP-RP-502,P.5-7' 'BSEP ~ Lesson Plan 3A, Radiation Control, Rev. 2 Sec. 2.14 pp. 11 Objective h.'(RO)-and g.(SRO) 294001K103- ..(KA's) P -ANSWER-1.'03 -(2.50) -a. 1At 500 deg F..[0.25), Ac moderator temperature. increases, neutron leakage out of the fuel bundles is increased, thus the control rod -1 is exposed to higher neutron flux and rod worth increases. [0.5] q (0.75) ) .b..:The withdrawn rod [0 25), neutron flux is higher in this area,- l .thus rod worth is greater. [0 5]. (0.75) ] ~At110 [0.25], the void content of the upper portion of the core is c. high'at operating conditions, the effects of deep control rod with- 'drawal can be substantial radially. The negative power effect of increased voids above the control rod is not seen because the channel length above the control rod is relatively short. [0. 75: ) (:.0) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) n n -) Gj i i I .OY i= i

r v.. m _7 .. l i '.. PRIElELENOF ' NUCLE 6R ' POWEB_ELSHI_QEEBATlQH1 Pag'e'27 IEEBMQDXHADEL__ BEAT. TBANSFER _AND_ELUID ELQH s "1 REFERENCE -Susquehanna Reactor Theory SC023 A-G rev.0, pg 9,& 10 and attachment.'A, pg-1 & 2. BSEP Lesson; Plan 2A, Reactor Theory,.Rev.2 Chapter 14'pp. 187.,189, and: -190.: Lesson' Objective No'. 59 o 292005K109< ...(KA's) l ANSWER. 1.04 .(1.00)' .c. (1.0). REFERENCE-

EIH:

L-RQ-667, p 10 1BFNP: Rx Heat Balance LP';. RQ 85/03/05 BSEP: Lesson Plan, Heat Transfer, Chap.

8. pp. 8-50 th.

8-59.. Objective is~second from bottom of page 8-1,no assigned number. 293007K111 ..(KA's)' LANSWER l'.05 -(1.50) a. Decreases b. Increase's c. Decreases

3 @ 0.5 ea.] (1.5)

' REFERENCE -Second Law of Thermodynamics BFNP: BFN Entropy LP; BFN Energy, Power, and Enthalpy LP; RQ 85/03/03 & 04 BSEP: Lesson Plan-Heat Transfer, Chapter 6. Lesson Objective = last 3 .on page 6-1 (no numbers assigned) 293003K123 ..(KA's) i (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) i;l' l: L_

OL,.; '

LgI PRINQlELES OF NUCLE &R POWEB PLANT OPERATIOHz Page 28 IHEBMQDluddICS; HESI_IB&ESEEB_&HD FLUID FLOW p ANSWER-11.06-

(2i50 0 a.

Remains constent. [0.25]' Flow is' controlled by the HPCI. flow controller which will attempt to maintain a. constant output flow regardless of reactor pressure [0.5]. (0.75) b '.. Decreases.. [0.25] The flow controller functions to maintain a constant flow, :thus pump discharge pressure is: decreased along-H with-the. decreasing reactor pressure to maintain constant flow. OR Since the flow controller maintains a constant flow to the reactor, as reactor pressure decreases, the pump discharge head must decrease to maintain a constant flow-(constant NPSH). [0.75] (1.0) Decreases. [0.25] To. maintain a constant flow, turbine RPM ic.-. must also decrease. [0.5] (0.75) REFERENCE SSES Fluid Mechanics: pumps, SC023 E-4 HPCI, SYO17 C-6 rev 0 .BSEP Lesson Plan - 14B Rev. 4,HPCI, Sec. 3.B.1. pp. 12 & 13. Lesson Objective No. 4. 206000K505 ..(KA's) ' ANSWER 1.07 (2.00) l a. 295 deg F (+- 15 deg F) i b. Increase c. Increase d. 450-psia (+- 50 psia) [4 @ 0.5 ea.] (2.0) [ j (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) \\ ._--___-__--_-___A

l' ~EEINDIELES OE_HURLEAB'POWEB PLANT OEEBATIQS1 Page 29 IDEBMODINAMICS J EAT TRANSEEE_AND_EL'UID FLOW REFERENCE Steam Tables /Mollier Diagram BSEP Lesson. Plan - Heat Transfer Chapter 4. Lesson Objective No. 3 and 5 from bottom of page 4-1(no number assigned) 218000A101 ..(KA's) ANSWER 1.08 (0.50) a. (0.5) REFERENCE BSEP, HEAT TRANSFER, CH. 9 Page 9-51 Lesson Objective = Third from top of page 9-1A (no number assigned) 293008K131 ..(KA's) ANSWER 1.09 (2.00) The prompt drop (0.25) (is caused by the prompt neutron multiplication change) caused by the change in Keff. (0.25) During the transition phase,(0.25) the delayed neutron precursors are adjusting to the new value of reactivity.(0.25) The stable period (0.25) will be established at approximately -80 sec.(0.25) after the short-lived delayed neutron precursors die out (0.25) and the long lived precursors control power decrease (0.25) (2.00) REFERENCE BSEP Lesson Plan - Reactor Theory Rev. 2 pp. 144 & 145. Lesson Objective No. 46 292003K106 ,.(KA's) (***** CATEGORY 1 CONTINUED ON NEXT PAGE ***+*)

' 1.' PBINGlELES OF NUC.J468_EQWEB_ELANT OPERATION Page 30 1 T.HEBdQDXHAMics. HEA CCBAMSFER AND FLMID FLOW ANSWER 1.10 (1.50) Worth due to voids = (1 X 10E-3 dK/K/%V) (1.5%V) 1.5 X 10E-3 dK/K = (1.0X10-5 dK/K/F) (40 F) Worth due to fuel temp. = = 0.4X10E-3 dK/K 1.9 X 10E-3 dK/K [3 @ 0.5 ea.] (1.5) ROD WORTH = VOIDS + FUEL TEMP. = REFERENCE l Reactor Theory Sec. 1 Pg. 16,14,and9. BSEP Lesson Plan 2A, Reactor Theory, Chapt. 14 pp. 172 & 181. Lesson Objective No. 58. 201003K506 ..(KA's) ANSWER 1.11 (2.00) The reactor is now producing less steam to go to the turbine.(0.5) There will be less extraction steam and reheater drain steam going to the feedwater heater (0.5) Therefore less feedwater heating will occur resulting in colder feedwater entering the vessel (0.5) which will cause reactor power to increase (about 3%) from the positive reactivity addition (alpha m) (0.5) (2.0) REFERENCE SSM BOOK 9, CH 18 - A,Sec. 2.2.4 Pg. 39. BSEP Lesson Plan, Reactor Theory pp. 164 Lesson Objective No. 52 AOP 03.0, Moderator Temp. Decrease Sec. 2.0, Symptoms. 259001K604 ..(KA's) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

pc; 7:

t ie

2. :

1' "',L1' PRIHQIELES 'OFJ NUCLEAR - POWEB_ELAHI_QEERATION_,;. Page 31 7THERMQDXHAMICS. HEAI_TB6HSEER AND FLUID FLOW' L IANSWER: 51.l'2 ' (3.00)' a,- Feedwater'. temperature Feedwater flow' RPVLpressure 1 RPV water level [any-3 @ 0.'5'ea.): (1 -~ 5 ). to, 1. High-flow,.High power-(0.5) 2. HighLflow, High' power.[0.50], due'to-the. increased inlet-subcooling'from the' increased feedwater flow. [0.50) (1.0) REFERENCE. r GE BWR ACADEMIC SERIES ON HEAT! TRANSFER AND FLUID FLOW. LBSEP HO-10-2A'(RECIRC FLOW' CONTROL);PG 42., -202001K402 ..(KA's) 1 LANSWER 1,13 (1.00)- Obtain corresponding' temperatures from the steam tables by interpolation. I 1000 psig'= 546.3 deg F'for 1014.7 psia. (0.25). H 250 psig =J406.0 des F for 264.7. psia. (0.25). j Determine'the temperature change: 546.3'- 406.0 =140.3 deg F. (0.25) 1 Determine the rate of cooldown: 140.3/1.75 = 80.2 deg F/hr (0.25) J (Will accept + or - 2 deg. on temps.) i REFERENCE. 4 l ' Steam Tables j 4 BSEP Lesson. Plan, Heat Transfer, Chap. 4, Steam. Lesson Objective 5 from bottom.of page 4-1(no assigned number. 293003K123 ..(KA's) ' ANSWER 1.14 (0.50) c. (0.5) J l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) m

u : o _....,. 2'. ERINQIELES OF NUCLEAB_EQWEB_EL&NT OPERATION. Page 32 .ISEBUQpYNAdigh_UEAI_TBAHSFEB_AND FLUID FLOW REFERENCE l 1 USEP Lesson Plan,-HO-02-2-A, Figs 45-49. 292004K113 ... ( KA ' s ) ANSWER 1.15 (2.50) a. The delayed neutron fraction is the percentage-(fraction) of fission neutrons:that.are born delayed. (1.0) I b. U-238 (0.5)- c, Decrease (0.25) As Pu-239 production increases (0.25) and U-235 decreases (0.25) the core average will decrease.due to Pu-239's beta being so much smaller (0.25). (1.0) REFERENCE NUS Reactor Theory section 11.3 BSEP: Lesson Plan HO-02-2/3-A, pp 132&137. 292003K104 ..(KA's) ANSWER 1.16 (2.50)

a. MSIV closure causes reactor pressure to increase [0.5) which collapses the voids (decreases the-void concentration) adding

[0.5) (1.0) positive reactivity /. . // c ce,m r., Qt

b. All L91 Safety Relief valves actuate.

(0.5) c. Decay heat is causing pressure rise due to SRV closure. (0.5) i

d. SRV actuation reduces vessel pressure causing level swell (0.5)

REFERENCE l 1 1 -BSEP: HO 05-2-A, Anticipated Operational Occurrences Lesson Obj. b, pp. 1 (Figure is ATTACHMENT 8) ) 239001A210 239001A204 239001K316 239001K308 ..(KA's) l j i (***** END OF CATEGORY 1 *****) l

L' I2' PbABI_ DESIGN _lN_CLUDlHfi_SAFET AND_.EMERGENC1 .61STEH& PageL33 {! 1 ANSWER; .2.01. (1.00)- i

.1.

Containment Penetrations (0.5) u l 2. Drywell. Air Coolers (0. 5 )- y

REFERENCE-LBSEP: SSM,fPrimary and~ Secondary Containment, P.

11, Lesson Objectively. '223001A101- ..(KA's) l . ANSWER' 2.02 (1.00)' 1. Reactor Pressure < 140 psig .(0.5) '2. Both. shutdown cooling suction valves are open (O.5) IREFERENCE' BSEP: SGM, Primary and Secondary Containment, P. 64, Lesson Objective g. 205000K403 ..(KA's) j l o ' ANSWER 2.03-(2.00) i a. To prevent bypass flow from one SBLC pump to the other i s 'SBLC pump. (1.0) a. The relief valve on the discharge of the SDLC pump. (1.0) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) _ _- __ - L

o[.;j.:q b' 2 '.- ! PLMLDESIGN INCLUDING SAFEILMD_EMEBQElLQX 1Page 34.

SYSTEMS

.y ' REFERENCE

BSEP:.SSM,fSBLC P.7, Lesson Objective e.

2110000007' ... ( KA s ). Ol fANSWER-2.04- '(2.00)> 1. .The local power may be lower than the core average. power. ( 1. 0 ).

2.

Several"of the' highest reading:LPRM'S normally fed to the RBM might be bypassed. (1 0) -REFERENCE-BSEP: 'SSM, RBM, P. 6,LLesson Objective a. '215002K102

215002K101

..(KA's) ANSWER 2.05 (1.50) If.a withdraw' signal is sent to the directional control valves [0.25] for more then 2 seconds [0.25] the auxiliary timer will time out. .When~the auxiliary timer-times out, it will' generate a select block which will deselect the rod. [0.50].This prevents aLfaulty master Ltimer'from causing.an uncontrolled withdrawal signal. [0.5] (1.5)- REFERENCE' i BSEP: SSM, RMCS, P. 14, Lesson Objective d. 201002A303 (KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) w_

.~ s. . ja ji!a'., ".1 gmPkN.1 DESIGN 1EGLUDINGSAFETY-AHDEMERGENCY-Page 35-l ' L-SYSTEMS ANSWER. '2.06 '('3.00) 1 a '. Non-interruptible-air; sys .-m (0.5)' -b, Drywell Nitrogen System (O. 5 )' c. fl.., Provide:a minimum of 5 cycles-(0.5) 2. . Hold the.~ valve open-30. minutes -(0. 5 ). .d. .i. Non--interruptible air header pressure low '( < 95 psig).- 2. LOCA" signal--.(from core spray) 3. Loss of power:to.the air-supply valves [any 2 @ 0.5 ea.] (1:0) l REFERENCE ~ i i BSEP: SSM, ' ADS,: P.10, Lesson Objectives 9 & 10 l218000K404

218000A203 218000A103-218000K604

..(KA's) . ANSWER' 2.07' (2.00) s a. True (0.5) 'b.

Falso (0.5) c.

True (0,5) i d. Falso

(0,5) l

-REFERENCE BSEP: SSM, ADS, PP. 8, 9 & 10,. Lesson Objectives 3, 5 & 7 239002A308 239002K504 239002K503 239002K405 ..(KA's) { \\ l;, l i (***** CATEGORY-2 CONTINUED ON NEXT PAGE *****) l l i ~l =__=:.____-__

I M.7,.... ; 2 PLAHT__ DESIGN INCLHD. LNG SAFETY AND EMEEGEHGY-Page'36 SYSTEMS 3 2.081 L(1 00)- 'I

ANSWER.-

f The CST could drain:to the suppression pool via-the'.' minimum flow-

line.

