IR 05000423/1987023
| ML20236P603 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/06/1987 |
| From: | Keller R, Temps R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236P493 | List: |
| References | |
| 50-423-87-23OL, NUDOCS 8711180063 | |
| Download: ML20236P603 (95) | |
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'U.S. NUCLEAR REGULATORY. COMMISSION REGION.I.~ '
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0PERATOR : LICENSING' REQUALIFICATION ~ EXAMINATION REPORT ' , s f REQUALIFICATION EXAMINATION REPORT.NO. 50-423/87-23(0L) FACILITY DOCKET-N FACILITY LICENSE N0. NPF-49 . l LICENSEE: Northeast Nuclear Energy Company P.O. BOX 270-Hartford, Connecticut 06141-0270~ FACILITY: ' Mill ts one 3 Nuclear Power Station EXAMINATION. DATES: June 12, July 31 and September 4, 1987 CHIEF EXAMINER: (Nh/L2 R.R; Temps, Operati t Engineer Ju-cca 7-
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l , APPROVED BY: R.M. Keller, Chie[1I 1////67 Date Pressurized Water' Reactor Section ~ Division of Reactor. Safety
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SUMMARY: Requalification written examinations'anci operating tests were administered to six senior reactor operators (SRO's)- and three reacto operators (R0's). All SRO's and RO'_s' passed the operating portion of the examinations; however, two R0s and three SRO's failed the written examinations.
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l l: 8711180063 871112 . . tj PDR ADOCK 05000423 V PDR
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. . i U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING REQUALIFICATION EXAMINATION REPORT-REQUALIFICATION EXAMINATION REPORT N0. 50-423/87-23(OL) FACILITY DOCKET N FACILITY LICENSE NO. NPF-49 i LICENSEE: Northeast Nuclear Energy Company P.O. BOX 270 H'artford, Connecticut 06141-0270 FACILITY: Millstone 3 Nuclear Power Station EXAMINATION DATES: June 12, July 31 and September 4, 1987
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CHIEF EXAMINER: [f h. 8e R.R. Temps, Operations' Engineer
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APPROVED BY: , R.M. Keller, Chief Date Pressurized Water Reactor Section 1 Division of Reactor Safety i
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SUMMARY: Requalification written examinations and operating tests were ' administered to six senior reactor operators (SR0's) and three reactor i operators (RO's). All SR0's and RO's passed the operating portion of the i examinations; however, two R0s and three SRO's failed the written examination j l l l OFFICIAL RECORD COPY OL REQUAL MILL 3 - 0003. /3/87 , _ - _ - - - - _ . _ _
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DETAILS TYPE OF EXAMINATIONS: Requalification l EXAMINATION RESULTS: l R0 l . SRO l l Pass / Fail l Pass / Fail l 1 I I I I I I IWritten i 1/2 l 3/3 l l 1 I I I I I i 10perating l 3/0 l 6/0 l l l 1 I I I l l 10verall l 1/2 l 3/3 l 1 I I I CHIEF EXAMINER AT SITE: Robert R. Temps - Operations Engineer
- OTHER EXAMINERS: David Silk - Operations Engineer t Bill Hemming - EG&G Contract Examiner 1 Frank Jaggar - EG&G Contract Examiner 3.
' The following is a summary of generic deficiencies noted on operating test This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is require DEFICIENCIES During control room discussions, several SR0's used the EPIP's for l l Unit 1 or 2 when asked to classify various emergencies.
l The following is a summary of generic deficiencies noted from the grading ' l of written examinations. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is require ? 0FFICIAL RECORD COPY OL REQUAL MILL 3 - 0004. /3/87
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. DEFICIENCIES R0 Examination by Question Number i 1.06a Effect on shutdown margin when boron concentration is ' lowered while maintaining constant power and. rod j positio .08 Effect on indicated pressurizer level when the water I temperature in the pressurizer is change l l 1.10b Effect of lowering steam pressure on enthalpy for the 1 steam generator pressure maintained during normal I operatio .02a Unable to state both flowpaths for RCP #1 seal leakoff ; during safety injectio ' 2.03 Response of the Emergency Generator Load Sequencer when ; a CDA occurs following a loss of powe .04 Design basis for the interlocks associated with the RHR i loop suction valve I 2.05 The three signals which will automatically close the Condensate Storage Tank supply valve,to the AFW pump .03a Operation of the steam dump system when lowering.the setpoint of the pressure controlle .03 The four federal quarterly radiation exposure limit .04b Alternate lineups available to avoid a Reactor /RCP trip if seal injection drops to 6 gpm and only an CCW pump is availabl .05 Basis for a precaution taken from OP 3335D, " Radio-active Liquid Waste System".
SR0 Examination by Question Number 5.11 Unable to choose the one item from a list which was not an example of an evolution which could cause water hamme .04 Same comment as 2.0 OFFICIAL RECORD COPY OL REQUAL MILL 3 - 0005. /3/87
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i 6.07b Alarms / indications which alert an operator as to insufficient spray bypass flo .02 Requirements .for when an RWP is required for. entry and work in low radiation area . l
7.04a The three conditions which require that the RCP's be stopped when in E0P 35 FR- .01 Shift Supervisor. responsibilities in ACP 6.03 related to liquid waste discharge ] ! 8.04 The three categories of individuals allowed access to the s control room during emergencies.
l 8.08 Actions to be taken if during movement of irradiated fuel '{ in the spent fuel pool, both emergency generators are J determined to be inoperabl . Simulation Facility Fidelity Repor J During the conduct of the simulator portion of these operating . l tests, the following performance and/or human factors discrepancies i were observed: l l The expected response described for simulator malfunction number RC11 ( does not model the actual simulator response to this malfuctio : Specifically , the increase in system pressure is significantly lower than that described in the malfunction boo ! l 6. Personnel Present at Exit Interview:
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R.R. Temps, Operations Engineer G.S. Barber, Resident Inspector Facility Personnel J. Harris, MP-3 Operations Supervisor I M. Moehlmann, ATS-Operating training j B. Ruth, Manager, Operator Training J
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l R. Stotts, MP-3 Operator Training 7. Summary of NRC comments made at exit interview: The chief examiner presented the number and type of examination conducted over the previous three months. In addition, generic weak-nesses noted from observation of the operating examinations were also presente OFFICIAL RECORD COPY OL REQUAL MILL 3 - 0007. /06/87 *
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DETAILS I l TYPE OF EXAMINATIONS: Requalification EXAMINATION RESULTS:
l RO l SR l l Pass / Fail l Pass / Fail .I i l l l l I I l' l Written l l'/ 2 l 3/3 l l 1 I I I I I I l0perating l 3/0 l 6/0 l l l l l , I I l l i l0verall l 1/2 l 3/3 l a l l l l 'h l
' ! CHIEF EXAMINER AT SITE: Robert R. Temps - Operations Engineer OTHER EXAMINERS: David Silk - Operations Engineer Bill Hemming - EG&G Contract Examiner Frank Jaggar - EG&G Contract Examiner The following is a summary of generic deficiencies noted on operating '
tests. This information is being provided to aid the licensee in g upgrading license and requalification training programs. No licensee response is require .t DEFICIENCIES I i During control room discussions, several SRO's used the EPIP's for i Unit 1 or 2 when asked to classify various emergencie " The following is a summary of generic deficiencies noted from the grading of written examinations. This information is being provided.to aid the 4 licensee in upgrading license and requalification training programs. No licensee response is require ,
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DEFICIENCIES R0 Examination by Question Number 1.06a Effect on shutdown margin when boron concentration :is lowered while maintaining constant power ~and ro positio .08 Effect on indicated pressurizer level when the water temperature in the pressurizer is change .10b Effect of lowering steam pressure on enthalpy for the steam generator pressure maintained during normal operatio .02a Unable to state both flowpaths for RCP #1 seal leakoff during safety injectio .03 Response of the Emergency Generator Load Sequencer when a CDA occurs following a loss of_ powe .04 Design basis for the interlocks associated with the RHR-loop suction valve ) 2.05 The three signals which will automatically close the Condensate Storage Tank supply valve to the AFW pump I 3.03a Operation of the steam dump system when lowering the j setpoint of the pressure controlle .03 The four federal quarterly radiation exposure limit .04b Alternate lineups available to avoid a Reactor /RCP trip if seal injection drops to 6 gpm and only an CCW pump is availabl l 4.05 Basis for a precaution taken from OP 3335D, " Radio-active Liquid Waste System". ! SRD Examination by Question Number 5.11 Unable to choose the one item from a list which was not an example of an evolution which could cause water hamme .04 Same comment as 2.0 l _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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6.07b Alarms / indications which alert an operator'as to ! insufficient spray bypass flo ]
7.02 Requirements for when an RWP is required for entry and work i in low radiation area ' 7.04a The three conditions which require that the RCP's be l stopped when in E0P 35 FR- l 8.01 Shift Supervisor responsibilities in'ACP 6.03 related to ) liquid waste discharge .04 The three categories of individuals allowed access to the control room during emergencies, l 8.08 Actions to be taken if during movement of irradiated fuel'
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in the spent fuel pool, both emergency generators are determined to be inoperabl . Simulation Facility Fidelity Repor ' During the conduct of the simulator portion-of these operating { tests, the following performance and/or human factors discrepancies were observed: l i The expected response described for simulator malfunction 1 umber RC11 j does not model the actual simulator response to this malfuctio Specifically , the increase in system pressure is significantly lower j than that described in the malfunction boo . Personnel Present at Exit Interview: NRC Personnel l R.R. Temps, Operations Engineer l G.S. Barber, Resident Inspector Facility Personnel l J. Harris, MP-3 Operations Supervisor M. Moehlmann, ATS-Operating training , B. Ruth, Manager, Operator Training R. Stotts, MP-3 Operator Training 7. Summary of NRC comments made at exit interview: The chief examiner presented the-number and type of examir,ations conducted over the previous three months. In addition, generic weak-nesses noted from observation of the operating examinations were also presente t - _ - _ - _ _ _ _ - _ _ - _ - _ - _ _ _ _ _ - _ - - - _ . _ _ _ _ _ _ _ _ _ . . - --- - !
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. Examination Review A review of the written examinations was conducted immediately following the examinations. Facility comments were discussed on a line item basis. The number of facility comments were minimal and were resolved to the satisfaction of the chief examiner and the license Facility comments ,an be found in Enclosure 3. Additional changes to the examination key made as a result of grading are listed in Enclosure l
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Enclosures: Written Examination and Answer Key (RO) Written Examination and Answer Key (SRO) j Facility Comments on Written Examinations j Additional NRC Changes to Written Examinations Answer Keys -l
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. Examination Review A review of the written examinations was conducted immediately following the examinations. Facility comments were discussed on a line item basis. The number of facility. comments were minimal and were resolved to the satisfaction of the chief examiner and the license Facility comments can be found in Enclosure 3. Additional changes to the examination key made as a result of grading are. listed in Enclosure Enclosures: Written Examination and Answer Key (RO) Written Examination and Answer Key (SRO) Facility Comments on Written Examinations , Additional NRC Changes to Written Examinations Answer Keys 1
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U. S. NUCLEAR REGULATORY COMMISSION I FEACTOR OPERATOR REQUAL:IFICATION EXAMINATION I
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DATE ADtV: 3TERED: 87/09/04 i EXAMINEP: SE _ __
CANDIDATE __ __________,___.__.______ j i 10SIEYGI1ONS_IQ_C@dDlDeIEi j l Read the attached instruction page carefull This examination replaces ) the current cycle tacility acministerea requalliacatien examinatio Pettaining requirements for failure of this examination are the same es i for tallure of a requalification examination prepared and administered oy ) vour training s t a f -F . Fo i r, t e for each question are indicated in j parentheses after the questio The passing grace requires at least 70*4 ! In each category anc a final grade of at least 80%. Ex ami na t i on papers j will be picked up four (4) hours after the examination starts, l l '
% OF CATEGCRY '. OF CANDIDATE'S CATEGORY
__MOLUE_ _IDIBL _ SCOSE_ _,_ _ M e L U E _ __ _ ,_ _. _ ,, ,,,_ _ ,, _, _ _, _ C 9 I E G O S y _ _ _ ,, _, _ _, ... .. l _lh90__ 2h09 - . _ _ _ . _ _ _ . _ _ _ . _.__ PRINCIPLES OF NUCLEAR POWER l ! PLANT OFERATION, THERMODYNAMIC l l HEAT TRANSFER AND FLUID FLOW _1199 _ _2h09 _ _ . _ _ _ _ _ . _ _ _ _ _ _ . _ . _ _ _ PLANT DESIGN INCLUDING 3AFETY AND EMERGENCY SYSTEMS j _.1h09_ _ _52.99 _ _ . _ _ . . _ _ . _ _ . _ _ _ _ . _ . _ _ _ _ _ INSTRUMENTS AND CONTROLS l l _IMz99 2h99 ___.__ __ _ _ _ _ . _ PROCEDURES - NORMAL, ABNORMAL, l EMERGENCY AND RADIOLOGICAL CCNTROL e ' 00 _ _ . _ _ . _ . _ _ _ _ _. _ % Totals Final Grade 411 work done en this examination is my ow I how nei ther given nor r-acetved al _ _ _ _ _ . _ _ . _ . . _ _ . . _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _
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- e . a NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS i During the administration of this examination the following. rules apply:
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l 1.- Cheating on the examination means an automatic denial of your applicati'on j and could result in more severe penaltie . Restroom trips are to be limited ~ and only one candidate at'a time may leav You must avoid all contacts with anyone outside the examinatio ! room to avoid even the appearance or possibility of cheatin . Usa, black ink or dark pencil only to facilitate legible reproduction . Print your nanie in the blank provided on the cover sheet of the-examinatio Fill in the date on the cover sheet of the examination , (i f necessary). Use only the paper provided for answer . Print your name in the upper.right-hand corner of the first page of each-section of the answer sheet.
I 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 "Last Page" on the last answer shee Numoer each answer as to category and number, for example, 1.4, I l 10. Skip at least three lines between each answer.
l . l ti. Separate answer sheets from pad and' place finished answer sheets face down on your desk or tabl . Use abbrevi ations only if they are commonly used in facility literatur , 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 require j j 14 Show all calculations, methods, or assumptions used to obtain an answer l to mathematical pr oblems whether indicated in the question or'not.
l l td. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l DUESTION AND DO NOT LEAVE ANY ANSWER BLAN t If parts of the examination are not clear as to intent, ask questions of the examiner onl . 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 ccmpleting the examinatio This must be done after the examination has been complete .
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18. When' you compl ete" your' ex ami nati on , 'you 'shall:
, Assemble your examination as follows: (1) Exam questions on to . -
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(2) Exam aids -' figures, tables, et (3) Answer pages including figures which are part of'the answe b, Turn in your copy of the examination ~and.all.pagesLused to. answer the examination questions., ,;
i Turn in all scrap paper.and the balance of the paper that you'did i not use for answering the question Leave the examination area, as defined by.the examine If after leaving, you are found in this area'while the examination is still
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PRINCIPLES OE_UyGLEAR FOWER PLANT QEERAT' ION2 4
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QUESTION 1.01 (1 00) ; i Multiple Choice I During a reactor startup, the first reactivity addition caused'the count , rate to increase.from 20 to 40 cp The second reactivity addi tion l caused the count rate to increase.from 40 to 80 cp Which-'ONE of the l following answers is correct? The first and second reactivity additions were equal .
! The first reactivity addition was large The second reactivity addi tion was large There i s not enough data given to determin I
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I l QUESTION 1.02 (1.00) 1 Multiple Choice ,
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Which GNE the f ollowing statements most correctly describes tne change in the Fuel and Moderator Temperature Coefficients (FTC q and MTC) as the core ages? <
] FTC becomes more negative and MTC becomes more negativ I l
i FTC becomes more negative and MTC becomes more positiv j FTC becomes more posi tive and MTC becomes more negati v I I FTC becomes more positive end MTC becomes more positiv ' l
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l QUESTION - 1.03 (2.00) e
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A Xe-free reactor startup is in progress with power leveled out at 10-8 ] l' amps for critical data. Describe the effects, if any, on the parameters j ! listed below if rod D-4 (control bank D). drops to the bottom. Include- ] in your description both the transient behavior'and the fina j steady state conditio Initially Tave = 546 F and Primary j Pressure = 2235 psi l i
'1 Tave- ( 0. 50 )' Primary Pressuee (0.50) Reactor Power (1.00) ]
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QUESTION 1.04 (1.00) q
r1ultple Choice s Choose the correct phrase to correctly complete the sentence, l h As the core ages from BOL'to EOL, the ratio of PU-239 atoms lto U-235 j l atoms increases. This changing ratio causes the _______ reactor period to decreas I , void coefficient to become less negative, i 1 moderator temperature coefficient to become less negativ I delayed neutron fraction to increase.
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-1, PRINCIPLES OF-NUQLE68_EQWES_EL8NI_QEEBSIlQN t- Page- 6~
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QUESTION 1.05 (1.00) Multiple Choice Reactor power is lowered to 80%, following 100 hours of continuous operation at 100% powe Which ONE the.following statements best describes Xenon behavior during.the first hour following the. power decrease?
(NOTE: EXe] denotes xenon concentration.] Direct EXe] increases, indirect CXe] decreases, total CXe]
decreases, Direct [Xe] increases, indirect EXe] increases, total CXe] increase Direct EXe] decreases, indirect CXe] decreases, total CXe] decrease Direct EXe] decreases, Indirect EXe] increases, total EXe] increase .
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l DUESTION 1.06 (2.00)
State how each of the f ollowing will affect' shutdown margi Limit j your answer to INCREASE, DECREASE, or NO. CHANG Consider each cese ! separately. Assume EO i Boren concentration is decreased 20 ppm while maintaining constant ; power and no rod motion
) Dank D rod height is increased from 125 steps to 200 steps while l maintaining constant power and baron concentration 1 Reactor trip l l While shutdown, the RCS is cooled dcun by 40 degrees )
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l QUESTION 1.07 (1.00) Multiple Choice (Fill in the Blank) I During a Xenon-tree reactor startup, critical data was inadvertently 1 taken two decades below the required Intermediate Range (IR) leve l Assuming RCS temperatures and boron concentrations-were the.same, the j critical rod position taken at the proper IR level _________.th ;i
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critical rod position taken two decades-below the proper IR leve j Is Less Than
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' QUESTION 1.08 (1.50) l ' At normal hot standby conditions, in which direction will -INDICATED pressurizer level change as a result of the following transients?