'(.1'.0)

REFERENCE

" BSEP: ' SSM,: HPCI;:P.16,-Lesson Objective'4 206000K417 ..(KA's)' 2.00) (

ANSWER 2 09

.1. Condensate-storage tank low ' level '2. Low reactor' water level 3. High'drywell pressure 4 '. HPCI suction valves from the suppression pool are open [4 0 0.5 ea.] (2.0) 4 , REFERENCE; ~BSEP: SSM,'HPCI, P.5, Lesson. Objectives S & S 206000K410 206000K419 ..(KA's) l i i l l -(***** CATEGORY 2 CONTINUED ON DEXT PAGE *****)

' ' l '. PLANT DESIGR INCLUDING SAFETY AND EMERGENCY Page 37 .EYSTEME_ i t i ANSWER 2110 (1.50) l 1. Low-Pressure Coolant Injection 1 2. Containment. Spray' l 3. Suppression Pool Cooling and Spray 4. Shutdown Cooling /2knd t ray c o m m,,,f' J,ja 4 5. Service Water Injection { 6. Fuel Pool' Cooling Assist [6 @.025 ea.] (1.50) .l ' REFERENCE BSEP: SSM, RHR, P.2 Lesson Objective 2 230000K101 226001K101 205000K104 203000K502 ..(KA's) I ANSWER 2.11 (2.50) Reactor Pressure less than 410 psig (/o conJua4((ou cJiN A (0.5) a. I c,co m o.,e d J. // LOCA SQM ) b. Prevent over pressurization of the low pressure piping upstream of the injection valves (0.5) .c. F015A/B: Remove initiation signal (0.5) F017A/B: S minutos after initiation signal is received (0.5) d. F017A/B-can be throttled to control injection flow (0.5) ) REFERENCE BSEP: SSM, RHR, P.22 Lesson Objective 10 203000K410 203000K402 ..(KA's) 1. (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l

?' s R *, ELAHI DESIGN -INCLUDING S&EETY ANILE.JERGENCY Page 38 EYS_TXt1S l ANSWER' 2.12 '(1.50) a. High side: Above-core plate pressure (0.5) 3 Low side: Core spray sparger' pressure (0.5) b. Negative (0.25) c. 'Yes (0.25) REFERENCE BSEP: SSM, Core Spray, P.8 Lesson Objective 9 209001K502 209001K404 209001K113 ..(KA's) ANSWER 2.13 -(1.00) i a. False (0.5) b. True (O'5) REFERENCE i BSEP: SSM, Core Spray, PP.10 & 11, Lesson Objectives 10 & 11 2090010008 209001K408 ..(KA's) ANSWER 2.14 (2.00) il i 1. Low Lube Oil Pressure (0.5] 27 psig +/- 1 [0.25] (0.75) i 2. Overspeed [0.5] 590 RPM [0.25] (0.75) 3. Loss of Diesel Generator Control Power y g e o efo /,f-f%// (0.5) D eWeren $a l DV!f tuffe nl ga tt o g/t, w,,p } p,, SAll&/loq r /En/t/.fr fm.)tr e (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) I

a

m. %

" J NE$L6HT 'DESIGJ._lB9LQ.QlBGjAEETNAND~ EMERGEH9X - 'Page.39- ?SYEIEd3 3 1

c. REFERENCE '

,d a k

1.,

-BSEP:l SSM, Dies'el' Generators', P.35,. Lesson Objective g. 1 E264000K4021 ..(KA's)^ -s E: ANSWER

2.15:

(1.00)- The? standby' position disables the automatic function of the SBGT system. (1.0) REFERENCE,, BSEP: SSM, SBGT, P.12. Lesson Objective e. 290001A102-290001A410 ..(KA's) p 1 tANSWER 2.16. .(2.50) a. The.RCIC gland exhauster system provides a means of preventing . radioactive-steam leakage.to the atmosphere (0,5) b. Barometric' Condenser Receiving Tank or Vacuum Tank Condensate Pump Vacuum Pump [4 @ 0.5 ea..] (2.0) i REFERENCE l t 'BSEP:. SSM, RCIC, P.8 Lesson Ob'jectives 1 & 8 217000G007 217000K405 .(KA's) i: (**3** END OF CATEGORY 2 *****)

p. 3 i

v. j.g 3 *. -INSIBUMEllI LAND Q0_HTROLS Page 40

. ANSWER 3.01 -(1.'50) 14- ' Actuate-SBGT deluge. valves 2. Deenergizes the SBGT. fan 3. Deenergizes the-SBGT heater [3 @ 0.5 ea.-]:(1.50) REFERENCE -BSEP: SSM, SBGT, P.16,. Lesson Objective f. .2900010007 . (KA's) i ANSWER-3.02 (0.50) False (0.5) ' REFERENCE BSEP: SSM, 230 KV Electrical, P.27 Lesson Objective 8 262001G008 ..(KA's) ANSWER 3.03 (1.50) [0.25] /~? 3 CPS [0.25] (0.5) a. Downscalo d om m en t f. o] b. Upscale [0.25] 1 X 10E05 CPS [0.25] (0,5) e Detector not full in [0.25] with less than 100 CPS [0.25] (0.5) REFERENCE l USEP: SSM, SRM'S, P.28 Lesson Objective g. I 215004K401 .(KA's) l-- l (***** CATEGORY 3 CONTINUED ON NEXT PA'3E *****) l

.. = .7 s " 1.. +

c. ;

'U Q b itiSIEUMENTSLAND CQNJBQLS: PageS41 d i f P. J-I i& l H. LANSWER 3iO4-(2.00): (The, normal supply: break'r must be open'beforeithe alternate canLbe Il e u .a. ' closed. . (0. 7 5 ). .The alternate supply. breaker:must-be:open.before the normal supply:-. breaker'can1be: closed. - ( 0. 75 )- 1 lb..To preven'tiparalleling of sources' .(0;.5)' l 1 REFERENCE _ 'BSEPi.SSM, 480 Volt Electrical, l?.'7 Lesson Objective 2 262001K406 262001K405 262001K403 .,(KA's) ANSWER' 3.05' (1.75) i a. .The:EHC system.has a. backup power supply from a PMG which:would continue to supply' power; [0.25) therefore, there would be no' effect on.the, turbine. [0. 25]. (0. 5 )' b. Reactor feedwater pump speed would decrease to minimum (for MGU). [0.25] Eventually a low level trip would result. [0.25] (0.5) I c. Removes the' ability.to move control rods [0.25] and a loss of- 'l the rod position dicplay' panel..[0.25] (0.5). 3

d.

Loss of power to the RWM would result in a rod block (0.25) REFERENCE BSEP:'SSM, UPS, P.18, Lesson Objective d. '262002K307 262002K306-262002K305 262002K301 202002K315 l ..(KA's) l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 'I l

y.... S3'L ' INSTRUMENTS'dND CONTROLS Page 42-2' L-. E 1 $ ANSWER '3. 06 ' (2.00):. i i True .a. 4 -bj False o c..- True' a 'd. False- [4 @ 0.5 ea.]~(2.0) REFERENCE BSEP:. SSM,. Instrumentation, P.6, Lesson Objectives 1 & 2 216000G007-216000K414 ..(KA's) ANSWER ' 3.07 (1,50) a. Opens.-the scram scram discharge volume vent and drain valves (0.5) ib. Shutdown-[0.25] or Refuel (0.25] (0.5) c.~ Rod Withdrawal Block (0.5) 'REFEFi,ENCE' -BSEP: SSM, CRD Hydraulics,.P.15, Lesson Objective e. '201001A105 201001K406 .(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) L

- }l,;. A. [.

I3NdNhTRUtfENTSI1N.lp CONTROLSI Page!43 'J-ANSWER. f3.08 (2.00)- a.; Afflow restricting orifice is-installed in the charging line. ( 0. 5~) ~ ,b. The. charging line,.o located between the flow sensing point-[0.5] and theLflow control valve. [0.5].When the increased flow signal is-seen by the flow control' valve it closes [0.5]' (1.5)" directingLfull flow,to the HCU'S, REFERENCE-j 'BSEP:-SSM, CR'D Hydraulica',.P.27, Lesson Objectives d & e '20100iK401 ..(KA's) 'ANSWERL 3,09~ '(3.00) a,

Low reactor water level' RWCU system area high temperature RWCU~ system high differential' flow RWCU system area' ventilation high dT

[4 @ 0.5 ea.] (2.0) b. ' Low Pressure: Between the restricting orifice [0.25] 0.25] (O.5) [ and FCV-F033. High Pressure: Between the FCV-F033 valve [0.25] and the radwaste system isolation valve F035. [0.25] (0.5) -REFERENCE l BSEP: SSM, RWCU, PP.16 & 43, Lesson Objectives b & e 204000K407 204000K404 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) \\ l

l._- r, I. TNS h &ENTS AND'CollTROLS Page 44' { I . ANSWER 3.10 (2.00) L .a..True b. Falso c.- False 'd. True- [4 @ 0.5 ea.] (2.0) .l REFERENCE i BSEP: SSM, APRM'S, PP. 11,.16 & 26, Lesson Objective e. d 21500rA104 215005K605 215005K407 215005K505 ..(KA's) ANSWER 3.11 (3.00) a. -Output of the individual pump controller [0.25] 89% [0.25] (0.5)

1-Limiter <28%

(0.25)

2 mimiter <45%

(0.25) d b. Limiter #1: Recire. pump discharge valve [0.25] >90% open [0.25] andfeedwaterflow[0.25]p20%[0.25] (1.0) dom,4 J,/ One or more Rea) tor feedwater pumps c [0.25] at <20% Limiter #2: rated flow-[0.25] and a reactor vessel [0.25] low level alarm (182") [0.25] is received. (1.0) REFERENCE \\ BSEP: SSM, Reactor Recirculation System, P.26, Lesson Objectives e. & j. 202002K402 202002K407 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) {

~,. ~ - ' } gi_ilt4 STRUM 5HIS_AHD: CONTBQLS - Page 45 I ' ANSWER '3.12 -(1.00) 1 -b; (1.0) ' - REFERENCE - .BSEP: SSM, EHC Electrical, P.9 Lesson Objective 9 - 241000K403' ..(KA's) ANSWF9 .3.13 (1.50) -l - 1. Mechanical-'overspeed trip device. 2. Mechanical trip-solenoid. 3. Manual mechanical trip. [3 @ 0.5 ea.] '(1.5) REFERENCE { j BSEP: SSM, EHC Mechanical, P.18, Lesson Objective 12. 241000K110 241000K406 .,(KA's) ) .) i sANSWER 3.14 (1.00) 1 I l a '. Increase b. Increase [2 @ 0.5 ea.] (1,0) i i I l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) j 4

~ f. dHETHLJMMIS bND COHIRQLS Page 46 i REFERENCE BSEP: SSM, Main Generator,.P.44, Lesson Objective h. 2620010007 262001A103 ..(KA's) ~ i . ANSWER 3.15 4-2.004 /s5 a. 90% (0.25) i g-((Fre&1 o.vj i b. ---h b uic by.nn n s vnivar g~ ii L6.u1 eiil purbine control valves shut [0.5] with a 30 second time delay (0.25] c. NO (0,5) o REFERENCE BGEP: SSM, RPS, PP.47 & 48, Lesson Objectives-c. & d. 4 215005A101 215005A104 ..(KA's) l ANSWER 3.16 (1.00) a. False (0.5) b. Falso (0.5) REFERENCE BSEP: SSM, CAC and Dilution, PP. 9 & 10, Lesson Objective 5 272000K118 272000K109 (KA's) (***** END OF CATEGORY 3 *****)

r~ W. LPaoCEDUREE_-iNORMAL....ABHQBMALi EMERGENCY Page 47-d 'JAND_BARIQLQGIRAL CONTRQL-1 l T -) ANSWER: 4,01~ (3.00)~ 'l ql ll Low-Ilow reactor. water level i 3 12. Main 1 steam ~line high-radiation 3. ' Reactor bldg.' steam tunnel high temperature 4.

Turbine bldg. steam tunnel high temperature l

5. Main steam line high-flow ~ <? > 6. Reactor pressure low.with mode switch in the RUN resitiUn

i 7.

Low' condenser vacuum l ~ -[any 6 @ 0.5 ea.): (3.0)' 3 . REFERENCE I 'BSEP: OP 25, Section;4.0.C, PP.7 & 8-239001K401 ..(KA's) i i ANSWER 4.02 -(1.00) l Reactor scram, because reactor power greater then 30% with the stop valves closed is censed by.155 psig first stage pressure. (1.0) REFERENCE l BSEP: OPL26, Section 5.2.B, P.17. SSM, RPS, P.17. '245000K104 ..(KA's) I i l L (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) i

p < w qc;7--- - -

3

. fy ,.s 14 i PROCEDURES' - MORMAL. ' AB@M6L L EtiEBQEEY Page 48 g

A@ RADIOLOGICAL CONTROLi t-

[.. < EAN WERi 4 iO3.- ._(1.00) c. T u_. V During a loss:of; coolant' accident and'the only remaining source of water Eto: flood theLreactorLis: service water. (1.0) . REFERENCE. 'BSEP: OP 43,LSection 4.0 A.', P 6 - 20'3000G007 ' ..(KA's) t

ANSWER; 4.04

. ( 2. 00 ). 11 -. 'The channel fasteners.are located at one corner of each bundle-adjacent to the center 1of the control rod. (0.5) 2. The identification boss of the bail points toward-the adjacent contro11 rod-(all. bosses point to center of the fuel cell). (0.5) 3.- The channel spacing buttons are adjacent to the control rod blades. (0,5) 31s 4. -There is cell-to-cell symmetry (except on the core periphery) (0.5) REFERENCE 5 Fuel Handling Procedure, FH-11, Section 4.27, P.4 .BSEP: BSEP::SSM, Fuel, P. 2, Lesson Objective 12 '234000K505 ..(KA's) o (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) 1

,r ,,,t

31. LPRQGEDUEEF -' HQPMALm.AHHQEMAL, EMEEGENCY Page 49 I

i AND_BADT%QGLQ6L CONT.BQL I j f N . o, 7J ' i h ~ ANSWER 4.05 (--i. 50 ) { \\ s rem / quarter rum / sear Whole Body 3 7.5 f+ Shin 3

a..u..

G110 Extremities 18.75 46 [fr @ -996 ea].( 1. 5 ) REFERENCE ] / BSEP: Radiation control and protection, Volume VIII, Section 4.1.1.1, i f P.14 fo CFR 20 .2 0< /61 / Md \\' ~ i 294001K103 ..'(KA's) i. A .I i ANSWER 4.06 (1.00) .i i If a technical specification and/or ESF system is not involved. (1.0) REFERENCE BSEP: Equipment Clearance Procedure, AI-56, Section 4.1.2.2,5.6' 294001K102 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l .1

n, 3 } o,.., 93 4 '. MNQGEREBEE r UQBMAL, APAGEti4L. EMEBQEERY Page 50 1 AHR_B&DlQkQG.1 GAL 3 QNIBQL a-k [ ANSWER 4.07' (2.00) s. F.a h o q fra.9 IQ False j t c.. False d. 'Truc [4 @ 0.5 ea.] (2.0) y REFEREt4CE .N BSEP:. operating Instruction 01-13, Sections, 4.0, 4.1, 4.1.1 & 4.1.4, P.2 4 7u.e/i-y Is.shu t9~ fileJA y/s//f) 294001M107 294001K101 ..(KA's) .t ' ANSWER 4.08 (1.00) o 'T "f' 1. Misor -ation in the " Automatic" mode is confirmed (0.5) 2. Adequate core cooling is assured (0.5) REFERENCE BSEP: AOP 03.0, Moderator Temperature Decrease, Section 4.1, P.4 295031K101 ..(KA's) l l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) i _________-_O

,y. 7 r h,,.,. e, ;. M: EBQQEDUBES NORMAL ABNORMAL EMERGENCY Page.51 ANQ_B&QLQLQQ1QAL CONTROL.