(INCREASES, DECREASES, STAYS THE SAME). Assume the letdown and
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charging flows are equall:ed and the pressurizer level control system i s i n manual . Ccnsider each transient' separately.
, The reference leg heats up from 120 F'to 200 F due to the I relocation of a ventilation duc (0.50) The pressurizer heaters fail and the pressurizer water cools from normal operating temperature to 590 (0.50) For case 2 above, following the cooldown, is the indicated level GREATER THAN, LESS THAN, or EQUAL TO the actual level?
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I: 1 PBIUQlELES_QE_ NUCLE 68_EQWE8_EL@NI_QEE88IlgN 2 .Page- 8
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, OUESTION 1.09 (1.00) l State how an INCREASE in each of the following parameters"affects the Departure from Nucleate Boiling, Ratio (DNBR). ' Limit your answer to INCREASE, DECREASE, or NO EFFECT.
' Coolant temperatur Coolant flo QUESTION 1.10 (1.50) How will each of the f ollowing affect the results of a secondary calorimetric power calculation? Limit- your answer to CALCULATED LO'ERw THAN ACTUAL, CALCULATED HIGHER THAN ACTUAL, or CALCULATED.SAME AS ACTUA Consider each case separatel Measured feedwater temperature is 10 degrees lower than actual f eeclWater temperatur Measured steam generator pressure is 30 psig lower than actual-
, . ' steam generator pressur Measured feedwater flow is 1E5 lbm/hr higher than actual feedwater flo DUESTION 1.11 (1.00)
Multiple Choice Assume the plant is in Mode 3 at a temperature of $35 F, and'the steam dumps are NOT operabl To what value must Tavg rise be+ ore causing the power-operated steam generator pressure relief val ves (MSS *PV-20A, B, C, & D) to lif t? Assume the pressure setpoint controllers are set for normal power oper at i o (Choose the MOST CORRECT answer.) .1 F .7 F .6 F .2 F
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QUESTION 1.12 (1.00) I
' State how an INCREASE in the following parameters affects' Net )
Positive Suction Head (NPSH) available at the suction.of a centrifugal i pump. Limit your answer to INCREASE, DECREASE, or NO EFFEC i System flow rat , 1 System temperatur .l
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n 2 PLANT DES (GN I NCLdlD I NQ_S6ESTL A110_gligRQgNC Y Page 10 .J
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OUESTION 2.01 (2.00) 1 What TWO interlocks must be satisfied to allow the CVCS Orifice Isolation Valves to open? 4 What TWO conditions will cause the CVCS OrificeEIsolation Valves
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to automatically shut other than the interlocks stated above? R .. . DUESTION 2.02 (2.50) ,
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! l Briefly describe how'the No. 1 RCP shaft seal responds.to an T l injection pressure increase of 50 psig over' normal pressur Include in your explanation a discussion of ~the change-in forces on the top and bottom of the seal and the final equi 1ibeium position of the seal'. ' ( 1. 50 ) .[ k State the TWO flowpaths for the RCP-#1 seal leakoff during a . satety injecti o (1.00)
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QUESTION 2.03 (1.00) j l Multiple Cnoice l l Which ONE of the following statements best describes the response of-the Emergency Generator Load Sequencer (EGLS) when a' Containment Depressurination Actuation (CDA) occurs-45 seconds followinq a Loss Of Power (LOP)? There is no SI in Progres i l The EGLS will stop the load sequence, the loads will be stripped j and the EGLS will start the CDA/ LOP sequence at. time O.
I The EGLS will stop the load sequence, the loads will be stripped' and the EGLS will start the CDA/ LOP sequence at time 45 se The EGLS will stop the load sequence, the l'eads will be stripped l that are not required by'CDA/ LOP , and:the EGLS will start the CDA/ LOP sequence at time The EGLS will stop'the load sequence, the loads will be stripped-that are not required by CDA/ LOP , and the EGLS wil'l start the j CDA/ LOP sequence at time 45 se '
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QUESTION 2.04 (2.00) What is the design'. intent (basi s) for the. interlocks and automatic functions associated with the RHR loop suction valves (MV 8701A/B)? , QUESTION 2.05 (2.50) 'l i What THREE signals will automatically close the Condensate Storar,e 1 . Tank supply valves (AOV 23A and B) to the Au>iliary Feedwater Pumps { l (AFW)? (1.50) ]
i What is the purpose of the cavitating venturis just downstream of j l the flow elements in the AFW discharge lines? (1.00) I \ l
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OUESTION 2.06 (1.00) State the FOUR sources of makeup water to the Spent Fuel Poo Identify between normal and emergency (or "last resort") sources of l makeup, AND state any pref erential order of use for the normal and/or H the emergency source QUESTION 2.07 (2.00) aix List FOUR of the +tve radioactive liquid effluent monitors which have automatic control functions, AND briefly describe the control function j for each monitor listed.
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QUESTION 2.08 ( 1. 00 ) ~
, Match the pressure at which injection. starts in' Column B to the component of Column !
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Column A Colum j _____ ___ __________ 1. High Head Injection Pumps psig 2. Medium Head Injection Pumps b ,, 2540'psig
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I psig I I QUESTION 2.09 (1.00) State t-M- TWO automatic f unctions provided by the F-11 interlock, (set at 1985 psig), when RC3 pressure -i s being INCREASED. Do not inclu'de indication function t i ! l (*++** END OF CATEGORY 2 *****) l _ _ _ _ _ _ _ _ _ _ _ - - _
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-l QUESTION 3.01 (2.00) J A leak develops in the ref erence leg associated with the automatic !
level controller of the Volume Control Tank (VCT). As a result the ( indicated level in that leg f ails high. Describe the VCT level .f transient assuming that no operator action is taken and.that the:VCT j is in the automatic makeup mod Include.the reasons.WHY level change (tJurc ' D e ' a b. .a w .Un r ekw t < di a te cI THe r r:. a :u : An y v t,e ,ew .,J m. h,,w,,.a
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a n n u s o' s - wd n r n e . fw rw r c* , , , ' QUESTION 3.02 (2.00) ! J
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l Assume steady state operation at 100% power when the Master Pressure Controller setpoint for the pressurizer is inadvertently ] changed from 2250 psig to 2385 psi Assume a step change:in setpoint and assume that pressurizer pressure control is in automati What automatic action (s), other than the actuation of alarms / annunciators, will occur immediately? (0.50)
' Describe the pressurizer pressure transient that wl11' occur if l l
no operator action is take Include in your answer any other l automatic actions, other than alarm / annunciator act uati on s , that .] ' take plac ( 1 :. 50 ) l QUESTION 3.03 (3.00)
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l For each case below explain the resultant operation of the Steam. Dump l system AND indicate the approximate final RCS Tavg (+/- 2 F).
Assume all systems are normal except as stated and that no operato" action' is take Consider each case separately, The normal setpoint on the steam dump system steam pressure i ' controller (MSS-PK-507) is reduced from 1092 psig to 1007 psig while in Hot Standby awaiting reactor startu (2.00) The Train A steam cump bypass interlock selector switch is taken to "OFF" while stable at 5% reactor powe (1.00) i f i
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QUESTION 3.04 (1.00) With regard to the main feedwater pump speed control c i r c ui t r y ', what . i s the reason that the output from the steam + low summing ampiifier-is-conditioned within a lag. circuit? DUESTION 3.05 (2.50) State the.FIVE uses o+ the output of the first stace impulse pressure 'l transmitter (PT-505). Setpoints are NO T require t OUESTION 3.06 (1.00)
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Would INDICATED steam flow at 100% power be LESS THAN, GREATER THAN, j
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or THE SAME AS,. the ACTUAL steam. flow if during the power. increase to 100%, the associated steam pressure signal had stuck at its;50% value? I QUESTION 3.07 (1.50) What instrument signal is sent to the Train A PORV programming circuit to develop the pressure setpoint when operating in the Cold 09crpressure Pratection (COPPS) mode? (O. $O)- b What TWO other conditions, in addition to exceeding setpoint,.must { be met in order for a PORV to open automatically in the COPPS' mode?
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l l QUESTION 3.08 (2.00) l l List FOUR o+ the functions of the P-10 permissiv I l l
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QUESTION' 4.01 (' 1 '. 50 )
' Change 4 to Rev. 1 of Procedure OP 3204,;"At Flower Operation,," i I
l necessitates raising'ths. Steam Generator Low-Low level, trip; setpoi nts 'to ')/= 36. 6% pri or ' to achi evi ng. 70%'. reactor .. power . 'Why:
.. was this procedure change required? . ( 1. 00) ,
.q In conjunction'with the change in Low-Low level ' trip'setpoints,. 'the .
operating water level for.the steam generators ?was raised, to 58% Why was this higher' water level adopted? ! !
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QUESTIO .02 ( l '. 50 ) f Refer to attached Figure 7.1: - During a.resctor startup, you notice that the reactor has achieved criticality, and.that rod posttians are as follaws: I All shutdown ' banks f ull y wi thdraw Control' bank A fully withdraw Control bank B at 138 steps withdrawn.' .l
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Control bank C at 25 steps withdraw J Control bank D fully inserte State the THREE actions which are required by procedure'OP 3202
" Reactor Startup" in this condition, l
l- i ! QUESTICN 4.03 (2.00) '..;
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j i State the FOUR federal (10CFR20) quarterly expecure limits for manimum permissible occupational exposure for individuals' eighteen years or 1 olper, for whom current quarterly and lif eti me exposures are know . } Include in your answer the numerical limit and,the: effected portion of l the bod j
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QUESTION 4.04' (3.00) The f ollowing caution is 'f ound in AOP 3561, " Loss of, Reactor Plant Component' Cooling Water": j
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-1 On a loss of Reactor Pl ant Component Cooling Water (CCW), i if the :RCP thermal barrier cooling flow is lost AND the !
seal injection flow CANNOT be maintained greater than 6 gpm, then the reactor must be. manually tripped and the affected f
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RCP secure What is the basis for securing the RCP? I Considering the above caution, state TWO line-ups that can be used j to avoid a reactor trip /RCP trip if only one CCW pump is available i and seal injection flow drops below 6 gp (Valve numbers are-not required.) (2.00) QUESTION 4.05 (1.00) State the basis for the following precaution of OP 3335D, " Radioactive , Liquid Waste System": J Do not operate the waste evaporator during a plant cooldown when two RHR heat exchangers are in servic , l ObESTION 4.06 (1.00) What constitutes Adverse Containment? (Include specific parameters and values.)
I OUESTION 4.07 (2.00) The following concern EOP 35 FR-H.1, "Rresponse to Loss of j Secondary Heat Sink" ) l State THREE conditions, each of which require that the RCP's be stoppe (Acverse containment values NOT required.) (1.50) l l State ONE adverse consequence of NOT stopping the RCP's, a= recuired by this procedur (0.50) i i
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GUESTION 4.08 (3.00) Answer the following in accordance with,AOP 3566 " Immedi at e ' Bor ati on " : ?
' ' List.FOUR.of FIVE conditions that. require i mmedi ate' b'orati o (2.00) . Describe the TWO' flow' path's available>for:immediate boration,.
indicate the preferred and alternat ( 1'. 00 ) i
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.. . .{. PRINCIPLES'OF NUCLEBB_ POWER PL. ANT OFERATIONi Pagei18 l
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ANSWER- 1.01' (1.00) b.
L REFERENCE MP3, Reactor Operations lesson pl an , pp. 13-15- , Neutron' Sources and Subcritical Multiplication lesson plan, I pp. 11-22 l Objective-1986 RO, Reactor Theory .RQ'48-1
l 192008K104 .193006K110 193006K104 ..(KA's) l l l AN5WER 1.02 (1.00) , ;f 1 REFERENCE
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MP3, Reactivity Coefficients and Defects lesson plan pp. 8 & 20 Objective-1986 RQ, Reactor Theory 13 & 1 RO 9B-1, 192004K107 193003K125 ..(KA's) ANSWER 1.03 (2.00)' Temperature is unaffected by the dropped ro (0.50) Pressure'is unaffected by the dropped re (0.50) d
'). The reactor power will initiall y drop promptly CO.25] and then slowl y ~{
decrease CO.253 to a new steady state level as supported b subcritical multiplication Co.50]. l
REFERENCE MP3, Reactor Oprirations lesson plan, pp. 5-18 Delayed Neutrons lesson plan, p. 21 ,q Objective-1986 RO, Reactor Theory 3 ' 192008K112 192005K103 193009K107 ..(KA's)
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ANSWER 1 04 (1.00). ]1 ,
.0 REFERENCE MP3, Reactivity Coefficients'& Defects.: lesson p1an', p. 2 Delayed Neutrons lesson plan, p. 9 Objective-1986 RO, Reactor. Theory 10 and 1 .a
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REFERENCE
MP3, Xenon and Samarium lesson plan, p. 17 .
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Objective-1986 RO, Reactor Theory 3!. 1987 RO 3A- ) 192006K106 193009K102 ..(KA^s) ANSWER '. 06 (2.00) decrease no change
{ no change
.] decrease REFERENCE MP3, Reactor Operations lesson plan, pp. 33-35 Objective-1986 RO, Reactor Theory 3 RO C-18 :i i
192002K114 192002K113 192002K110 004000K519 001000K508- ) 001000K104 ..(KA's)
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:MP3,' Reactor. Operations lesson. plan, pp. 5-18 a Objective-1986 RG, Reactor Theoryc2 ,
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192OO8K110 OO1000K407 ..(KA's) ANSWER 1.08- (1.50)
.i ' , INCREASES' STAYS'THE SAME (1.00) GREATER THAN (0.50)
REFERENCE' MP3, Mitigating Core Damage, pp. 6-8 NSSS P:r Pressure & Level, pp. 20-23' Objective-1986 RO, Describe the operation of the p:r press. & level control sy C3OO1K103 012OOOK604' ..(KA's)
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ANSWER 1.09 (1.00) Decrease Increase REFERENCE MP3 Boiling Process lesson plan, pp. 24 & 25 MP3 1987 RO 9A-1 OOSK105 059000K405 ..(KA's)
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ANSWER 1.10 (1.50). calculated' higher than actual b, calculated higher than actual calculated higher than actual
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~ REFERENCE MP3, Pl.an t Cycles l esson plan , pp. 27 &,28:
Objective-1986 RO, Reactor 1 Theory 5 fa K108 193OO7K106 -002020K501 ..(KA's)
~ ANSWER 1.11- (1.00) GEFERENCE Steam Tables MP3 S.G. lesson plan, pp. 23-2 K12 ..(KA's)
ANSWER 1.12 (1.00) Decrease Decrease REFERENCE MP3 Fluid Properties lesson plan, pp. 25 & 26 MP3 1987 RQ 10A/8-4 j 191004K114 191004K106 191004K101 ..(KA's)
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2 PLANT DE@l@@_lyGLUQ1NQ_E6EETy_@ND_EMEBQENCYL .Page.22
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EYSI?d .. ANSWER 2.01 (2.00) . >17 percent P:r Leve . Letdown isolation valves .(LCV-459 and 460) ope . > . Loss of Powe . . ' Loss of Instrument Ai REFERENCE' MP3.'NSSS CVCS..p.'4 Objective-1987 RQ 18-2 ANSWER 2.02 (2.50)
' As pressure increases, a closing f orce is exerted on the top of'
the seal ring CO.5]. The narrowing between the' seal faces restricts the flow and increases the pressure f elt on - the underside of the seal face Co.5 The increased pressure pushes.the seal ring back up, opening the flow passage which allows more flow to escape [0.5], thus re-establishing a correct equilibrium posi tion.
, Through #2 seal to the CDTT E0.53 and the #1 seal-return line l relief valve to the PRT. CO.5] l l-l REFERENCE MF3 NSSS RCP, pp.6-9 MP3 NSSS CVCS, p 19 l Objective-1987 RO 2A-2 002000K602 ..(KA's) l ANSWER 2.03 (1.00) C.
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REFERENCE MP3 BOP Di esel Generator. Sequencer, p. 4 Objective-1986 RO,' Describe the operation of the Diesel Generator Sequence K411 064000K410 OOOO56K301 ..(KA's) ANSWER 2.04 (2.00) To protect the low pressure RHR piping C1.00] and. preclude.the possibility of uncontrolled'RCS depressuricationCO.50].to
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[0.25] or containment sump CO.25].
REFERENCE MP3 NSSS RHR, p. 2-4 Objective-1987 RQ 10C-1 OO5000K407 073OOOK401 ..(KA's) ANSWER 2.05 (2.50) Safety Injection signa Loss of Power signa Auxili ary f eedwater pump start signa (1.50) To limit the flow of water into a faulted steam generato (1.00) REFERENCE i MP3 NSSS AFW, pp. 7& 16 i Objcsctive-1986 RO, Describe the operation ofL the AFN syste l 061000K404 061000K101 061000K105 OOOO11K312 ..(KA's)
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i ANSWER 2.06 (1.00) Normal: Primary Grade Water System [0.2], and RWST CO.2]. j Emergency: Fire Protection Water System CO.2] - Preferred CO.2], and i Service Water System CO.2 ; I REFERENCE j M P ::.. 1987 Requal Ob jecti /es , , Item 10C/5 MP' procedure 3305, p . 033OOOK404 033OUOK401 194001K103 ..(KA's) ANSWEP 2.07 (2.00) 'I
Any four of the following: hfjj-fd _, ! Waste Neutralization Sump Monitct -Condensate Polishing Facilit y: j i.. u. , . .2m ,- _ u . a m a v_: &a m J L c. 2t d ? mi ner- Sum p:
' l d; u l, u p is o s v /a red.(WQf=.g ' Turbine Building Floor Drains: Turbine building sump effluent is q diverted to the turbine plant component cooling drain sum (Lw,- A6 ?0 , Liquid Waste Monitor,: Liquid waste effluent is isolated from the discharge cana (wc.ac.L sO l 4 Regenerate Evaporator MonitofCondensatePolishingFacility:)
Regenerate evaporator system effluent is-diverted to the regenerate ' evaporator feed tan (Sc.?.pSC0 Steam Generator Blowdown Mon i t or, : Steam generator blowdown is isolate b (cN.44 0.2 Each function: 0.25] (2.00) i
/M ! .'E s <'r vO (o ule , .s C aEach rc 1%moni, :rv .tor
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MP7 1987 RQ C-2B4 MP3 TS Table 3.5-1 MP! BOP Rad Monitors lesson plan, pp. 46-47 '/ ) * 9 %
,m ? 1 S., f (c h , src R'm%e m 1. t e e.> /< >s., n p k ,,, p .A / ;
073OOOKA01 OOOOO9K321 ..(KA's)
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ANSWER 2.08 ( 1. 00 )- 1-a 2-d 3g 4-f i REFERENCE l MP3 NSSS ECCS lesson plan, pp. 113-11/ Gbjactive-1987 Ru 3C-2
OU6020K603 OO6020K601 OOOO54K304 .. (KA's) l ANSWER 2.09 (1.00)
+; ,, y T w'd s ' cl e 4, C . 4y :
4 Sends an open signal to accumulator isolation valve CO.50] : Unblocks the low pressure SI signa .[0.503 3_ v ,, s ta S ,~ :n ir en ~ /, , e . . ./* o ,., e, > r e. i, e 4 p.es 2 :t '. . t c c., y & G wak s eaa, s r e. ~ use i n w h ri. < u n '>re** /r ee ,w;-rwe 4;$6 - ra re o f f. e s s <e e s. ye p ,3 . } f, V sie a k 3 l. * e Sd", ; ,,,:g
,m ., o , i res *.