ANSWER 4.09 (2.00) 1; : Ensure the cause,is not a recirculation " Runback".

(0.5)- 2. ' Lock'the. scoop tube on the affected pumps (s), '(0.5) ,l ~ 3.- Monitor and Maintain reactor water level [0.5] between the the low and high' alarm setpoints.[0.5] ( 1. 0). REFERENCE BSEP: AOP 04.0, Recirculation Flow Control Failure - Decreasing Flow, Section 3.1, P.3 295001G005 ..(KA's) l . ANSWER '4.10 (1.00) It could cause neutron flux oscillations [0.5]:due to core thermal { hydraulic instability [0.5] (1.0) REFERENCE i i BSEP: AOP 04.1, Recirculation Flow Control Failure - Increasing Flow, Section 4.1, P.3 295001K102 ..(KA's) L l ANSWER 4.11 (0.50) 68 % (0.5) L (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) I u_____._

G ;".m ;; L } O w ' 2.- PBQ9EDUEED - NOEMALZ ABE9EMAL mEMEEGEHD1-Page 52 E' 4 ANH_BADIQL99194L_90BIERL-4 L 4 REFERENCE.

l l'

l 'DSEP:.AOP 04.3, Recirculation Pump Trip, Section 3.2, P.4 l n '295001K305 295001K103 ..(KA's) ,l

1 c

J ANSWER ~ 4.12 (1!00) -To slow cooldown of the reacter vessel bottom head. -(1.0) REFERENCE. 3 r+ \\ F BSEP:'AOP 04.3, Recirculation Pump Trip,.Section 3.2, P.4 r '295001K305 ..(KA's) ANSWER-4.13 (2.50) 1; Decrease.in generator megawatt output. l 2. Decrease 1n core thermal power. -3. Increase'in total core flow.

4.. ' Core plate differential pressure decrease.

5,- Recirculation loop flow increase in loop with failed jet pump. ~ [5 @ 0.5 ea.] (2.5) REFERENCE .BSEP AOP: 04.4, Jet Pump Failure, Section 2.0, P.3 202001G005 202001K601 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) .____-______-_-_________-________a

gr-p

,j u..',i L ;; " +

'1 =

T.i LE8QQEDURES - NORMAL. ABNORMAk J EBQKNCY Page 53 AND;BADIOLOGIGAL._G2NIBQL f

i 7 ANSWER ' 4.14, (3.00) a s 1". !ServiceLair. valves PV-706-1 & 706-2 close or Service airLheader' i 'J isolates. 1 2. Air compressors A, B, and-C start and load. 3. Instrument air valves PV-722-1 & 722-2 close or Interruptible-instrument air. header, isolates.. 4. ' Standby reactor 3 building air compressors start. l l 1 . Division I or II backup nitrogen supply. opens or RNA-SV-5482 & i 5. SV-5481Jopen. [5 @ 0.5 ea.] (2,.5) j b. Manually scram the reactor. (0.5); . REFERENCE- !j -l BSEP: AO'!-20.0, Air System Failures, Sections, 2.0 & 3.1.'S PP. 3&4 -} 2950'10G005 295019K303 295019K302 295019K301 ..(KA's) 1 ANSWER-4.15 (2 25) a. Cycle the affected SRV's control switch to OPEN and CLOSE or AUTO i several times. [0.5] Leave switch in CLOSE or AUTO.[0.5] -(1.0) ( ) b. As soon as it is recognized the SRV will not close. (0.5) -When Suppression Pool Temperature [0.5] Reaches 110 F. [0.25](0.75)

REFERENCE.

BSEP: AOP-30.0, Safety / Relief Valve Failure, Section 4.0, PP.3 & 4 239002G008 ..(KA's) L l L 1 I l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) i l

' 4 ', PBQQEDURES - NQRUAL ABNoBUAL, EMERGENCY Page 54 At.(D RARIQLQ.GICAL CONTROL h I I L ANSWER 4.16-(1.00) q l a. Reactor vessel level should be raised [0.25] to greater than 200 4 inches. [0.25] (0.5) i b. To ensure natural' circulation flow'through the core. (0.5) l i REFERENCE BSEP: OP-17, Residual Heat Removal System, Section 7.2.B, P.40, 205000A206 205000K303 ..(KA's) \\ i i i i i i i l l l l 1 (***** END OF CATEGORY 4 44***) l (********** END OF EXAMINATION **********) 1 l l l i ]

... *:.. ' t. ~ TEST CROSS REFERENCE Page 3' -QEESTION VALUE BEEEBENCE 11.01 1.00 ZZZ0000001 '1.02' O.50 ZZZ0000002 1.03 2.50 ZZZ0000003 1.04 1.00 ZZZ0000004 1.05 1.50 ZZZ0000005 l'.06 2.50 ZZZ0000006 1.07 2.00 'ZZZ0000007 '1.08 0.50 ZZZ0000008 1.09 2.00 ZZZ0000009 1.10 1.50 ZZZ0000010 1.11 2.00 .ZZZ0000011 -1.12 3.00 ZZZ0000012 1.13 1.00 ZZZ0000013 1.14 0.50 ZZZ0000014 1 15 2.50 ZZZ0000015 1.16 2.50 ZZZ0000016 26.50 i 2.01 1.00 ZZZ0000017 2.02 1.00 ZZZ0000018 2.03 2.00 ZZZ0000019 2.04 2.00 ZZZ0000020 2.05 1.50 ZZZ0000021 2,06 3.00 ZZZ0000022 2.07 2.00 ZZZ0000023 2.08 1.00 'ZZZ0000024 2.09 2.00 ZZZ0000025 2.10 1.50 ZZZ0000026 2.11 2.50 ZZZ0000027 2.12 1.50 ZZZ0000028 2.13 1.00 ZZZ0000029 2.00 ZZZ0000030 2.14 2.15 1.00 ZZZ0000031 2.16 2.50 ZZZ0000032 27.50 3.01 1.50 Z7ZO000033 3.02 0.50 Z130000034 3.03 1.50 ZZZ0000035 3.04 2.00 ZZZ0000036 3.05 1.75 ZZZ0000037 3.06 2.00 ZZZ0000038 3.07 1.50 ZZZ0000039 3.08 2.00 ZZZ0000040 3.09 3.00 ZZZ0000041 3.10 2.00 ZZZ0000042 3,11 3.00 ZZZ0000043 3.12 1.00 ZZZ0000044 3.13 1.50 ZZZ0000045 3,14 1.00 ZZZ0000046 3.15 2.00 ZZZ0000047 3.16 1.00 ZZZ0000048

.,.... v 4.01 3.00 ZZZ0000049 TEST CROSS REFERENCE .Page. 2 i QEEEIIRH YAldlE BEEEBEECE l 4.02 1.00 ZZZ0000050 4.03 1.00 ZZZ0000051 '4.04 2.00 ZZZ0000052. 4.05 1.50 ZZZ0000053 4.06~ 1.00 ZZZ0000054 4.07 2.00 ZZZ0000055 4.08 1.00 .ZZZ0000056 4.09 2.00 ZZZ0000057 i 4.10 1.00 ZZZ0000058 4.11 0,50 ZZZ0000059 4.12 1.00 ZZZ0000060 4.13 2.50 ZZZ0000061 4.14 3.00 ZZZ0000062 j 1 4.15 2.25 ZZZ0000063 4.16 1.00-ZZZ0000064 i 25.75 MWW--W 107.0 l l a

NAST!R CPY U. S. NUCLEAR REGULATORY COMMISSION 4 SENIOR REACTOR OPERATOR LICENSE EXAMINATION j FACILITY: BRUNSWICK 1&2 i REACTOR TYPE: BWR-GE4 1 DATE ADMINISTERED: 87/08/24 l EXAMINER: BISHOP. M. CANDIDATE INSTRUCTIONS TO CAHDIDATE: Use separate-paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%.- Examination papers will be picked up six (8) hours after the examination utartu. l % OF l CATEGORY % OF CANDIDATE'S CATEGORY I VALIIE_. TOTAL SCORE VALUE CATEGORY 2L.o 24.24 -efr:-50"g Mtg 5. THEORY OF NUCLEAR POWER PLANT 1 OPERATION, FLUIDS,AND l THERMODYNAMICS l /S.23g 27.00 -24.Ou 8. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

24. )G 28.00 25.01 7.

PROCEDURES - NORMAL, ABNORMAL, g EMERGENCY AND RADIOLOGICAL CONTROL 7k* 00 24,2 9 -27.00 24.0S 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS . lo 'l O d.^y.L 5 # Tota 1s Final Grade 1 All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature NASTB CPY )

.e v NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During tuo administration of this examination the following rules apply: l J 1 1. Cheating on the examination means an automatic denial of your application j and could result in more severe penalties. i 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination I room to avoid even the appearance or possibility of cheating. l I 3. Use black ink or dark pencil only to facilitate legible reproductions. l 4. Print your name in the blank provided on the cover sheet of the examination. 5. Fill in the date on the cover sheet of the examination (zf necessary). 6. Use only the paper provided for answers. 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet. 8. Consecutively number each answer sheet, write "End of Category _" as l appropriate, start each category on a new page, write only on one side l of the paper, and write "Last Page" on the last answer sheet. 9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk cr table.

l

12. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

15. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16. If parts of the examination are not clear as to intent, ask questions of the examiner only.

17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.

This must be done after the examination has been completed. j

. r: s.&n o L L 7F ] h18.c.When'you complete your' examination you shall: [ .J c'. .Assemblelyour1 examination:as'follows: 'i o u l L - (1); Exam questions'on' top. ~i tables,=etc. d (2)~ Exam' aids -: figures, 1 (3).: : Answer:pages including figures:which are part of theLanswer. b; . Turn in'your copy of the examination and all pages used-to answer- ) "the examination. questions.- 'I J Ec. LTurniin?allLacrap paper and.the balance of the paper.that:you did-j nots _use"for answering the. questions. d.: Leave the examination area, as defined by the examiner. If after. i l leaving, you1areLfound-in this area while the examination.is still l ~ 'in: progress,fyour license;maysbe denied or revoked.. i I i 'I. s i i l l l l l

1 M c Ge:- y y50 UTHEORY OF NUCLEAR POWER PLANT OPERATION. ~ Paga 4-FLUIDS.AND THERMODYNAMICS j , -. s y..

QUESTION ~

5.01 (2.50) I

Answer EACH of the following in'regards to THERMAL. LIMITS':-

u

e. What are THREE (3) parameters required by the process computer E'

for the-MCPR calculation?-

(0.75) .b. What-is the relationship between MAPRATJand MAPLHGR7 ( O'. 5 ) .c.' Is'a MAPRAT of 1.05 within operating-limits?- '(0,5) l d.:What-could be the physical consequence ~of. operation outside

the MAPRAT limit?,

(0.75)

QUESTION 5.02

-(0.50)

Which ONE (1),-of the following radiation doses'would have the WORST Abiological effect on a' human?

a. 1 Rad'. of-GAMMA b. 1 Rad.'of BETA c. 1 Rad. of NEUTRON d. 1 Rad, of ALPHA (Internal) { l 1 ) i (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

.c 5, THEORY'OF NUCLEAR POWER PLANT OPERATION, Perp '5 1 FLUIDS.AND THERMODYNAMICS i i l a 1 QUESTION 5.03 (3.00) { i For EACH of the pairs of conditions listed below, state WHICH condition j would have the GREATER rod worth and BRIEFLY EXPLAIN WHY, { a. ' Reactor moderator temperature of 150 deg F or 500 deg F. i 1 b. For an inserted rod.next to a fully withdrawn control rod or next to a fully inserted control rod. [ Assume average core flux is constant) c. For a rod at position 10 or position 40 of a core operating. at 100% power. l l QUESTION 5.04 (1.00) A reactor heat balance was performed (by hand) during the 0000-0800 shift due to the Process Computer being 000. The GAF's were computed, but the APRM GAIN ADJUSTMENTS HAVE NOT DEEN MADE. l 1 Which ONE (1) of the following statements is TRUE concerning reactor power 7 a. If the feedwater flow rate used in the heat balance calcu-lation was LOWER than the actual feedwater flow rate, then the actual power is HIGHER than the currently calculated power, b. If-the reactor recirculation pump heat input used in the heat balance calculation was OMITTED, then the actual power is HIGHER than the currently calculated power. 1

c. If the steam flow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is HIGHER than the currently calculated power.

d. If the RWCU return temperature used in the heat balance cal-culation was LOWER than the actual RWCU return temperature, then the actual power is HIGHER than the currently calculated power. (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

.. i _.,., ; ', (

.i.

!5; THEORY OF' NUCLEAR' POWER PLANT OPERATION. Pcss 6 E. .FkUIDS.4ND THERMODYNAMICS (: QUESTION . 5 05: .(1.50) " STATE whether the following thermodynamic' properties' INCREASE, j DECREASE, or, REMAIN.THE SAME as steam travels from the inlet to the outlet of the BSEP H.P. Turbine. I c. Enthalpy b. Entropy c. Quality-l QUESTION 5.06 (1.00) If' reactor power'is 1,Kw and control rods are withdrawn to establish a 100 sec. period WHAT will' reactor' power.be after 7-minutes? Assume no operator action after the stable period'is established-and the 7 minutes starts at the time the stable' period is established. Express your answer in MEGAWATTS. i 1 t 4 ) l I i s i 1 t-i l l L J L j (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ________________________-____D

[.,' 2

l. I 5.

THE'ORY'OF NUCLEAR-POWER PLANT OPERATION. Paco 7-l- FLUIDS,AND THERMODYNAMICS I- {> l c QUESTION '5;07 (2~00) l During your Shift,.an SRV inadvertently opens from 100% power and 1000 psia. Use a Hollier Diagram or the Steam Tables to anewer EACH of the following: (ASSUME A SATURATED SYSTEM.AND INSTANTANEOUS HEAT TRANSFER) al STATE the tailpipe temperature',' assuming atmospheric' pressure in the Suppression Pool and No Reactor Depressurization. .I 1 b. If the'. Suppression Pool Pressure were to INCREASE, STATE whether the Tailpipe Temperature would INCREASE, DECREASE,.or REMAIN THE -SAME.' I .c. .IfLthe reactor starts to depressurize'whe. the SRV opens, -STATE whether the Ta11 pipe Temperature.will initially INCREASE, DECREASE, or. REMAIN THE SAME in relation to what it would have done iffthe pressure had not decreased. d.- STATE the Reactor Pressure at which the Tailpipe Temperature would be at'its MAXIMUM value (during the depressurization). f QUESTION 5.08 (2.50) Referring to attached Figure 1, " Closure of All MSIV with Valve Position Scram" for Unit 1, answer EACH of the following;

a. EXPLAIN WHY neutron flux increases sharply at approximately two seconds into the event (Point A). (A complete response will include _both the action which causes the spike and the mechanism j

which causes the power increase.) (1.0) j

b. WHY.has vessel steam flow returned to approximately 90% of rated at Point B7 (0.5)

c. WHY.is reactor pressure increasing at Point-C (eight seconds into event)?