ha.fa3sw~c ! REFERENCE l
MP3 1987 Requal Objectives, p. 3, Item 2A- j r1P3 "P:r. Prassure ?.e Level" Lesson Plan, . ' l 012OOOK604 0450006001 ..(KA's) l l l
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ANSWER 3.01 (2.00) With control level indicating high the actual: VCT level will drop CO.503 because (charging continues but) letdown 'is diverted from the VCT CO.50 The VCT will eventually be completely drained l CO.50] because the charging pump suction Will not shitt to the RWST-
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CO.50 REFERENCE MP3 NSSS CVCS, pp. 8&9 Objective-1986 PO, Describe the plar.t. response in the event of an instrument-failure with no operator. respons ;j 004000K605 004000K106 004000A301 004000A207 003000K614
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ANSWER 3.02 (2.00) All pressurizer heaters energiz (0,50) l Primary pressure risesCO.53 and.then stabilizes at the setpoint of the power operated relief val ves C O. 5 A single PORV will automatically openCO.5]. REFERENCE MP3 NSSS P:r Fressure and Level Instrumentation, pp. 6-16 Objective-1987 RQ 2A-3 1986 RO, Describe the plant response 10 the event of an instrumen failure 3a th no operator respons K607 194001K103 ..(KA's)
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ANSWER 3.03 (3.00) j (In Hot Standby Mode the Steam Dumps are being controlled in Steam-Pressure Mode.) Reducing the setpoint to 1007 psi will.cause.the , steam dumps to open to reduce pressureEO.503. . The Steam Dumps will- i close when primary temperature cools to the P-12 -(or lo-lo Tavg) 'l interlock setpoint (of 553 F)to.50]. The Steam Dumps will then 'I oscillate open and shut as- RCS Tavg oscillates around the.P-12 set / reset E0.50]. -Thus, final RCS Tavg will be approximately, 553 F (+/- 2 F)CO.503'. Steam dump cperation would be blocked CO.503. (Secondary pressure
. R would rise to the setpoint of the secondary pressure relief 1 valve, j which would operate to maintain pressure at 1125.psig). As a resul 'i the RCS Tavq will steady out at 561F (+/- 2 F) Co.50]. 3 I
REFERENCE Steam Tables .
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MP3 NSSS Steam Dump System, pp. 10-16 I Objective-1987 RQ 28-3 1986 RQ, Describe the plant response in the' event of an instrument failure with no operator respons J 039000K408 039000K404 039000K402 039000A204 045050K401 )
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ANSWER 3.04 (1.00) i
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Permits the feed regulating valve to provide fine control of feed I flow; (makes the feed pump speed respond slowly during and after secondary plant transients).
REFERENCE MP3 1987 Requal Objectives, , Item 3B- K405 001000G001 ..(KA's)
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ANSWER 3.05 (2.50) Used to block automatic rod withdrawals ( C -d , j Used to develop the P-13 signa . Used to generate Tref. ( LS Cdrol [G If, 6elW fdIc Used to generate. steam generator water level progra . Usea to generate a rate of change of power (in the automatic rod control circuit). u>, duller [~19 & o. so wQ REFERENCE Skeam Mrs CTret) \u A <9ed MP3 NSSS I & C Failures, pp. 41-42 Objective-1986 RO, State all the Instrument outputs, control functions and alarms f or PT-50 K407 001000K403 00005eK101 ..(KA's) ANSWER 3.06 (1.00) GREATER THAN REFERENCE MP3 NSSS SGWLC, p. 12 Objective-1986 RO, Describe the plant response in the event of an instrument failure with no operator respons A203 068000 GOO 1 ..(KA's)
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T8 Auctioneered low CO.25] wide range loop '>& CO.25 The Train A (arm / block) switen is in ARM. Co.50] ] The PORV (c l ose/aut o/open ) control switch is in auto. [0.50] l I i
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. REFERENCE MP3 NSSS P:r Pressure and Level, pp. 11-16 ,
Objective-1986 RQ, State all the Per Pressure Control outputs, control-f uncti ons., and alarm State the interlocks. 1987 RQ 10A/B- K403 194001A116 ..(KA's) 'l
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ANSWER- 3.08 (2.00)
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Any four of the following: Allows. manual blocking of the intermediate range high flux tri ! Allows manual blocking of the C-1 rod sto . Allows manual blocking of the low setpoint power range tri . Automatically restores intermediate range trip (when power falls below the P-10 setpoint). Automatically restores the low setpoint power range trip'-(when power falls below the P-10 setpoint). Automatically restores C-1 rod stop (when power falls below the-P-10 setpoint). Provides input to the P-7 circui . Serves as a back-up to P-6 (by preventing the operator from ! inadvertently y reenergining the source range high voltage with power above P-10).
REFERENCE i l MP3 NSSS RPSAS, p.67 Objective-1986 RO, State any RPSAS Interlock K610 194001A103 ..(KA's) i
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l ANSWER 4.01 (1.50) To assure that the amount of heatsink postulated in accident analysis is available to mitigate an accident (while an evaluation i of steam generator level instruments is conducted).
or i Because design inadequacies in the SG reference legs can result in I errors in indicated SG 1eve (1.00) I To lower the potential for challenging the SG level trips (b l Placing the operatung band between the high level turbine safety ; trip and the Low-Low level reactor trip). (0.50) l REFERENCE MP3 OP 3204 Change 4 documentatio MP3 LER 87-022-0 K401 194001K105 ..(KA's) I ANSWER 4.02 (1.50) I Terminate the startu . Drive rods i . Commence boratio REFERENCE MP3 1987 RQ, C-3E MP3 OP 3202, precaution 4.1 *** Include MP3 Figure 7.1 from OP 3202 " ROD BANK INSERTION LIMIT vs. TRERMAL POWER, FOUR LOOP OPERATION" with examinee's packag *** 001000G001 194001K111 ..(KA's)
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ANSWER 4.03- (2.00) i I .25 R/qtr to the whole body, (gonads, head and trunk, blood forming j organs and lens of the eye). j i l l .5 R/qtr to the ski !
'! .75 R/qtr to the extremitie i R/qtr to the whole body not to exceed a total. lifetime exposure i of 3(N-18) Re )
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REFERENCE ] MP3 SHP 4902, pp. 6&7 .g Objective-1986 RO, State the federal quarterly exposure 1.i mi ts . ] 1 194001K103 000036G003 ..(KA's)
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ANSWER 4.04 (3.00) f i (Loss of the RCP thermal barrier cooling flow and low seal injection I flow represents a serious challenge to the RCP seals.) In order to- d preclude, or at least minimite, damage to the seals it is necessary to secure the affected RC (1.00) l Can cross-connect the CCW supplies to the RCP thermal barrier . _ j thrcugh the CCW pump' suction and discharge valves (3CCP*V92,93.94,95 j and 3CCP+V7,8,9,10) E1.00] or through the containment header cross connect valves (3CCP*AOV179A,1798,180A,1808) [1.00 j REFERENCE MP3 NSSS. RCP p. 17, AOP 3561 step 2, OP 3301D step Objective-1986 RQ, State reasons behind steps of AOP A203 000026K303 016000G005 ..(KA's)
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ANSWER 4.05 .(1.00) To preclude exceeding flow limits (8100 gpm) on.a CCW trai or l l To preclude experiencing . adverse tube vibration af f ects on the: shell l side of a CCW heat exchanger.
l REFERENCE l MP3 OP 33304 p. . 33 Objective-1986 RO, State reasons behind no.tes and precautions.of op GOO 5 ..(KA*s). j ANSWER 4.06 (1.00) Containment temperature [0.25] >180 F EO.253, containment radiation [0.2SJ >10E5 R/Hr CO.23 REFERENCE MP3 1987 Requal' Objectives, p. 30, Item 10C- OOOO11K312 OO6000 GOO 5 ..(KA's) ANSWER 4.07 (2.00) . If total feed flow to S/G's cannot be maintained >525 gp . If WR level in any 3 S/G's is <39% ( 54 adv. cont.). If p:r. pressure >/= 2350 psi (1.50) Drvout of the S/G's will occur-earlier (less time available fo establishing secondary heat sink or RCS feed and bleed);
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Causes RCS feed and bleed to be less effectiv ( 0. 50) .