(0.5) l

d. WHY is reactor water level increasing at Point D even though a steam flow is much higher than feed flow?

(0,5) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) I I

e l' s r* 5. THEORY OF NUCLEAR POWER PLANT OPERATION, Paco 8 FLUIDS;AND THERMODYNAMICS i I I l ( /,50) QUESTION 5.09 E2M Answer EACH of the following by referring to the attached Figure 2, Unit 2 Recirculation. Flow Control Failure - Increasing Flow; a. Briefly' EXPLAIN WHY the neutron flux increases at Point A when the MG set fluid coupler causes the Recirculation pump speed to increase? (Include discussion of reactivity coefficients) (1.0) b. WHY does indicated reactor water level decrease ut Point B?. (0.5) c. WHY is this accident considered most severe if initial reactor power is 65% and initial core flow is about 50%? (0,5) f)olo/h J QUESTION 5.10 (1.50) Unit 1 power is increased by control rod withdrawal. The void fraction increases 1.5% and the fuel temperature increases 40 degrees as the result of the rod withdrawal. What was the reactivity WORTH of the PORTION of the control rod that was withdrawn ? SHOW ALL WORK AND STATE ALL ASSUMPTIONS. QUESTION 5.11 (2.00) The reactor has been operating at 95% power for several days. The operator rapidly reduces reactor power to 60% by reducing the speed of the recirc pumps. During the next few minutes ( 2-3 minutes) the operator notices reactor pcwer slowly increases approximately 3%. BRIEFLY EXPLAIN the cause c.? this effect. 1 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) { i

Wmn kf,, '. g' 3, gp. ,l52 THEORY'OF'NCCLEAR POWER PLANT OPERATION, Pega 9. FLUIDS.AND..THERHQDYNAMICS ^ .i 0 . QUESTION 15.12 ' (3.00) 1 .c. ListiTHREE.(3)iparameters which< contribute--to:AVAILABLE NPSH (Net Positive Suction Head)-for a recirculation pump. Limit your? answer toLthose. parameters which are DIRECTLY indicated

in the. CONTROL-ROOM.-

(1.50) l b. . Consider TWO'(2) Reactor Plant conditions: Low' Power'and Low Flow (<10%) OR ~- High Power.and High Flow'(>85%). -1.

During:which' condition is the REQUIRED NPSH for a Recirculation ~ pump greater?

(0. 50 )

2. 'During which condition is AVAILABLE NPSH-for a recirculation' ' pump l greater and WHY is it greater? (1.00)- i. ' QUESTION 5.13 . ( 1.' 00 ) EUsing the steam tables, CALCULATE a reactor cooldown rate (dog.F/hr) for: a reactor pressure decrease from'1000 psig ' to 250 psig' ir. 'one hour and forty five mir.utes (105 minutes total). SHOW ALL WORK.' LQUESTION 5.14 (0.50) 'The void coefficient of reactivity becomes less negative for which ONE of the following changes? a. Percent void changes from 30 to 40. b. Average fuel temperature changes from 500 deg.F to 550.deg.F. c. Core age' changes from BOL to EOL. p d. Control rod density changes.from 20% to 25%. l l-j: e. Reactor pressure changes from 980 peig to 1010 psig. l l l-(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) i i L

7., 4,.; - - - p s ~ lh__1 .is* ] i "62 THEORY-:0F NUCLEAR' POWER PLANT'OPERATIONi Pzsa ' m; FLUIDS AND_ THERMODYNAMICS u o e o fQUESTION 5.15. (2.50)- J a. Define the term BETA with regard to delayed' neutrons? (1.0)- .j -b. When' comparing the;ihdividual BETA from thermal fission'of < U-235, ePu-239,L 'and f ast fission of U-238, which BETA is .l largest?.. (0.5)-

c. From BOL - to EOI,, does the core average beta INCREASE,.

DECREASE,'or REMAIN THE.SAME7 EXPLAIN your answer.. (1.0) 1 i i l l, p-t ![ (***** END OF CATEGORY 5 *****) l

~J;l V; d4 ,pk

6[ PLANT SYSTEMS: DESIGN, CONTROL,JAND INSTRUMENTATION' Pass ll' 7

j + . QUESTION".6'01:

(2.00)-

^ The'RBM-gainichange'is~done so the RBM output will be equal.to or EgreaterJthan the reference'APRM output. WHAT;are1-TWO (2)' reasons;for changing the gain? i QUESTION 6.02 (2.25) r

The HPCI minimumiflow valve.is designed to automatically close under-F three conditions..

1 WHAT are these THREE (3) conditions and WHAT is the basis for each of H

the conditions?

QUESTION' 6,03 ('2.'00) i Answer'the.following in regard to the RHR system: .l I a.JWhat is the BASIS for the 2.7 psig DRYWELL pressure interlock in the containment cooling logic? (1.5) b 'What.is'the BASIS for the three minute interlock in the heat i exchanger bypass valve logic? (0.5) i i l l l 1l i ) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) 1 L______=__ ~

m a 0.c m. ii.' PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pega 12-4 'l QUESTION 6.04 (2.50). ' Answer EACH of.the following in. regard to the core spray system:- a; WHY are' operators: cautioned-about shutting d.own a core spray-pump'with an' initiation signal present or bsfore the initiation signal has been reset? -(1.5) b.' The Core Spray _ Inboard (F005) and Outboard (F004) valves have interlocks associated with'each other's position during manual operation utilizing the hand switch. What are these TWO (2) interlocks? (SETPOINTS REQUIRED) (1.0) . QUESTION 6.05 (3.00). Answer EACH of the following in regard to the Control Building HVAC System: .a. WHAT design feature of the discharge check valves allows them to close in the' event of a tornado? (0.5) b '..What TWO'(2) signals will automatically shutdown the Emergency Air Filtering' Trains if they are operating with an automatic start signal present? (Setpoints not-required.) (1.0) I o. List THREE (3) conditions necessary for the make-up air dampers to be in their normal positions. (1.5) I J QUESTION 6.06 (1.50) Temperature switch TS7 is located in the inlet duct of the SBGT filter train. What THREE (3) automatic actions are defeated when .this temperature switch sees inlet temp increase to 180 deg. F.? l (Setpoints not required) l~ t 1 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) n

5" D.: .6 l l L i /6p PLANT SYSTEMS DESIGN,L CONTROL, AND INSTRUMENTATION-Pcca.13 L E 4 11 } -l QUESTION 6.07-(3.00). 1 1; Answer EACH of.the following concerning the RCIC system.. ) I

a. :What are TWO-(2) reasons the Inboard steam' isolation valve (F007)Lhas an AC; operator instead of a DC operator?

(1.0) b. List FOUR-(4) trips that will result in the Trip and Throttle-valve closing. .(2.0) 'QUE'STION 6.08-(3.00) Answer EACH of'the followP4g in. regard to the Reactor Vesse1~(RPV): .a.?Briefly' EXPLAIN HOW axial stability and lateral uupport are f provided.to the RPV7

b. What are THREE functions performed by the BIOLOGICAL SHIELD 7 I

c. The Shroud Head is tensioned to 50 ft. lbs, of torque when it is installed. This is only a portion of the tension required at full power.

Briefly EXPLAIN HOW the remainder of the required tension is applied and WHY this design was utilized, l QUESTION-6.09 (1.50) 1 i What are the THREE (3) ways or methods by which the Turbine EHC Hechanical Trip Valve is or can be tripped? ] l l I 1 l QUESTION-6.10 (1.50) List THREE (3) of the four design bases of the Condenser Circulating l Water System. l 4 l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) L J

l' .i 6, PLANT' SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Peca 14 i I l 1 QUESTION 6.11 (2.00) Answer EACH of the following regarding the Service Water System : a. List THREE (3) loads supplied by the Vital Service Water System Division II Header. (1.5) b. WHAT signal automatically initiates service water flow to the RHR pump seal coolers ? (0.5) i ' QUESTION 6.12 (1.25) With tne TBCCW system in normal operation, the system temperature starts .to increase. BRIEFLY EXPLAIN what automatic actions occur within the system tx) keep the temperature in the normal range. INCLUDE in your answer any applicable setpoints, the normal temperature range, and the high temperature setpoint. QUESTION 6.13 (0.50) I State whether the following statement is TRUE or FALSE; RBCCW to the Drywell does not automatically isolate in the event of a LOCA signal. i QUESTION 6.14 (1.00) l As a result of the design of the RPS system, what TWO (2) conditions would result in a portion of the control rods scramming with no valid scram signal present 7 (***** END OF CATEGORY 6 *****) L_ _ ___ __

f' : ' s,f, ]

f. 7,;

.i L 07:. ' PROCEDURES'- NORMALT ABNORMAL,' EMERGENCY Pap 15 ,-AND RADIOLOGICAL CONTROL- .J \\ i trl fQUESTION:.7.01: . ( 2. 00 )i j . List FOURc(4)' entry conditions for Emergency Operating Procedure E0P-01 CCP,.ATWS-CONTAINMENT CONTROL PROCEDURE. (SETPOINTS REQUIRED.)'

QUESTION 7.02' (2.50)

]

a. Define the term " Boron Injection Initiation Temperature" per

EOP-1-UG,' USER'S GUIDE.'

(1.0) i b.~WHY;has a " Boron. Injection Initiation Temperature" been H established? (1.0)- .c. What is the VALUE (deg.F) of'this' limit? (0. 5 ). .QUESTIONL 7.03 (1.50)- List the THREE-(3) entry conditions for E0P-01-EDP, EMERGENCY q DEPRESSURIZATION PROCEDURE 7-QUESTION 7.04 (2.00) l According t'o AOP-10.0, Moisture Separator / Reheater Tube Failure, WHAT 1 are FOUR (4) s3rmptoms or indications of a failure in the reheater j tubes? j i QUESTION 7.05 (2.00) l According to AOP-12, Loss of Uninterruptible Power Supply (UPS), WHAT are FOUR (4) symptoms or indications of a loss of the UPS bus? (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) u_ii__:__ __ _ m I

_ _ _ = _ _ _ y m.m 1 7l. PROCEDURES - NORMAL ABNORMAL. EMERGENCY Pcg2 16 AND RADIOLOGICAL CONTROL j fQUESTION 7.06 -(1.50) =According to AOP-26,0, Bigh Reactor Coolant or Condensate Conductivity, q WHAT are THREE-(3) symptoms or indications which could be present.if 1 resin injection is the problem? j I 1j I ' QUESTION 7.07 (0.50)' l -{ . 1 AOP-32.0, Plant' Shutdown From Outside the Control Room cautions'the op'erator.not to increase steam flow to 3 x 10E 6 lb/hr while reducing 1 Unit 2 reactor pressure to approximately 700 psig. WHAT is.the BASIS for this caution? i . QUESTION' 7.08 (2.00) i i What are FOUR (4) of the six IMMEDIATE ACTIONS that should be carried out prior to reducing pressure to approximately 700 psig -while executing AOP 32.0, Plant Shutdown From Outside Control Room? 3 1 l l 1 I l l l 1 l l l 1 l t i J (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) i i l l. \\ a

c <c 1.,.. :n co -t t i 7-PROCEDURES -NORMAL.' ABNORMAL'. EMERGENCY Pegs 17 o AND RADIOLOGICAL CONTROL,

5.
QUESTION-7.09 (2.50)~.

Answer EACH:of the following with regard to the~ primary containment: a. iDuring a~ reactor plant startup. when must the. Oxygen concentration"be less:than 4% 7 (1.0) b. Upon increasing temperature of the Suppression Pool, STATE the~ temperature at which a Technical Specification Limiting ~ Condition'.for Operation in FIRST entered. (0.5)- c.. The~ reactor shall be scrammed if suppression pool temperature. reaches (0,5) d. During reactorcisolation conditions, the reactor pressure vessel i shall be depressurized to less than 200 psig at normal cooldown rates if the' Suppression Pool temperature reaches (0.5) -QUESTION 7.10 -(2.50) i a. STATE the exposure rate limits (for a major portion of the body) which characterize EACH of the following: (1.5) 1. Radiation-Area 2. High Radiation Area 3. Locked High Radiation Area b. ' STATE the definition of EXTREMITIES as it pertains to radiation exposure of personnel. (1.0) i . QUESTION 7.11 (1.50) I What TWO (2) operational limits are recommended for the Recire pumps by 1 the Unit 2 Operating Procedure 02, Reactor Reciro. System, in the event j of a sudden increase in the No. 2 Seal Cavity Pressure ? I 1 I (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l J L i ~

-e v 7 ' PROCEDURES - NORMAL.' ABNORMAL, EMERGENCY P2gs 18 AND. RADIOLOGICAL CONTROL QUESTIDN 7.12 (1.00) -Unit 2 Operating Procedure 05, Standby Liquid Control System, cautions the operator to maintain the SLC pump discharge accumulator nitrogen pressure between 450 and 500 psig. WHAT is the BASIS for this requirement? QUESTION 7.13 (1.00) Unit 2 Operating Procedure 07, Reactor Manual Control System cautions the operator not to withdraw rods with core flow greater than 60%. WHAT is the BASIS for this caution? I QUESTION 7.14 (1.00) Operating Procedure 08, Control Rod Drive Hydraulic System cautions the operator not to let the CRDHS flow rate become less than 30 epm. WHAT is the BASIS for this limitation? QUESTION 7.15 (2.50) I In accordance with General Plant Operating Procedure 01, STARTUP CHECKLIST, there are five requirements that must be satisfied in order for a system to be in " Standby Readiness". ) LIST these FIVE (5) requirements. 1 l l l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) 1 1

7- .,r ' i 7.. PROCEDURES - NORMAL, ABNORMAL,' EMERGENCY Pcc 19 t r AND RADIOLOGICAL CONTROL ' Ia i --QUESTION 7.'16 . (2.00) ~ ! 11 List FOUR (4)' separate visual indications used to confirm correct orientation of a fuel bundle placed-in'the core. 1 l J l L i . i., j 1 l 1 I I i i l l l (***** END OF CATEGORY 7 *****) L- - --

(T. ' i .s 8.. ADMINISTRATIVE PROCEDURES. CONDITIONS. P;;ca 20 AND LIMITAT10NS QUESTION 8.01 (2,00) i Answer EACH of the following in regard to the Emergency Notification System, ENS (Red Phone): V If the ENS is not operable when it is needed, LIST (In orcif?the

a. THREE (3) alternate methods available for NRC notification.

(1.5) i i

b. WHAT BSEP document specifies the NRC alternate notification methods?

(0.5) I ( Exact title or number not required.) (3: 00) QUESTION 8.02 (+.-O&) Answer EACH of the following questions concerning AI-58, Equipment Clearance procedure. a. There are three major groups of clearances -- station, radwaste, and local. WHO may AUTHORIZE the cancellation of EACH of the THREE (3) clearance groups? (1.5i b. State the COLOR and under what conditions (WHEN) the following tags will be used. 1. Paper Caution Tag (0.5) 2. Plastic Caution Tag (0.5) 3. Caution Label (0.5) c. What does a RED clearance plastic CAP placed on the following control switches on the RTGB (Reactor Turbine Generator Board) mean. 1. Valve Control Switch (0.5) 2. Breaker Control Switch (0.5) [h l l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) 1 1

pq --

^

8 ADMINISTRATIVE PROCEDURES, CONDITIONS Pcco 21 s AND LIMITATIONS -QUESTION 8.03 (0.50) l In accordance with the PEP's (Plant Emergency Plans), STATE the 1 ONE RESPONSIBILITY the Site Emergency Coordinator SHALL NOT delegate until the position of Emergency Response Manager (or EOF) is i activated. j QUESTION 8.04 (1.00) Which ONE (1) of the following correctly describes the Technical Specification definition of an " Instrument Functional Test?" f a. The adjustment of an instrument signal cuput so that it corresponds, within acceptable range and accuracy, to a known value of the parameter which the instrument monitors. f b. The injection of a simulated signal into the instrument's primary sensor to verify the proper instrument channel response, alarm, and/or initiating action. i c. The qualitative determination of acceptable operability by observation of instrument behavior during operation, including, where possible, comparison of the instrument with other independent instruments measuring the same variable. ) d. A tost of all relays and contacts of a logic circuit to 1 insure all components and instruments are operable per the j design intent. QUESTION 8.05 (2.00) l L LIST FOUR (4) conditions that require a Radiation Work Permit in accordance with the E&RC Manual procedure Radiation Control and { Protection. (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) i )

.. s !- ADMINISTRATIVE PROCEDURES. CONDITIONS. Paca 22 1 l81 AND LIMITATIONS-9 -1 QUESTION 8.06 (2.50) l Answer.EACH of the.following according.to Operating Instruction 01-01, Operating Principles'and Philosophy: 4 ~ a. EXPLAIN what'"Believe Your Indication" means. (1'.0) -1. WHO can approve the transfer of a "Part/ Item" from one unit or l 6 system to the other unit or another system? Include the l requirements.foriboth " Safety Related" systems and "Other Equipment" transfers. (1.5) i QUESTION 8.07: (0.50) l t Operating. Procedure OI-02, Shift Turnover Checklist recommends an ~ Auxiliary Operator accompany the Shift Foreman and Shift Operating Supervisor when~they make a plant tour. What: is the purpose of this recommendation? i QUESTION 8.08 (3.00) Answer EACH of the following in accordance with Operating Instruction

OI-04,.LCO Evaluation and Followup:

a.' Define " Tracking LCO". (1.5)

b. TRuni is an Event Evaluation Check Sheet required for equipment removed from service for testing required by Technical Specifications and conducted per approved plant' procedures?

(0.5) .c. What Time and Date should be assigned to an LCO that is discovered by the Shift Foreman during shift --change review of WR/JO's? (0,5)

d. Uhen Technical Specification 3.0.3 (COPY ATTACHED) is entered, when does the 6 hour clock start?

(0.5) (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) L

J 8..

ADMINISTRATIVE PROCEDURES, CONDITIONS, Pega 23

AND LIMITATIONS QUESTION 8.09 (1.50)

AccordingLto Operating Instruction 01-13, Valve and Electrical' Lineup Verification, Valve Identification, and Locked Valve Identification and Locking,.WHAT are THREE-(3). exceptions to the requirement that'"All Valves-and Breakers Will Be Maintained In:the Position Required For the 'OP Valve / Electrical Lineup"? . QUESTION 8.10 (2.00) In'accordance with Administrative Procedure, AP, LIST FOUR (4) of the-six11tems evaluated to determine if a proposed Temporary Change to a procedure changes the INTENT of the procedure.

QUESTION 8.11 (2.00)

Answer EACH of the following in accordance with Administrative Instruction AI-59,Jumpering, Wire Removal, and Designated > meer: a. WHEN are Jumper and Wire Removal Tags not required to be utilized? (TWO REQUIRED FOR FULL CREDIT) (1.0) b. WHO can authorize jumper installation or wire removal in a safety-related. system that is inoperable? (0.5) c. WHO is responsible for maintaining the Safety-Related Jumper and Wire Removal Log? (0.5) QUESTION 8.12 (1.00) l Unit 2 Technical Specification 3.9.3 requires all control rods be fully i inserted in Condition 5. I WHAT is the BASIS for this requirement? (TWO REQUIRED) i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

+ [83 ADMINISTRATIVE PROCEDURES. - CONDITIONS. ' Pego!24 ,AND.. LIMITATIONS ] i ,y . y- ' h' 'iQUESTION ' 8.13 (2.50); WhatDare~FIVE (5) items / activities that require a: SAFETY EVALUATION;.

in:laccordance with the Technical Specifications?

m } QUESTION l-8.14- '(2.50) 'In accordance with'the Technica1'Specifica ions, the.-reactor N Ewas scrammed due to Suppression Chamber water temperature being. greater'than 110--degrees F. The. reactor is now in HOT SHUTDOWN, Suppression Pool' Cooling is.ON,'all heat input to the Suppression Chamber.has been stopped, and Suppression Chamber. water temperature l is 97 degrees'F. CAN YOU STARTUP THE REACTOR AND ENTER i OPERATIONAL CONDITION 27 EXPLAIN YOUR ANSWER. FULLY. ( 2. 5). I i i l l i I l (***** END OF CATEGORY. 8 *****) (********** END OF EXAMINATION **********) j

^ 4S ) )P)B I ) S (E P( S (PU s 2 I \\ N4 I iI I ET(O E R. SBCL s I AF iSR (B R RE .R .UFM PNPEVT I Ml W SP SSOA 3 E EERTOE 8 L S LUENE / ENINIA SABE 6 ) SERR ETUO VSTC ( I23436 2 6 5W Q O N L E F 5 S ( E E R M O 2 IT 0C 4 5 ~ ( N XULF ^ XT \\6 UA LE 5 i x 2 FH ,2 N ^ e ^ OE ^ RA ^ TF UR 1 M i i -~ I V EU I A NS SR1 i f i2 ~ 5 n 5 0 g s 0 MC E O S 2 2 L L 2 1 ^0 4r

g LNC 1

e AOY l)S FT C i m I P. TS p X REE MU MHt OI I S E EL CO) TF W S N% EO1 R N RT O P (I RP T U .D%5 I() L(Lw% UEI G E EA L S TE F E EO S V N NHW FVVL(I I X M EEEF OL U F UM OLA 5 RLL E TW L A L FE 2 4 FLC T HDDEO a C EA N UFRS 8 CEELL V W NSS FRE NF 3 NNI I O ULTL (EE E R SWE LESSE RI ~ TK S V UAEEDS VRROR .EEVEE ELWNCD NPAFV I2345 r' 2345 I ? ) g) C 3 C 4 5 M g ES ( E E 4 MI T I T A y 4 s 4 5,(' \\ I g

?3 5d o'

/ a = 5 = t 2 a = = =U S 0 O ,O C C 5 S 1 $g E I yI Nt 1 llC r

)SI SP S P{P - -W (E(CCO s ESEPPL I S F SRI 2 2 l I M 3 3 l R.RS.WWA S.ESOOE ERE LLT 0 RPRFF S 6 0 P PE RR 1 8 E LNEEE N SS I 1 EL9 UU8 t S 8 SMR FFR )% FF U ETU0 I VS10D T ( 2 5 I234S6 7 W 3N lqI J i O 8 j ) 4 L F L E L 5 C X R U 0 O 4 M F LE MT UA FH 4' . NE OC 5 4 RA 2 a TF UR / 3 EU = I 6 i i. NS 2 = I2 n ~ e. 'E.O O 0 O O O. 0 O 8 O-2 e 4 6 e, e_ 8 8 ) sWt:0t E T ) i R S i P. E I M M H C M U S C P P. EL N 1 TF W P I - N 3 i MM E ( RT O [ %W iSLLW O L EAW EEO I TE F T NH O VVL 4 5 t EEEF s X LM E R LL O UCE F A T L C FLA R E . HDD P E 6 %N 1 T 6 C EE L R E EF S QY'1 INSS H E NUR T A L NNI OFU ( LEE L R SWE ESS E P TM.DS Vt R U UAEE S 2 RO O EEVEE E . LWNC C MPAFV 4 2 I234S \\ 234 5 2 I 1 1 4 3 5 I /' M

a. T 8

2'hN w 5 4 I' 4 i y 4 j s I' 's h f" r sO 2 [u ~ 7, ~. f O 0 0 O 0 0 o O 0 g 0 s 5 0 5 5 1 1 8 8>2 o 5a. ~ 3 p $]

(- 3/4 LIMITINO CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION Limiting Conditions for Operation and ACTION requirements shall be 1.0.1 ' appitcable during the OPERATIONAL CONDITIONS or other states specified for each specification. Adherence to the requirements of the Limiting Condition for Operation I 1 0.2 and associated ACTION within the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, completion of the ACTION statement is ont required. In the event a Limiting Condition for Operation and/or associated 1.013 be satisfied because of circumstances in excess of ACTION requirements cannot those addressed in the specification, the unit shall be placed in at least HOT 9HUTDOWN within 6 hours and in COLD SHUTDOWN wichin the following 30 hours unless corrective measures are completed that pe rmit operation under the permissible ACTION statements for the specified time interval as measured from initial discovery or until the reactor is placed in an OPERATIONAL CONDITION to which the speciftention is not applicable. Exceptions to these requirements shall he 4tated in the individual specifications.

PHATI"!E ?"T!T!nN or other scecified applicability 3.1. 4 Entrv at t o an state shall not he ?.ade unless the conditions of the, Limiting Condition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provision shall not prevent passage through OPERATIONAL CONDITIONS required to comply with ACTION requirements.

s. 1.A.5 When a system, subsystem, train, ecmponent, or device is determined to be inonerable solely because its emergency power source is inoperable, or solely because its normal power source is ineparable, it may be considered apERABLE for the purpose of satisfying the requirements of its applicable 'Jmiting Candition for Goeracion, provided: (1) its corresponding normal or amergency power source is CPERA3LE; and (2) all of its redundant system (s), subsystems (s), train (s), component (s), and device (s) are OPERABLE, or likewise satisf y the requirements of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least HOT SHUTDORN within 6 hours, and in at least COLD SHUTDOWN within the following 30 hours. This specification is not applicable in Conditions 4 or 5. i l l i I .7 BRUNSWICK - UNIT 2 3/4 0-1 RETYPED TECH. SPECS. 'g Updated Thru. Amend. 7 l i N ( 1 \\ ) E_ _

l

e ; s' ' CONTAINMENT SYSTEMS s: = 3/4.6.2 DEPRESSURIZATION SYSTEMS SUPPRESSION CHAMBER . LIMITING' CONDITION FOR OPERATION 3.6.2.1 The suppression chamber shall be OPERABLE with: a. The pool water: 1. ' Volume' between 37,600 f t aad 89,600 ft3, equivalent to a level 3 between'-27 inches and -31 inches, and a y( 2. Maximum average temperature of.95'F during OPERATIONAL CONDITION 1 or 2, except that the maximum average temperature I may be permit:ed to increase to: a)' 105'F during testing which adds heat to the suppression - i chamber. 5) 110*? with !ERMAL POWER less than or acual-to l'. of RATED T222:iAL 20'iER. ? 41 : /. :a= cain 3:22m '.ina ;3cla:i:n valves closed 6 fol.:wia; 1 scram. a b. Two OPERA 3LE suppression chancer water :emperature instrumentation j channcis ui:h a minimum of !! cpersbie ITD inputs per channel. c. A 3 cal '.eakage from the drywell to the suppression chamber of.less

han :he equivalent leakage through a 1-inch diameter orifice at a differential pressure of I psig.

( 4 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. f i ACTION: i With the suppression chamber water level outside the above limits, 1 a. ras:Or2 the water level to within the limits within 6 hours or be in 4: least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN wl nin the following 24 hours. l. In OPERATIONAL CONDITION 1 or 2 with the suppression chamber average water temperature greater than 95'F, restore ene average temperature

han or equal to 95'F witnin 24 hours or be in at leas t ' OT d

1 o.ess

ZJ'OvWN sithin the next 12 hours and in COLD SHUT 00WN wi:hin the

'. acurs, excep:, as ;ermi::aa ace"e:

....n; 3RUNSWICX - UNIT 2 3/; 5-9 Amendment No. 103 l

i

- 1 l .o' . C l CONTAINMENT SYSTEMS ~ ~ LIMITING CONDIT.J.0NS FOR OPERATION (Continued) ACTION: (Continued) - 1 1. .With the suppr'essio'n c' hamber average water camperature greater than 105'? during :esting which.' adds heat to the suppression. . chamber, 3:ap all testing which' adds heac to.the suppression chamber 2nd restore the average temperature to less than or equal to 95'? 'si:hin 24 hours or be in at least HOT SHUTDOWN "I within the next 12 hours and in COLD SHUTDOWN within the following 24 f.o u r s. 2 '. LWith the suppression chamber average water te=perature greater-than lis*? sanually scram the reactor and operate at least one residual heat Eremoval loop in the suppression pool ~ cooling mode.. 3. With :na....;ression chamber average water temperature greater . i than 12/ ?, depressurize the reactor pressure vessel to less than 200 asil within 12 hours. I With one supprasa;:n chamoer water temperature instrumentation c. channel incper :'.d. restore the -inoperable channel to.0PERABLE status within. 7 da:'s -- verify suppression cha=ber water temperature to oe j ti:hin :h2.i;.- 1: '. 2 2 3 : once 7er 12 hours. I d. 'Ji:h both rap r2ssion enamber water camperature instrumentation . j channels inoperacle, ' restore at least one incperaole camperature ins t rume ntat i]n :nannel :o 0? ERA 3LI 3:atus 41:n;n 3 hours ar be in at least HOT 3HU! 4:. sitnin ene next 12 hours r" " ^ ^ ' ' S H'l'"DCWN within the !;..:v.ng 21 hours. With the crywa_ -:o-suppression chamber bypass leakage in excess of ) e., the limit, rast' ara the bypass leakage :o wt:hin :ne limit prior co - j increasing reactor coolant temperature socve 212'?. SURVEILLANCE REQUIREMENTS 1 l 4.6.2.1 The suppression chamber shall be demonstrated OPERA 3LE: a. By verifying che suppression :nameer sacar

1;:a :; ::e si : n _.: :13 1121:a at leas: :nca ;er ;- 1;urs.

kl 4 a l l 1 r i I l I, 1 3RUNSWICK - UNIT 2 3/4 6-LO Amend =ent No. '. 0 3 1 l-

L CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i b. At least once per 24 hours in'0PERATIONAL CONDITION 1 or 2 by verifying the' suppression chamber average water temperature to be less than or equal to 95'F, except: -W 1. At least once-per 5 minutes during testing which adds heat to the suppression chamber, by verifying the suppression chamber i sverage water temperature to be less than or equal to 105'F. )

2..