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REFERENCE 1 i MP3 1987 Requal Objectives, , Item 2B- 'l MP3 EOP 35 FR- i MP3 EOP Development Training Text HO EOP 35 FR-H, pp.21-2 j
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000054K304 ..(KA's) I ANSWER 4.08 (3.00) 1 Any FOUR of the following: Rod height below the Low-Low limi t each) . Failure of one or more rod clusters to f ully insert. cn a reactor trip or shutdow . Uncontrolled cooldown following a reactor trip or shutdow Uncontrolled or unexplained reactivity increas . Failure of the Makeup = system to borat I i j Preferred - Boric acid tanks to BAT pumps to Immediate boration valve (MV 8104) to suction of charging pumps. [0.503 ' Alternate - Boric acid' tanks to Gravity boration valve (MV 8507 ALB) to suction of charging pumps. [0.50] I' REFERENCE MP3 AOP-3566 pp. 2-3 Objective-1987 RO 9A-3A, 1987 RO 3B-2 004000K116 OOOO29G011 ..(KA's) 1
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U. S. NUCLEAR REGULATORY COMMISSION
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SENIOR REACTOR OPERATOR'REQUALIFICATION EXAMINATION l FACILITY: MILLSTONE 3 y
I REACTOR TYPE: PWR-WEC4 I
, DATE ADMINISTERED: 87/09/04 copy o+
EXAMINER: JENSEN, CANDIDATE I l INSTRUCTIONS TO CANDIDATE: l
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I Read the attached instruction page carefully. This examination replaces j l the current cycle facility administered requalification examinatio i Retraining requirements for failure of this examination are the same as ' l for failure of a requalification examination prepared and administered'by j - your, training staf Points- for each question are- indicated in l I 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 four (4) hours after the examination start '
% OF l CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY !
15.00 25.00 THEORY OF NUCLEAR POWER PLANT 2 OPERATION, FLUIDS,AND THERMODYNAMICS 15.00 25.00 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 15.00 25.00 PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 15.00 25.00 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
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60.00 % Totals Final Grade All work done on this examination is my ow I have neither given nor received ai . . -. _ _ _ _ _ . _ - _ _ _ _ . - _ , candidate's signature
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NRC RULES' AND' GUIDELINES' FOR LICENSE: EXAMINATIONS:
- During the administration of this examination the following rulesEapply: Cheating on the examination means an automatic'denialoof.your application and could result in more severe penaltie ' Restroom trips are to be limited and.only one candidate at astime.may, leav You must avoid'all contacts ~with~anyone'outside.the' examination room to avoid even the appearance or possibility of cheatin . Use black' ink or dark pencil.only'to facilitate legible reproduction . Print your.name in.the blank.provided'on the cover sheet ofethe examinatio . Fill in the date on the cover sheet of.the examination?(if necessary). Use only the-paper provided for answer . Print your name in the upper right-hand corner of the first page of each:
section of-the answer shee . Consecutively number each answer sheet, write "End of Category __" as appropriate, start'each category on a'new page,Lwrite only on one side of the paper, and write "Last:Page" on'the last answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least three lines between each answe i 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in' facility literatur 'l 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 require '! 14. Show all calculations, methods, or assumptions used to obtain an answe to mathematical problems whether indicated in the question or u no . Partial credit may be given. Therefore, ANSWER ALL' PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl ; 17. You must sign the statement on the cover sheet that indicates that the
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work is your own and you have not received or.been given assistance in completing the examinatio This must be done after the examination has been complete . .. _ _ _ _ --
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(1), Exam questions on to ..
.(2)- Exam aids - figures, tables, et .(3)
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Answer pages including figures.which are-part:of the. answe Turn in yo'ur copy of the examination an'dLall pages.used/to ai2swer-the examination. question ,
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, Leave'the examination. area, .as defined by th'e: examine ,If a hter-L l . leaving, you are found in this area'while the-'examin'ation.is'still- 'l in progress,.your license may be denied or revoke ,
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* THEORY'OF: NUCLEAR POWER PLANT OPERATION, Paso 1 FLUIDS,AND THERMODYNAMICS .
QUESTION, 5.01L (1.00) For each of the conditions listed below, state whether'the moderator temperature coefficient becomes MORE NEGATIVE, LESS NEGATIVE, or REMAINS THE SAME. Assume all other conditions are' unchange Control bank D is withdraw I Core age increase R
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I QUESTION 5.02 (1,00)
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State how each of the following will' affect the value of. differential: J Boron worth, assuming all other conditionsLremain unchange Limit your answer to LESS NEGATIVE, MORE NEGATIVE,ior NO EFFECT., q l Reactor coolant temperature decreases, Boron concentration increase l QUESTION 5.03 (1.00) State how each-of the following will~ affect the value of shutdown margin, assuming all other conditions remain unchange Limit your~ answer to INCREASE, DECREASE, or NO EFFEC Reactor coolant temperature decreases, Xenon concentration increase , t
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* THEORY OF NUCLEAR POWER. PLANT OPERATION, 'Page' 5 FLUIDS,AFD THERMODYNAMICS .
QUESTION 5.04 (1.00) Multiple Choice Which one of the following statements'MOST CLOSELY describesEt he va'lue of' Xenon reactivity following:a reactor' trip from 100% power? Assume equilibrium BOL condition Approximately~24 hours after the trip, Xenon reactivity-worth-will:be approximately 4500 pc Approximately 6 hours after.the trip, Xenon reactivity worthiwil be approximately 5650 pc Approximately 8 hours after the trip, Xenon reactivity worth will be approximately 4500 pc Approximately 8 hours after the trip, Xenon reactivityEworth wil be approximately 5300 pcm.
QUESTION 5.05 (1.00) l Multiple Choice During a.startup it was determined that Keff was equal to 0.9 when the- !
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Source Range (SR) instrument was reading 50' cp What would the source range instrument be reading if rods were withdrawn to bring Keff equal to-0.9757 Assume BOL condition cps cps cps cps l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE.*****). . _ _ _ _ _ _ _ _ _ . _ 1
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THEORY . 0F NUCLEAR POWER PLANT. 0PERATION , -Page"li FLUIDS,AND THERMODYNAMICS-
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-QUESTION- 5.06 :(1.00)
. Multiple Choice During a reactor trip recovery, the initial 1/Midata point was~1. .
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Upon. stopping rod withdrawal.to take 1/M data, you find'that.the second 1/M point i s . l .1' . Which'of the following explains.this increase in the 1/M ~ value? ~ f This'is not possible, the RO must have made an errorfwhen taking count rate dat The buildup of Xenon during the.1-hour delay added more negativ reactivity than'the rod' withdrawal had'added in positive reactivit The source-detector geometry is incorrec An inadvertent dilution is in progres QUESTION 5.07 (1.00) Multiple Choice What is the startup rate if i,wer increases from 3000 cpsfto 8000 cps-i twenty seconds? (Choose the MOST CORRECT answer.) .4 DPM .7 DPM
' .0 DPM .3 DPM
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' THEORY OF' NUCLEAR POWER PLANT. OPERATION, :Page 7 FLUIDS,AND THERMODYNAMICS '!
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QUESTION 5.08 -(1.00) Multiple Choic Which one of the following statements is? correct?' ! H .With all other conditions constant,Tthe reactor respondsLMORE- i QUICKLYLto a given reactivity change'at EOL.than:at BOL, because ! the value'of' Beta-bar effective is GREATE I With all other conditions constant, the reactor responds LESS
.QUICKLY'to a given reactivity change at EOL than'at BOL, because the value of Beta-bar effective is GREATE With'all other conditions constant, the. reactor responds MORE QUICKLY to a given reactivity change at EOL than at BOL,.because:
the value of Beta-bar' effective is' LOWE With all other conditions constant, the reactor responds.LESS QUICKLY to a given reactivity change at EOL.than at BOL, because the value of Beta-bar effective is LOWE QUESTION 5.09 (1.00) I l True or False 7 Xenon oscillations are more likely to be DIVERGENT as the core ages, because fuel is depleted from the center. regions of the core more-rapidly than from the outer regions.