At least once per hour when suppression chamber average water i temperature is greater than 95'F, by verifying: I I l a) Suppression chamber average water temperature to be less { than or equal to 110*F, and b) THERMAL POWER to be less than or equal to 17. of RATED THERMAL POWER. l ( 3. At least once per 30 minutes following a scram with suppression chamber average water temperature greater than 95'F, by j verifying suppression chamber average water temperature less than or equal to 120*F. By an external visual examination of selected. emergency core cooling c. system suction line penetrations of the suppression chamber enclosure price to taking the reactor from COLD SHUTDOWN after safety / relief valve operation with the suppression chamber average water temperature greater than or equal to 160*F and reactor coolant system g l pressure greater than 200 psig. l-d. By verifying at least two suppression chamber water temperature instrumentation channels OPERABLE by performance of a: 1. CHANNEL CHECK at least once per 24 hours. 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and 3. CHANNEL CALIBRATION at least once per 18 months (550 days). l L wtth the temperature alarm setpoint for high water temperature less j than or equal to 95'F. (CAC-TE-4426-2 thru 131 CAC-TY-4426-1; i CAC-TR-4426-1) (CAC-TE-4426-15 thru 26; CAC-TY-4426-2; CAC-TR-4426-2) l e. At least once per 18 months by: 1. A visual inspection of the accessible interior of the suppression chamber and exterior of the suppression chamber enclosure. j l BRUNSWICK - UNIT 2 3/4 6-10a Amendment No. 111 L j

a, CONTAINMENT SYSTEtiS '

l l SURVEILLANCE - REQUIREMENTS (Continued) 1 2. Conducting a drywell-to-suppression chamber bypass leak test at i an initial differential pressure of l psig and verifying that ) the differential pressure does not decrease by more than 0.25 inches of. water per minute for a 10 minute period. J J ~; '.j l 4 i 1 l l 4 i 32.* :S LICK 'lNIT 2 3/4 o-106 Amena ce n t :lo. 103 t

.] ^" ,. : g w 'M EQUATION SHEET ] L f = ma -- - v = s/t: j g

- w = mg sl= v,tL+3 sat -

Cycle efficiency =' "I ' 1 E = mc a = (vg - y )/t\\ 2 / . KE = hmv ~* A = AN 'A=Ae f = v, +1 a v .PE ='mgh .m = e/t, A = in 2/tg = 0.693/tq; ~ W =;v4P } "'(t,)(ts) AE = 931Ah! (t'+tf) g Q = inC AT 7,7,-Ex. '1 p ,g ,,-Q = UAAT 7,7,-px ~ Pwr" =' W m" -x/TVL D ^] g I=I 10 y = p. loSUR(t), TVL = 1.3/u e /T HVL ai'. 0.693/u t P=P .o. 'SUR = 26.06/h ~ T_= 1.44 DT SCR = S/(1 - Keff) I x eff -SUR = 26 1 CR =.S/(1 - Keffx) (A-p x 'T " '(1*/p ) + [(fi-' o)'/A,gg ] 1(1 ~ K8ff)l.= CR (1 - K,ff)'2' =~ 2 o .T = 1*/ (p fy M " I/(1 ~ Kefg).= CR /CR g O T " N " 9)I A P eff H = (1 - K,gg)0 (l ~ Kdff}l / p = (K,ff-1)/K,ff = AK,gg/Kaff SDM = Q ~K,gg)/K,gg [1*/TKygg.] + [I/(l' + A,gg )) 1* = 1 x 10 seconds p'= T ~ ~ P = I$V/(3 x 1010) ~ ~1 A,gg = 0.1 seconds A I = No diy=Id22 WATER PARAMETERS Id=Id g 2 1~ gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters) 2 4. 1 Sal. = 3.78 liters R/hr = 6 CE/d (feet) 1 ft = 7.48 gal. MISCELLANEOUS CONVERSIONS 3 ~ 10 Density = 62.5.lbm/ft 1 Curie = 3.7 x 10 dps Density = 1 gm/cm 1 kg = 2.21 lbm i Heat of valorization = 970 Etu/lbm I hp = 2.54 x 103 BTU /hr ) 6 Heat of fusica = 144 Btu /lbm 1 Hw = 3.41 x 10 Beu/hr 1 Atm = 14,7 psi = 29.9 in. I'g. I Btu = 778 f t-lbf 2 I ft. H O = 0.4333 lb'f /in 3 inch = 2.54 cm 2 F = 9/5 C + 32 "C = 5/9 (*F - 32) L

m _- -. I f:. i e 't y THEORY OF NUCLEAR POWER PLANT OPERATION.- Paga 25 p-FLUIDS.AND TIIEBtiODYNAMICS E .i 7 ANSWER 5.' 01L (2.50) la.-Power Core Flow-Pressure Inlet' temp.- f (Any'3.0 0.25 ea)- -b.)MAPRAT K APLHGR(act)/ MAPLHGR(LCO) % N M M [/ M 48. ((o;s) t C. 7 51 -

c. No

(0,5).

d. The clad temp. could. exceed.2200 Deg. F. during a DBA.

(0.75). l REFERENCE- -BRUNSWICK THERMAL-LIMITS./ chapter 9, pp.9-68 th 9-78. Lesson Objectives. -on pp.- 9-1A (Several no numbers. assigned)- ANSWERL .'5. 02 '(0 50)- d REFERENCE -BFNP: GET;-Mitigating Rx Core' Damage LP,p 4. GGNS: 'OP-RP-502,P.5-7 BSEP Lesson Plan 3A, Radiation Control, Rev. 2 Sec. 2.14'pp. 11 Objective h.(RO) and g.(SRO) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

I e 5.. ' THEORY OF NUCLEAR POWER PLANT OPERATIQL' -Paca 26 FLUIDS.AND THERMODYNAMICS f L { ' ANSWER f5.03. .(3.00) l A. At 500 deg F. [0.25], As moderator temperature increases,. neutron leakage out of the fuel bundles is increased, thus the control rod is exposed to higher neutron flux and rod worth increases. [0.75] B. The withdrawn ' rod: [0'. 25], Neutron flux is higher in this area, thus rod worth is greater..[0.75] l C. At 10 [0.25], The void content of the upper portion of the core is. high at operating conditions,'the effects of deep control: rod with-3 drawal can be substantial radially. The negative power effect of-i . increased voids above the control rod is not seen because the channel length above the control rod is relatively short. [0.75] l l REFERENCE Susquehanna-Reactor Theory SC023 A-6 rev 0, pg 9 & 10 and attachment A, pg 1 & 2. I BSEP Lesson Plan 2A, Reactor Theory, Rev.2 Chapter 14 pp. 187,189, and 190. Lesson Objective No. 59 ANSWER 5.04 (1.00) c 1 1 l REFERENCE EIH: L-RQ-667, p 10 l BFNP: Rx Heat Balance LP; RQ 85/03/05 BSEP: Lesson Plan, Heat Transfer, Chap. 8 pp. 8-50 th. 8-59. Objective is second'from bottom of page 8-1, (no assigned number) ) l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

h ut. l5..' ' THEORY OF NUCLEAR POWER PLANT'OPERATIONu ' Para 27 O a, FLUIDS.AND THERMODYNAMICS 4 ') l ANSWER 5.05' ' ( 1. 50 )- .a. Decreases- -;b.. Increases

c.

Decreases- '(0.5'ea) REFERENCE Second Law'of Thermodynamics ~.BFNP: BFN Entropy.LP;-BFN Energy, Power, and Enthalpy LP; RQ 85/03/03 & 04 BSEP: Lesson Plan--Heat Transfer, Chapter 6. Lesson Objective : last 3 on page 6-1 (no numbers assigned) - ANSWER-5.06 (1;00) P = P(init.) e E t/T t 7 minutes =~420 sec. 1 Kw ='1000 watts. 1 1 P=( 1000 watts) ( e E 420/100 ) (0.25) J = (1000 watts) (e E 4.2) (0.25) j 1 (1000. watts) (66.7) (0.25) l = = 66700 watts = 0.0667 MW. (0.25) (Steps may be combined) i l l 1 ) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) j i

( A' s. I 15; THEORY OF NUCLEAR POWER PLANT OPERATION. Peco.28. ' FLUIDS.AND TWRMODYNAMICS

REFERENCE.

-1 i BSEP Lesson Plan, Reactor: Theory, pp.-131. Lesson Objective No. 43. .{ I l'! 5.07: _(2.00) ' ANSWER. la.- '295l des F (+- 15 deg F) b.- Increase o. Increase -d. 450 psia (+- 50 psia) [4 @ 0.5 ea.] (2.0) REFERENCE l Steam Tables /Mollier Diagram BSEP Lesson Plan - Heat Transfer Chapter 4. Lesson Objective No. 3 and 5 from bottom of page 4-1(no number assigned) l ANSWER 5.08 (2.50)

a. MSIV closure causes reactor pressure to increase [0.5) which collapses the voids (decreases the void concentration) adding

-positive reactivity. [0.5] (1.0)

b. All-(9) Safety Relief valves actuate.

(0.5)

c. Decay heat is causing pressure rise due to SRV closure.

(0.5)

d. SRV actuation reduces vessel pressure causing level swell (0,5) l i

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) 1

a; 4 6 '!Sa' THENRY OF NUCLEAR POWER PLANT' OPERATION, Pega 29 FLUIDS.AND THERMODYNAMICS m, 'MREFERENCE q B'SEPi HO"05-2-A, Anticipated Operational Occurrences Lesson Obj..b, Epp..i';(Figure is ATTACHMENT ~8) (l. 50k ' ANSWER ~.- 5.09 -(e-etO i a. Rapid. increase-in core flow causes decrease ~in transient void fraction (0.5) by replacing steam voids-with subcooled. liquid ,(0.25) thus adding positive reactivity and causing reactor power to. increase.-(0.25) b. Pump speed change increases drawdown in downcomer region causing indicated water level to decrease. (0.5) Low power and_ flow conditions permit thermal margins to more closely c. approach safety limits. (0.25) Power at 65% and flow of 50% are at the' low end of the automati w, control characteristic curve.(0.25)- . REFERENCE BSEP: HO 05-2-A, Anticipated' Operational Occurrences Lesson Obj. b, pp. 1 (Figure is ATTACHMENT 27), P. 8,53,54

ANSWER 5.10 (1.50)

Worth due to voids = (1 X 10E-3 dK/K/%V) (1.5%V) (0.5) = 1.5 X 10E-3 dK/K Worth due to fuel temp. = (1.0X10-5 dK/K/F) (40 F) (0.5) = 0.4X10E-3 dK/K ' ROD WORTH = VOIDS + FUEL TEMP. = 1.9 X 10E-3 dK/K (0.5) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

r m.. . ;- _p -

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. Pago 30, Pf FEUIDS.AND THERMODYNAMICS li L 1 REFERENCE I t ' )ReactoriTheoryLSec..liPg.116,14','and9. ,/LBSEP1 Lesson:: Plan 2A,'. Reactor Theory, ~ Chapt._14 pp. 172 & 181. Lesson KlObjective No.:58.. ANSWER 5.11) L ( 2. 00 )l .q <The reactorlis now producing less steamLto go to the turbine.(0.5)~ There will'be less extraction steam and reheater drain steam going to othe feedwater1 heater (0.'5)1Therefore,'less feedwater heating-will; occur Jresulting:in. colder'_feedwater entering.the-vessel (0.5) which will cause reactor power to increase'(about 3%) from_the_ positive _ reactivity

addition (alpha: m) -(0.'5)

) IREFERENCE-SSM-BOOK;-9,1CH'18 - A,Sec. 2.2.4, Pg. 39. -BSEP Lesson Plan 2A, Reactor' Theory Chap. 14'pp. 164 Lesson Objective ~ - No.s 52.'And. AOPLO3.0 pp. 3 (for effect) ANSWER 5.12 (3.00) a, 'Feedwater temperature-Feedwater flow RPV pressure RPV. water level (0.50 ea., any 3) ' b. _1. High' flow, High power (0.50) 2.- High flow, High power'(0,50), due to the increased inlet .subcooling from the increased feedwater_ flow. (0.50) i j (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

m pl.i. j, ! } l# i ,i',,s' ^ THEORY OF' NUCLEAR: POWER PLANT OPERATION <

Pega'31_

%,!Et

FLUIDS. AND THERMODYNAd@

p ..o EREFERENCE . u: n-

GE!BWR ACADEMICiSERIES ON HEAT TRANSFER AND FLUID FLOW cBSEP HO-10-2A1(RECIRC, FLOW: CONTROL):PG:42 -

7BSEPiLessonrPlan 10A,' Reactor Reciro. Rev. 2:.- pp. 42,26,and 39 Lesson

Objectivesag,j,& e.

l .Less~on Plan, Heat Tranafer,; Chap. 7. pp.7.-93.:Th 7-97 4 1 ANSWER? 5.13 l(1. 00 ). j 10btain'correspo'nding temperatures from the steamLtables by interpolation. .1000 psig = 546.3 deg-F'for-1014.7 psia. .(0.25) 250'psig.='406;0 deg F.for 264.7 psia. .(0. 25 ) - 1 Determineithentemperature' change: 546.3;- 406.0.=140.3 deg~F (0.25) . Determine the rate of cooldown: -140.3/1.75 = 80.21deg F/hr. (0.25)' .(Will accept + or - 2 deg. on' temps.) REFERENCE BSEP. Lesson. Plan,-Heat Transfer. Chapt 4. Steam. Lesson Objective. No,'sL3 and~5 from bottom of page 4-1(no assigned number) ANSWER 5.14. (0.'50) . (c)- u l REFERENCE BSEP Lesson Plan 2A' Reactor Theory, Rev. 2 pp. 174 Lesson Objective ~ I No.6-l I l' 1 ly -(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