! ' The primary method of dampening Xenon oscillations is to follo secondary load changes by boration and dilution while holding control rod position constan QUESTION 5.10 (1.00)
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State how an INCREASE in each of the following parameters ~affects'the Departure from Nucleate Boiling Ratio (DNBR). Limit your answer to-INCREASE, DECREASE, or NO EFFEC Coolant temperature, Coolant flo (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ,
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cause water hammer? R d Sudden closure of-a valve'-in a system.in whichfthere-is water: flo q
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QUESTION 5.12 1 l]i
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Multiple.. Choic Assume the plant is in Mode'3 at'a temperature of1535 F, and the-steam: - dumps are NOT operable. To what value,must Tavg1 rise before. causing- ! the power-operated. steam generator pressure. relief. valves'(MSS *PV-20A,- l B, C,.& D) to-lift?- Assume the. pressure sotpoint controllers are set-for normal power operation. - (Choose the'MOST. CORRECT answer.) .1 F .7 F
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FLUIDS,AND THERMODYNAMICS
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QUESTION- 5.13 (1.00) ,
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Multiple. Choice Indicate whichLitem below does NOT' ENSURE that the Enthalpy? Rise' Hot! w N Channel, Factor-(F. Delta-H)' remains within prescribed' limits': ' Control rods in group move together within-+/ 12 step Axial flux difference-(AFD) is maintained within: limit Tavg vs. Tref are kept matched toiwithin:5.0F.- I ' Control rod groups.are properly' sequenced'and overl'pped.a I QUESTION 5.14 ( 1. 00 -)
State how an INCREASE in the following parameters affects Net' . . Positive Suction' Head (NPSH).available at the suction of a centrifugal" pump. Limit your answer to INCREASE, DECREASE, or NO EFFEC System flow rat System temperature.
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* THEORY OF NUCLEAR POMER PLANT OPERATION, Page 10 FLUIDS,AND THERMODYNAMICS ,
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i QUESTION 5.15 (1.00)' l
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Multiple Choice Choose the CORRECT definition of Axial Flux Difference (AFD). The difference in normalized flux signals between the maximum upper excore detector calibrated output and the minimum lower excore detector calibrated outpu b. The difference in normalized flux. signals between the minimum upper excore detector calibrated output and the maximum lower l excore detector calibrated outpu A c. The difference in normalized flux signals between the' top and' l bottom halves of a two section excore neutron detecto d. The ratio of the maximum upper excore. detector calibrated output ; to the average of the upper excore. detector calibrated output I l i I l ' l l ' I
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* PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Pace _ll ,
QUESTION 6.01 (2.00) i State the SIX rod control system interlocks which inhibit outward rod movemen Indicate whether each is effective in the MANUAL-mode, AUTOMATIC mode, or BOTH. Include setpoints and coincidence as applicabl QUESTION 6.02 (1.00) Assume steady state operation at 100% power when-the Master Pressure Controller setpoint for the pressurizer is inadvertently changed from 2250 psig to 2385 psig. Assume a step change in setpoint ) and assume that pressurizer pressure control is in automatic, i What automatic action (s), other than the actuation of alarms / annunciators, will occur immediately? (0.25) Describe the pressurizer pressure transient that will occur if no operator action is take Include in'your answer any other ' automatic actions, other than alarm / annunciator actuations, that take plac (0.75) 4 i l QUESTION 6.03 (0.50) With regard to the main feedwater pump speed control circuitry, what is i the reason that the output from the steam flow summing amplifier is conditioned within a lag circuit?
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QUESTION 6.04 (1.00) What is the design intent (basis) for the interlocks and automatic functions associated with the RHR loop suction valves (MV 8701A/B)? QUESTION 6.05 (1.50) List SIX of the eight conditions which must be satisfied to energize the white " Ready To Auto Start" light on MB-8 for a diesel generato (Do NOT include power available to indicating lamp.)
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QUESTION 6.06 (2.00) A leak develops in the reference leg associated with the automatic level controller of the Volume Control Tank (VCT). As a result the indicated level in that leg fails high. Describe the VCT level I transient assuming that no operator action is taken and that the VCT is in the automatic makeup mod Include the reasons WHY level change LNere ' T ke o c.< - . .u y i . > u .. <t 6 rt.- rei r i, e <e .- < i e ~ ', J;< real rin r- ,e,> r e ;., The s cie d w- r ear c e s , e h, / , a < e s ., /r e t. e , ,w , , r l }er V:/ sI.n ne/ f.,ils ). , y +,, ,, J rm te . ., m , a ., J . J s re w,m:., n;,.,en .-.A:,,,is re,. . ,.,J cl ,f rt.e p e,,,,, . s ., a.,.) , , ,1.,:... br "#- 3 QUESTION 6.07 (2.00) , What would be the MAIN consequence of inadequate spray bypass flow to the pressurizer, including the effect, the component (s) affect- ; ed, and the reason this occur (Do not discuss boron or tempera- i ture equalization.) (1.50) i State the provision made in the control room to warn the operator I of insufficient spray bypass flow, including any applicable setpoin (0.50)
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QUESTION 6.08 (2.00) Assume the plant is shutdown at 500 F and 2000 psig and the Cold i Overpressure Protection System (COPPS) is inadvertently arme De-scribe how EACH TRAIN of COPPS will respond if a loop wide range Tc instrument fails lo INCLUDE the inputs to each COPPS train, AND whether or not that train's associated PORV will open to reduce plant l pressur QUESTION 6.09 (1.00) State the FOUR sources of makeup water to the Spent Fuel Foo Identify between normal and emergency (or "last resort") sources of makeup, AND state any preferential order of use for the normal and/or the emergency sources.
l l QUESTION 6.10 (2.00) List FOUR of the h$ e radioactive liquid effluent monitors which have automatic control functions upon activation of high radiation alarms, AND briefly describe the control function for each monitor listed.
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AND RADIOLOGICAL CONTROL i' QUESTION 7.01- (1.00) What constitutes Adverse Containment? (Include.spe'cific parameters and values.)
'l i QUESTION 7.02 (1.50) H'
'An operator-needs to' enter a posted neutron' radiation"areaLwhere.'the total measured neutron' dose is 1.8 mr/h '(The measured beta-gamm ,
i radiation levels are insignificant.) He will remainiin the area"fo y 1.5 hour Is a Radiation WorkLPermit'(RWP). required?. Briefly:EX--
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QUESTION 7.03 (2.00) , '! State the FOUR SI Termination Criteria as"listedEin EOF 35 E-l', " Loss-of Reactor or Secondary Coolant." Be specific,,and. include /ALL_ , listed options for'each criterion. ' Adverse containment valuestar NOT require ' !,
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QUESTION 7.04 (2.00) ;
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The following concern EOP 35 FR-H.1, " Response to:LossLof Secondary Heat Sink": ,, State the THREE conditions, each'of which require that the RCP's be stoppe (Adverse' containment values NOT required.) :( 1. 50 ) i' State ONE adverse consequence of NOT stopping the'RCP's, as-required by this procedur .(0.50)
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l QUESTION 7.05' (1.00) l The following concern procedure OP 3204, "At Power Operation": In case of an emergency load reduction request from Convex, at what rate shall load be'shed? '(0- 50 ) ! . ' What guidance is given concerning maintaining AFD within the I
target band during an emergency load reduction?- (0.50).
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* PROCEDURES - NORMAL,LABNORMAL,1 EMERGENCY Page l'41 ,
AND RADIOLOGICAL CONTROL' QUESTION- 7.06 (1.50)- According to.0P 3301D, " Reactor Coolant Pump Operation",fstateethe THREE conditions.which-require determining that the pump shaft..is free,i (by. manual rotation of-the shaft), prior to pump star QUESTION 7.07 (2.50) A' maintenance man is working inside 'the L containment while the reactor is at powe He is. working in a radiation field of 500 mrem /hr'gamman and 45 mrad /hr combined (thermal and fast)' neutron. The mantis 35 years old and has a lifetime exposure'through'last quarter of 81.0' Rem on his NRC Form 4. Additionally, he has accumulated 1750 mrem so far this quarte How long can the man work in the area beforeLhe exceeds his 10CFR20 limits?. Show all work and state all assumptions. Round (1.50)
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off your answer to the nearest. minut During a declared emergency, this-individual-. volunteers toLenter-a high radiation area and perform work necessary-to prevent-furthe effluent releas In accordance with the MP3 Procedures, what is his maximum allowed whole body exposure?- (0.50)'< In accordance with the Health Physics' Procedures, whos authorization (by job title / position) is required in part b7 (0.50) QUESTION 7.08 (0.50) Change 4 to Rev. 1 of Procedure OP 3204 "At Power Operation'," neces-sitates raising the Steam Generator Low Level Trip setpoints to >/= 36.6% prior to achieving 70% reactor powe Why was this procedure change required?
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QUESTION 7.09 (1.00) Refer to attached Figure.7.1: During a reactor startup, the Reactor Operator reports.that the ) reactor has achieved criticality, and that rod positions are as follows: All shutdown 1 banks. fully withdraw Control bank A fully withdraw Control bank B at 138 steps withdraw Contrcl bank C-at 25 steps withdraw Control bank D fully. inserte State the THREE actions which are required by procedure OP 3202
" Reactor Startup" in this conditio QUESTION 7.10 (2.00)
State the FIVE indications which are used to verify natural circu- , lation flow in. procedure EOP 35 ES-0.1,." Reactor Trip Response," l if it is determined that NO Reactor Coolant' Pumps.can.be starte .