'...:: 5 ? LTHEORY OF NUCLEAR: POWER-PLANT) OPERATION. Page 32 FLUIDS,AND' THERMODYNAMICS .y,,, _y .x iANSWER 5.15 ~ (2l50 ), o I

c. The delayed; neutron fraction is-the. percentage (fraction)'of 1

fis' ion neutrons'that'are' born-delayed..(1.0) f s b'.,U-238;((0.5) c., Decrease.(0.25)..As.Pu-239 production increases:.(0.25) and U-235 decreasea (0.25) the core average will decrease due'to r Pu-239's' Beta:being.so much smaller (0.25). REFERENCE' LNUS-Reactor Theory section-11.3' - BSEP:' Lesson Plan, 2A Reactor Theory, pp. 120 th 123. Lesson o -. Objective No.'s~38 &.40. .j ) (***** END OF CATEGORY 5 *****) r l 1 C J

y .q. 1 8..~ LA'NT SYSTEMS (DESIGN,-CONTROL,'AND INSTRUMENTATION' =Pega'33-t ANSWER'

6. Ol' (2.00)-

'l.= The local power may be lower than the' core average power. (3.0)'

2.. :Several of the highest' reading LPRM'S normally fed to the i

RBM might be bypassed. (1.0) g

REFERENCE

'BSEP: SSM, RBM, P. 6, Lesson Objective a. ? ' ANSWER' -6.02 (2.25) phd ,p,

1. Flow 800 gpm or greater /.(0.25) The flow is high enough to provide-cooling for the pump.(0.50)

.g. _ g g,- C S T : t o. t h e s u p p r e s s i o n p o)o l. ( 0. 5 0 )(F001)(fully closed. (0.25) Pre

2. Steam supply valve 4

3. Turbine stop valve (V8) fully closed.(0.25) Prevents draining the CST to the suppression pool. (0,5) i i

REFERENCE- .BSEP Lesson Plan 14, HPCI,pp esson Objective 4. I ANSWER 6.03 (2.00). Prevent possible DRYWELL implosion (0.50) due to the external a. pressure exceeding 2 psig (0.50) during the rapid void collapse (0.50) b. Insures a discharge path for the RER pumps. (0.5) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Page 34 REFERENCE BSEP HO 142-3DR3,RHR pp 28 & 30 Lesson Objective No. 10 ANSWER 6.04 (2.50) a. If the pump is shut down prior to the initiation signal clearing, white light " INITIATION SIG SEALED IN" (0.5) uill remain illuminated until the initiation signal clears and is reset by the operator.(0.5) Until the signal is cleared and reset, the pump will not automatically start.(0.5) 'b.

1. Only one of them can be opened manually when Rx Press is >410 psig.

2. The Outboard valve can only be opened manually if the corresponding Inboard valve is closed, regardless of Rx. press.

(0.5 ea) REFERENCE BSEP LESSON PLAN HO 142-3ER3, CORE SPRAY pp 10 & 11. Objectives 10 & 11. l l l l 1 I l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) i

6..

PLANT'BYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Page 35 ANSUER 6.05 (3.00) a. The valve plate is attached to a counterweight inside the valve body.(0.5) i b.

1. High Chlorine at the building intake

(

2. High Carbon Bed temp.

(0.5 ea) I c.

1. Control Bldg. radiation monitors reset.

2. Control Room fire detection system reset.

3. Emergency and filtering train control switches selected to either the PREF.or STBY positions.

4.'No clorine detected in the control bldg. intake plenum (Any 3 @ 0.5 ea.) REFERENCE BSEP LESSON PLAN 36B, Control Room HVAC pp 4,24,25 & 26 Lesson Opjective No. d. (No SRO objective in L.P.) - ANSWER 6.06 (1.50) 1. Deluge valves for both carbon filters will not automatically actuate. 2. The heater and fan cutout (at 210 deg. F.) is defeated. 3. The (180 deg. F.) heater cutout is defeated. (3 at 0.5 ea) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

t l 6. ' PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Paso 36 . REFERENCE BSEP LESSON PLAN HO-15-2/3-F, SBGT, pp 16. Lesson Objective No. f. ANSWER 6.07 (3.00) a.

1. AC motors have less potential for sparking than DC motors. (0.5)

2. AC motors are more reliable in a hostile environment.(0.5) i b.

1. Overspeed

2. Manual Turbine Trip (RTGB) i

3. Actuation of local-manual trip lever.

4. RCIC pump low suction pressure.

5. Turbine Exhaust Pressure high.

6. Auto Isolation Signal (Any four at 0.5 ea.)

REFERENCE BSEP LESSON PLAN HO142CR4, RCIC pp.10 & 11 Lesson Objective No. 1&9 l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) L

6.. PLANT SYSTEMS' DESIGN, CONTROL, AND INSTRUMENTATION-Paco 37 i ANSWER-6.08' (3.00)

a. Stablizer Brackets-(8).are welded to the vessel and' allow for (four inches) vertical movement due to thermal expansion. (0.50)

These stablizers.(Spring loaded turnbuckles) tie the RPV to the biological ~ shield. (0,5) b.

1. Minimizes. radiation exposure-to personnel during drywell entry while, reactor is shutdown.

2. Extends'the' lifetime of drywell componments.

3.. Limits the neutron activation of componments in the drywell.

4. Acts as thermal shield for the primary. shield outside the drywell.

(AnyJ3 at 0.33 ea)

c. The remainder of the torque (tension) is applied as the moderator is heated up.(0.33)

The stainless steel expands more than_the inconel bolt thus more tension is applied to the head.(0.33) This design-makes it easier toiinstall and remove the head.(0.33) REFERENCE. BSEP LESSON PLAN - H0082-3A, Reactor Vessel Internals pp. 2,3 & 14. Lesson Objective No. 2. . ANSWER 6.09 (1.50)

1. Mechanical overspeed, trip device-3

2. Mechanical trip solenoid.

{

3. Manual mechanical trip.

(0.5 ea.) I l i 1 1 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

m 1.- Paga 38- ' 6.. PLANT' SYSTEMS' DESIGN. CONTROL.- AND' INSTRUMENTATION . ~. LREFERENCE' TBS'EPJLesson Plan EHC'iO2A pp. 18.1 Lesson. Objective-No. 12. ' ANSWER '6.10 (1.50) 1.. Capable of. supplying' cooling water to the main condenser to remove heat rejected from the power cycle. 2. Safe. operation 1at-extreme low tide and peak surge elevation ,resulting from the Maximum-Probable Hurricane. 3. -The head developed by the pumps can overcome system friction. 4., Pumps rated to-permit operation of Unit'2 at 105% Steam dump. without a trip in Unit 2. In Unit 1: the steam dump ~ can be used with only 25% design flow. (Any 3 @ 0.5 ea.) ~ REFERENCE' BSEP Lesson Plan HO-22-2/3-A1, pp. 1,. Lesson Objective d.(SRO) ' ANSWER 6.11 (2.00) a..1. RHR Pump Room Cooler "B"

2. CS Pump Room Cooler "B"

~3. RHR Pump'"B" Seal HX.

4. RHR Pump "D" Seal HX.

(Any 3 at 0.5 ca.) L .b~. Closure of the associated RHR pump motor breaker. (0.5) i i l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) i l-

6. PLANT SYSTEMS DESIGN,LCONTROL, AND INSTRUMENTATION Pcg2 30 REFERENCE BSEP Lesson Plan HO-22-2/3-B&C pp. 25. Lesson Objectives c-d & g. ANSWER 6.12 (1.25) i i The bypass valve will start closing to put more water through the HX.(0.5) The valve cycles to maintain temperature between 80 & 90 deg. F. (0.5) A high temperature alarm will be initiated at 105 deg. F. (0.25) (Will accept temps. within 5 deg. F.) i REFERENCE l BSEP Lesson Plan HO-23-2/3-A, pp. 8 & 9. Lesson Objective a. i ANSWER 6.13 (0.50) True REFERENCE l BSEP Lesson Plan HO-12-2/3-A pp. 12. Lesson Objective No. d. ANSWER 6.14 (1.00) i

1. A half scram in one channel AND

2. A failed relay or blown fuse in the other channel, i

) l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) I l i--

J2 ! PLANT-SYSTEMS' DESIGN. CONTROL. AND INSTRUMENTATION.- Pcasl40 ~- L-REFERENCE' BSEP: LLesson Plan'HO-28-2/3-A pp. 15 ' Lesson Objective No. b. g l I (***** END OF CATEGORY 6 *****)

N &.. 7. '. PROCEDURES - NORMAL. ABNORMAL'. EMERGENCY Pasi 41 AND RADIOLOGICAL CONTROL ANSWER-7.01' (2.00)

1. Suppression Pool' Temp. >-95 deg.'F.

'2. Primary Containment Volumetric Average Temp. >135 deg. F.

3. Drywell Press. > 2.0 psig.

'4.' Suppression Pool Level <:-31 inches gysnan ful h.wl > - 2 7l' ,y

0. 5 ea. ).

' -(g g ! REFERENCE ,BSEP: EOP-01-ACCP'Rev. 5 pp. 3 ' ANSWER 7.01 (2.50)

a.-The temperature of the Suppression Pool, during ATWS conditions, when Boron injection is' required.(1.0) b'.

The purpose is to eatablish a point at which Boron injectice must be initiated in order to shut down the Reactor before the Heat Capacity Temperature Limit is exceeded.(1.0)

c. This temperature is 110 deg. F. (0.5)

REFERENCE BSEP EOP-01-U0, USER'S GUIDE pp. 64 o L (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

n.m ' ~:<b In Y 7 '.i PROCEDURES - NORMAL. ABNORMAJo.JtERGENCY-Pagol42L AND RADIOLOGICAL-' CONTROL- [ ANSWER' i 7. 03 L (1.50)

1. Depressurization of the1 reactor is required'AND

2. Less'thanLthree SRV'S"can be opened AND

.3. Reactor Press. at'least-50 peig above suppression chamber pr6ssure. .(3 0 0.5 ea) ' REFERENCE-BSEP EOP-01-EDP,-EMERGENCY DEPRESSURIZATIO

N. PROCEDURE

Rev. 2 pp. 3. ANSWER -7.04 (2.00) T1. Erratic'MSR or reheater drain tank level indication, locally.or annunciators'on UA-3, 2. Abnormal feedwater heater drain flow. 3.,iHigh temperature differential on the L.P. ' turbine Steam inlet lines side to side. (Points'7 & 10 on TT recorder)

4.

Abnormal indication on the MSR leak. detectors. (XU-2) 5. High differential temperature across LP turbine.could result in distortion and ' rubb'.ng in the turbine (.*.ny four at 0.5 ea) ! REFERENCE' BSEP:AOP-10,-MSR Tube Failure Rev. 00 pp. 3 (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) I

_ =_ _:

n

.J.. PROCEDURES -' NORMAL. ABNORMAL. EMERGENCY-Pega IND RADIOLOGICAL CONTROh- = ANSWER .7.05 .(2.00) I 1. Loss ofLinstrument power to P503 and BOP control board. 2.. Loss 7ofLfull core display, j 3' Loss of process. computer.. 4.. UPS Primary Power Converter' Trouble Alarm (UA-6, 6-7). 5. UPS Standby Power Converter Trouble Alarm (UA-6, 6-8). lB. UPS Transfer to Reserve Bus Alarm (UA-8, 6-9). 7. Possibla loss of SJAE's. (Any four at 0.5 ea) REFERENCE BSEP.AOP-12.0, Loss of UPS Rev. 001 pp. 3. ANSWER . 7 ' 06. (1.50)

1. Increase on main steam line radiation recorder (XU-3) 2-.

MAIN-STEAM LINE RAL HI annunciator (UA-23, 2-6)

3. MAIN STM LN HI/HI RAD TRIP annunciator (A-3, 3-7)

4. Momentary-increases in power level, core flow, and pressure 7"per n- <;- cFf-Mr fd&H~ /kh or lhctu Cn~ Gk1 RAb. IlL ann ~nneloc.

. 5,' e ( Any three at 0.5 ea) . REFERENCE BSEP: AOP-26.0, High Reactor Coolant or Condensate Conductivity Rev. 00 pp. 3. I I I l l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) lL _ _ ___________ _ ______ _ _ _ a

w-l 7J PROCEDURES ' NORMAL.' ABNORMALM EMERGENCY-Pcrol44-AND' RADIOLOGICAL CONTROL. 1 ANSWER-L

7.07 (0.50)

Prevent a Group 1. isolation while trying to reduce pressure.(0.5). ) 1 . REFERENCE m 1 BSEP:'AOP-32.0 'Rev.ll app. 3 j s 4 l ANSWER 17.08 ,(2.00) '1. Obta'n the keys'to the remote shutdown panel from the SOS key locker' i

2. Manually Scram the reactor.,

j

3. Trip the maingturbine.

l

4. Verify or manually Transfer auxiliary power.tx> the SAT.

L5. Place Mode Switch to Shutdown (For Unit 2, when steam flow < ,1j i 3,000,000 lb/hr)

6. Trip both recirc pumps, j

(Any four at 0.5 ea)

REFERENCE l

BSEP AOP-32.0 Rev 11 pp. 3. q b ^ ANSWER 7.09 (2.50) i a. Within 24 hours (0.50) after exceeding 15% thermal power (0.50) b.- 05!deg F (0.50) c. 110 deg F-(0.50) d. 120 deg F,'O.50). l-I- e l .(***n* CATEGORY 7 CONTINUED ON NEXT PAGE *****) I' l

e ':i;. :

7.'. -

PROCEDURES - NORMAL',~ ABNORMAL',-EMERGENCY Pers_45-AND RADIOLOGICAL ~ CONTROL ).- REFERENCE BSEP GP-03,' TECH SPEC 3.6.6.3,'AND TECH SPEC 3.6.2.1. ANSWER 7.10 (2.50) a. 1. > 5.0 mrem in any one hour or 100 mrem in'5 consecutive days 2. > 100 mrem in any one' hour 3. > 1000 mrem / hour j (3 @ 0.50 ea.) - b. Hands'and forearms,. including the elbows; feet, ankles, and lower ~ legs, including the knees.- (equivalent answer accepted) (1.00). ' REFERENCE I BSEP: VOL.VIII'(RADIATION CONTROL AND PROTECTION) PG 8, 10, 3'. 6 l ANSWER 7.11 (1.50) i

1. Not run for more than 72 hours (0.50) at greater than or equal to 810 i

psig.(0.50)

2. Shut down at 980 psig.(0,5) j (Will accept + or - 10 psig.)

I REFERENCE ) i BSEP: OP-02 Rev. 56 pp.12 (Reciro System Operation) l (***** CATEGORY 7 CONTINUED ON NEXT PAGE f****) I -a--

1 g..... t 1 / 7.; PROCEDURES NORMAL. ABNORMAL'. EMERGENCY 'Pago 46 AND RADIOLOGICAL CONTROL j1 1 4 ANSWER 7.12 (1.00) If this pressure is not maintained.the accumulator bladder may ' collapse (O.5) when' full. system' pressure is achieved.(0.5) L

REFERENCE:

j.. .BSEP: 'OP-05',-Standby Liquid Control System Rev. 21 pp. 5. ANSWER' 7,13 (1.00)- 1 .To avoid the possibility of double' notching (0.5)'and subsequent PCIOMR violation'.-(O.5) REFERENCE I BSEP: Operating Procedure.07, RMCS, Rev. 40 pp. 7 ANSWER 7.14 (1.00) To' ensure proper flushing (0.5) of the CRD seals.(0.5) REFERENCEL I BSEP:: Operating Procedure 08, Control Rod Drive Hydraulic System Rev. 7 l pp.'.7. i l l l-l t. (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) C j

{ 7.- PROCEDURES - NORMAL, ABNORHAL, EMERGENCY Paco 47 i AND RADIOLOGICAL CONTROL i ANSWER 7.' 15 - (2.50) l i

1. The valve and electrical lineup is as stated in the operating procedure with actions' required by the appropriate operating j

procedure to place the system in standby completed. 2.'All necessary components are operational for the system to perform its intended function in the required manner. i

3. The system is ready for automatic or manual initiation.

4. All system operability periodic tests are current.

?

5. Maintenance affecting system operability and all associated post maintenance checks that can be performed in the present plant condition are complete.

(5 @ 0.5 ea.) REFERENCE BSEP: General Plant Operating Procedure 01, Startup Checklist Rev. 101 s pp. 6 l ANSWER 7.16 (2.00) ~ l 1 1. The channel fasteners are located at one corner of each bundle 1 adjacent to the center of the control rod. (0.5) l l 2. The identification boss of the bail points toward the adjacent control rod (all bosses point to center of the fuel cell). (0.5) 3. The channel spacing buttons are adjacent to the control rod blades (0.5) ] 1 4. There is cell-to-cell symmetry (except on the core periphery) (0.5) 7h andl MC-t+ vamb > n Ah bl k'dl I'~ I ** t'" W " 5: y fu en }l cu t-r s A //~ Se / oJ/ (by9's0.51c-) l 1 (***** CATEGORY 7 C,NTINUED ON NEXT PAGE

- - - - )

j J

_., _7 - ) I 1 t ,'.)', t 4 74~ PROCEDURES'-(NORMAL ABNORMAL', EMERGENCY .Psco 48

AND-RADIOLOGICAL-CONTROL w

-REFERENCE': .BSEP: Fuel Handling Procedure,'FH-11, Section 4.27,'P 4 BSEP: SSMi Fuel, P. :.2, Lesson Objecti.ve 12 l s \\ s, } 1 l 1 i I i 4 ) (***** END OF CATEGORY 7 *****) j L 1 L a L__ _-_ --- - J

'k' 4 . 8.. ' ADMINISTRATIVE PROCEDURES ' CONDITIONS, Para 49 ,.AND LIMITATIQHE-g LANSWER 8.01' '(2.00) ~ c.L1.: Commercial: Telephone System to NRC Operations Center. .2.' Health Physics Network (HPN)-

3. NRC Operator (g

of'(~0.25formethod&0.25fororder[)) s_ .b. 'RegulatoryfCompliance-Instruction (RCI)'06.5, NRC Reporting Requirements. (0.5). .(Exact title or number not. required) < REFERENCE' lBSEP: Regulatory; compliance Instruction 06.5, NRC Reporting Requirements Rev. 5'pp'. 24 (Att. 3) (%Q C0 . ANSWER-8.02 (4700) a. Station: SRO or Duty CO (0.25 ea) Local: SRO or Duty CO (0.25 ea) Radwaste: Radwaste CO or Radwaste Shift Foreman (0.25 ea) .1 b. 1. Orange. Used when deemed necessary for conditions of a i temporary nature. 2. Orange. Used when deemed necessary for conditions of a permanent nature. 3. Yellow. Used when deemed necessary for conditions of a permanent nature. (on control boards and panels)- (0.25 ea color, 0.25 ea condition) c.- 1. The valve is cleared in the open position 2. The breaker is cleared in the closed position #,/ { .(0.50 ea) g;/- 1 L j. (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) { L __ _u- _j

b-28.l 3 ADMINISTRATIVE PROCEDURES.' CONDITIONS. Para 50

AND LIMITATIONS REFERENCE ~:

BSEPi AI-58 PG 8,'9, 10, ~ 11 AND FIGURES. ' ANSWER 8.03 (O'.50), ~ The. responsibility-of recommending off-site protective actions. . REFERENCE' l BSEP: PEP 2.3, 2.4, 2.5 SEC-2.0 ANSWER 8;04 (1.00) b (1.00) REFERENCE BSEP: U2 TECH SPEC DEFINITIONS l ~i I 1 l 1 l. l 1 (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

1 [?8 INISTRATIVE PROCEDURES. CONDITIONS.' Pc t 51

AND' LIMITATIONS

'f ANSWER 8.05- .(2.00) .1. Entry into any area where radiation levels are in excess of 100 l mrem /hr. ~2. Any maintenance work which involves opening of any system which contains or could potentially contain radioactive materials. -3; Any-maintenance of contaminated or potentially contaminated equipment using methods involving abrasion, cutting,. machining, or welding. 4.

Entry into.any-area where_the airborne MFC ratio is equal to or less than 0.25.

5. Any maintenance _or operation which could potentially cause an unknown radiological' hazard at the discretion of E&RC personnel.- Maintenance in an area. contaminated in excess of 5000 dpm/100 cm sq. or'any1 work.in an area-with measurable neutron exposure will be reviewed by.RC to determine if an RWP is required. '( Any 4 at 0.5 ea.) REFERENCI BSEP: E&RC MANUAL PROCEDURE, Radiation Control and Protection Rev. 018 pp.41. ANSWER 8.06 (2.50) l 1 a. It means to investigate and corroborate alarms and instrument readings immediately (0,5), not to blindly take action based on one indication (0.5) t b SF. wiJ b pm.'an.., f g,;,. n y

b. Safety Related -AGeneral Manager or Manager of Operations (0.5)

Other Equipment - Shift Foreman,(0,5) with concurrence of the 1 SOS.(0.5) i i l l 1 (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) { l r l

8 -ADMINISTRATIVE PROCEDURES. CONDITIONS. 'Pega 52 AND LIMITATIONS RFFERENCE BSEP: Operating Procedure 01-01, Operating Principles and Philosophy Rev. 019 pp. 8. I ANSWER 8.07 (0.50) l l To transfer experience with particular emphasis on Auxiliary Operator watchstanding practices (0.5). REFERENCE { BSEP: Operating Instruction 01-02, Shift Turnover Checklist Rev. 022 PP.1. ANSWER 8.08 (3.00)

a. An LCO for which the minimum requirements of the technical l

specifications are met by redundant equipment (0.5), the technical specification equipment is not required operable due to existing operational condition (0.50), or a conditional release is associated with operability or other applicability state (0.5) i

b. When the equipment removed from service places the plant in Technical Specification 3.0.3. (&:4r end chc w/*- AinJ L b+ 4,w.j/~. 6,,- r.ww c 4A +.4 L ~ a<1 w 1eya,,{f G j.j,

,,p- (o,2.c) l

c. At time the WR/JO was re iewed. (0.5)

d. At the time T.S.

3.0 3. is entered (You always have time to reach hot shutdown without scramming the, reactor). (0.5) I a l l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

l w-____________________

J

, 7 17 --,-- 1 uir ~

v. 7M :,

gg 37..t f8P ADMINISTRATIVE PROCEDURES. CONDITIONS.- Pega 53 1 ( jAND LIMITATIONS; 1 1 l 1(REFERENCE. J BSEP: Operating Instruction'OI-04.Rev. 025 pp. 5 Sec. 5.8 & pp.~ 6-4 .Sec.6.1.1.1. & Sec.6.1.1.2 k jl x c " ANSWERS 8.09 (1.50); 1.?When'a; valve / breaker is under_ clearance, f . 2.1When~a. valve / breaker is-being operated in the course of an approved j

procedure..

3'-When deviation'is approved by'the Shift Foreman. 1 - 4;;When a permanent! revision-is.being processed to the OP to change its

re q '. position'.

..(Any.'3at 0.5 ea.) JREFERENCE- ]l

BSEP: Operating Instruction-01-13 Rev. 15 pp. 5 Sec. 4.6.1.

j I ANSWER 8.10 (2.00)

1. Does it change the scope of the procedure?: 2. Does it involve.a QC hold point?

l

3. Does it place the plant in a limiting condition of operation or J ower i

state if not already required by the original procedure ? I

4. Does it cause plant systems to exceed their-design' requirements?

{

5. Does it alter or delete surveillance acceptance criteria, setpoints, or required operating parameter ranges ?

6. Does'it alter any administrative controls necessary to assure safe plant operation such as provided in the Security Plan, Fire Protection Plan, Emergency Plan, etc?

(Any four at 0.5 ea) 1 l l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) j

eq7x. = d v.;; ' f L'f i6. ? ADMINISTRATIVE PROCEDURES, CONDITIONS, . Pegs'54 cAND' LIMITATIONS REFERENCE. a

BSEP::HO-07-2/3-DO, L.O. 8, p.

i. ADMINISTRATIVE-' PROCEDURE,.AP Rev. 104 Sec.:5.5.3.2. pp. 5-16. l~ t.. LANSWER~ 8.11 L(2.00)! La.

1'. During Emergencies

'2. For Momentary. jumpers.and wire lifts.(Less than 3 insta11ed'or lifted at one time'and not'left unattended ) (0.5 ea). b. ' Shift Operating Supervisor (0.5) OperationN Shift' Foreman (0.5) c. REFERENCE 'BSEP:' Administrative Instruction AI-59 Rev. 9, Jumpering, Wire Removal, 'and Designated Jumper, pp.4 Sec. 4.11 ,.pp. 6 Sec. 6.3.2., and pp. 6'Sec. 6.3.5. ANSWER. 8.12 (1.00) Ensures fuel will not be loaded into a. cell without a control rod l (0.5).and prevents two positive reactivity additions from occurring simultaneously (0,5) ' REFERENCE BSEP: Unit 2 Tech. Specs. pp. B 3/4 9-1 i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) ._-___--__---_-__-_-___-____u

%4 ( j . nf F - u,, 8f.4 ADMINISTRATIVE PROCEDURES CONDITIONSi ' Pass.55-f lAND LIMITATIONS ~ j g 3 . 'f i 1 -t 4 - j l NSWER 8113 , (2.50) l'... Changes'.:to procedures required by T.S.(6.8),--or changes to other ? -procedures'/that' affect nuclear safety. .2, Proposed tests;orl experiments affecting nuclear safety.. H

3.: Proposed: changes to the Technical: Specifications.

l .i4.JProposed; changes ~to the Operating; License. i 5.6 Proposed modifications to plant systems or equipment that affect, nuclear safety. g (n0.5'ea); .+. ) REFERENCE 1 i. BSEP::HO-07-2/3-DO, L.O. 8, p. i .. Unit.2 Tech. Specs.'Section 6.5.2.1. pp. 6-9 r. LANSWER !8.14 - (2.50) l .NO.(0.5)'You must have Suppresion Pool water temp less than 95 degrees.F before entering Operational. Condition 2 (1.0) because Tech. Spec. 3.0.4'does not allow you to enter an operational condition while relying on:an action statement.(l.0) REFERENCE .i BSEP: HO-07-2-B1, Tech Specs and Bases L.O. 6, p. i Unit 2 Technical Specification 3.0.4 and 3.6.2 a ~ ! l l (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********) 3

7 v j ;,7, S. f {o l e "o TEST. CROSS REFERENCE Pago: 1l QUESTION VALUE' REFERENCE 4 5.01: 2 '.' 5 0 ZZZ0000001' .I 5'.02- -0.50 ZZZ0000002 15.03 3.00 ZZZ0000003 5.04 1'. 0 0 - ZZZ0000004 ' 5. 0 5 - 1.50-ZZZ0000005' 5.06 .1'. 0 0 ZZZ0000006 5.07 2.00 ZZZ0000007. -5.08-

2. 50 ~

ZZZ0000008 -5.09' l,6&rfMFr# ZZZ0000000 { ~ 5.10: 1.50 ZZZ0000010' . 5.11- . 2. 0(F

ZZZ0000011 5.'12

' 3. 00-ZZZ0000012 5.13 1.00 .ZZZ0000013 5.14 0.50 ZZZ0000014 '5.15' 2 50.' ZZZ0000015 2b $YH' 6.01 2.00~ -ZZZ0000029 6.02 2.25-ZZZ0000016 .6.03 '2.00 .ZZ20000017 6.04 2.50-ZZZ0000018 6.05 3.00 ZZZ0000019 '6. 06. 1.50-ZZZ0000020 s ~ 6. 07-3.00-ZZZ0000021' 6.08 -3.00 ZZZ0000022 6.09 1.50 ZZZ0000023 6.10 1.50-Z3Z0000024 6'.11 2.00 ZZZ0000025 6.12 1.25-ZZZ0000026 6.13 0.50 ZZZ0000027 i ~6.14 1.00 ZZZ0000028 27.00 -7.01 2.00 ZZZ0000030 7.02 2.50 ZZZ0000031 7.03 1.50 ZZZ0000032 7.04 2.00 ZZZ0000033 7.05 '2.00 ZZZ0000034 7.06 1.50 ZZZ0000035 7.07 0.50 ZZZ0000036 7;08 2.00 ZZZ0000037 17.09' 2.50 ZZZ0000038 I, 7.10 2.50 ZZZ0000039 3 7.11 1.50 ZZZ0000040 7.12 1.00 ZZZ0000041 i 7.13 1.00 ZZZ0000042 7.14 1.00 ZZZ0000043 7.15 2.50 ZZZ0000044 l 7.16 2.00 ZZZ0000045 l 28.00 1 8.01 2.00 ZZZ0000046 E 1

'. 4' .; r[ i 1-- H. s Q.g b - TESTfCROSS~' REFERENCE Pcca: 2-QUESTION VALUE' REFERENCE Y.oD 4 nn_ Zggoooo047 e8.02! in. 38.03. L 0. 50'- xZZZ0000048-t '8.04 1.00 ~ZZZ0000049-8.051 2.00-ZZZ0000050 " 8 '. 0 6 ' .2.50' ZZZ0000051 8.07. 0.50 ,ZZZ0000052 1 l - 8.'08 - 3 '. 0 0 ' ZZZ0000053-i8'.09 1,50 ZZZ0000054. .8.10~ 2.00.. -- Z Z Z00000 55.' g .c, 8.- 11L 2.00. ZZZ0000056 ~ "8.12 .1. 00.- ZZZ0000057 8.13. 2.50 ZZZ0000058. - 8.14' 2.50-ZZZ00ij0059 ?2F4G ). . s. u v yv m

1,0:

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Text

N. PROCEDURE

REFERENCE:

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