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QUESTION 8.01 (1.50) I
The shift supervisor (SS)'is' responsible for. complying with-the limit )
.and policies set forth in ACPJ6.03, " Radioactive Liquid' Waste Discharge' j Policy." . State the other TWO i-tems for which-the SS is responsible, .i according to ACP'6.03, regarding radioactive liquid waste discharg )
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QUESTION .8.02 (2.00)- nl The following concern information~found in MP3:EPIP 4701, " Unit" !: Incident Assessment,. Clarification, And Deportability": .!
~ State the NRC emergency event' classification levels'AND their l corresponding state posture codes, in order from the least ';
to the most sever (1.00): What TWO determinations are made using'the State of: Connecticut . , j
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Posture Code System in state 1and~ local emergency plans?- ( l'. 0 0 )'
QUESTION 8.03 (2.00)
The following concern information found in MP3. procedure ACP 1.19, l
" Overtime Controls For Personnel Working At:The Operating' Stations": l In cases of overtime to be worked in excess of established limits, f '
briefly describe the THREE overtime situations for which only the-first-level supervisor's approval is require (1.50) ! Include in your answer whether each situation described INCLUDES, or l EXCLUDES shift relief / turnover tim (0.50) i l l ' QUESTION 8.04 (1.50) ,
.. t Which individuals shall be allowed access to the Control Room during emergencies, according to ACPl6.01 Control Room Procedure? (List-
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THREE categories of individuals.)
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QUESTION 8.05 (1.50) The following refers to procedure OP 3250, " Removing Equipment From Service For Maintenance."
Assume Train A of Safety Injection is being removed from service during Mode 1 operation to repair a pump cooling water lea State the operators' responsibility with regard to the OTHER Safety Injection train (Train B). (1.00) QUESTION 8.06 (1.00) State the requirement in the MP3 Technical Specifications which exists to minimize the possibility of radioiodine release to atmosphere in the event that an irradiated fuel assembly were to rupture, while seated in its storage rack in the spent fuel storage poo Specifically INCLUDE the parameter monitored, j AND the required value for the paramete e QUESTION 8.07 (2.00) State FOUR of the five bases for the technical specification requirement that the lowest RCS loop Tavg be >/=551 F whenever the reactor is critical, State the TWO surveillance items which must be performed to ensure that this requirement is me I i
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QUESTION 8.08 (2.00) j Assume that the plant is in Mode 6 during a refueling outage with ! water level 18 feet above the reactor vessel flange, and a-gradual cooldown'of RCS water in progress. No movement of core components is in progress in the containment. However, irradiated f fuel is being moved in the spent fuel pool. The "B" emergency diesel { generator (EDG) is disassembled for overhaul. Surveillance testing j has just been performed on the "A" EDG, and it is determined that, j due to an electrical problem, the "A" EDG will not automatically j start on a loss of power signa l In accordance with section 3.8.1.2 of the Technical Specifications,
"A.C. Sources - Shutdown", state the FOUR actions which are required j to be performed immediately in THIS situatio l QUESTION 8.09 ( .1. 50 ) i i
Answer the following in accordance with Technical Specification-3.5.1, " Emergency Core Cooling Systems - Accumulators": I State the action which is required to be performed IMMEDIATELY, l in the event that an ECCS accumulator is declared inoperable due to its isolation valve being shu (0.50) What TWO actions are required to be performed if ECCS accumulator water boron concentration has been 2400 ppm for greater than one hour? (1.00)
(***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********)
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: THEORY '0F N CLEAR POWER PLANT OPERATION,:-
Paga;-19 FLUIDS,AND. THERMODYNAMICS '
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' ANSWER- 5.01 -(1.00) >
q 'Less negative More negative > i REFERENCE' MP3.1986 Requal Objectives, Reactor Theory,q.#15 & #1 MP3 '.' Reactivity. Coefficients & Defects". Lesson Plan ~Rev.;1,'.p . 192004K106 ..(KA's) q il ANSWER 5.02 .(1.00) ; More negative ) Less negative '
-d REFERENC MP3 1986 Requal' Objectives, Reactor Theory, #2 MP3 " Neutron Poisons" Lesson Plan Rev. 1, p. 12, Item :
192004K110 192004K109 . .(KA's); . l
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ANSWER 5.03 (1.00) E "'T b Decrease C4 JA.rgv~ay' .r ,c,enie li ? a r t% e t~) ~ i V * * T' E^'"'"*~*""~*
. .;, 5,v,,,,,,,; n,,,7 o f .. .fn',o j
- Increase n a g,, H .,,e ;;; W g:,a ,, U t
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MP3 1987 Requal Objectives, p. 10. Item C-1B ' MP3 " Reactor Operations" Lesson Plan Rev. 1, pp. 33-3 K114 ..(KA's) ;
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ANSWER 5.04 (1.00) ., j j d.
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FLUIDS,AND THERMODYNAMICS
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REFERENCE 1
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MP3 Curve Book Cycle 1, OPS Form 3209-6 Rev..0, Page 2 of ; MP3 1986 Requal Objectives,LReactor Theory, #3 { 192006K112 ..(KA's) i
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ANSWER 5.05 (1.00) C l REFERENCE ; MP3 1986 Requal Objectives, Reactor Theory, # MP3 " Neutron ~ Sources & Subcrit. Mult." Lesson Plan Re , p. 1 K102 192004K107- ..(KA's) ANSWER 5.06, (1.00) l b REFERENCE MP3 1986 Requal Objectives, Reactor Theory, # MP3 " Reactor Operations" Lesson TextLRev. 1, p. 1 K106 192008K103 192004K107 ..(KA's)-
ANSWER 5.07 (1.00) i REFERENCE MP3 1986 Requal Objectives, Reactor Theory, # l MP3 " Delayed Neutrons" Lesson Text Rev. 1, p. 1 < 192003K109 192004K112 ..(KA's) ! I i ANSWER- 5.08 (1.00) l
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' . . * THEORY OF NUCLEAR POWER PLANT OPERATION, Page 21 l FLUIDS,AND THERMODYNAMICS .
REFERENCE ! MP3 1986 Requal Objectives, Reactor Theory, #1 ( MP3 " Delayed Neutrons" Lesson Plan Rev. 1, pp. 7-1 . 192003K107 192005K107 ..(KA's) ANSWER 5.09 (1.00) True False REFERENCE MP3 1987 Requal Objectives, Item C-1A MP3 " Xenon and Samarium" Lesson Plan Rev. 1, pp. 21 - 2 K106 192005K107 ..(KA's) ANSWER 5.10 (1.00) Decrease Increase l REFERENCE > MP3 1987 Requal Objectives, p. 25, Item 9A-1 MP3 " Boiling Processes" Lesson Plan Rev. 1, pp. 24-2 l-193008K105 192005K109 ..(KA's) l AN54ER 5.11 (1.00) ! l f l \ - j REFERENCE l MP3 1987 Requal Objectives, , Item C-1C MP3 " Plant Processes" Lesson Plan Re , pp. 26 - 2 j 193006K110 193006K104 192005105 ..(KA's) 1 ANSWER 5.12 (1.00) b.
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j l REFERENCE - Steam Tables-- MP3 "S.G." Lesson Plan.Rev. O, pp. 23-2 J 193003K125 192008K124 192008K121' ...(KA's) j J ANSWER 5.13 .(1.00) .
'l1 ;l REFERENCE l
MP3 1986 Requal objectives, Heat Transfer and Fluid' Flow, #2 d No Facility Reference Identifie ) 193009K107 192003K111- ..(KA's) , ; l k ANSWER 5.14 (1.00) Decrease i Decrease j REFERENCE s MP3 1987 Requal Objectives, p. 29,' Item 10A/B-4 MP3 " Fluid Properties" Lesson Plan Rev.-l',.pp. 25-2 ' 191004K114 191004K106 191004K101' 192003K101 ..(KA's) I1 l ANSWER 5.15 (1.00) I e REFERENCE MP3 1986 Requal Objectives, Heat Transfer and Fluid Flow, #2 MP3 TS, Definition K102 192006K108~ ..(KA's)
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* PLANT' SYSTEMS. DESIGN,ECONTROL,iAND. INSTRUMENTATION ;Pago .23; :;
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Lh I ANSWER' 6.01- (2.00)' ' _; i C-1: (Intermediate range' overpower)'- 1/2L[0.118] intermediate? range- H channels. exceeds-20% current equivalent. power [0.118]i bothL[0.118]. j i C-2 (Power range?high1 flux) - '1/4 : [0 '.118] power! range channels exceeds 103% power _[0.118], both [0.118]'. .
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0 C-3-(OT Delta-T) - 2/4 [0.118].OT Delta-T channelsLwithin'3% of 4' their (continually' variable) trip ~setpoint'[0.118]; both-[0.118]'.
' C-4 (OP Delta-T):- 2/4 [0.118] OF Delta-T channels-within 3% of; ?
h their (continuallyLvariable) triptsetpoint [0,118], both-[0.118]. , l)
' C-5 (Turbine power) - One channel (PT-505) [0.118]f.turbinei o impulse pressure indicates.less than 15%' power?[0.118],.
auto [0,118]. 'C-11c(bank D withdrawal limit) - Control bank D at 223 steps; y
[0,118], auto [0.118]. !
REFERENCE MP3 1987 Requal Objectives, p. 5, Item 3A- MP3 " Rod Control" Lesson Text Rev. 0, pp.'62-6 K407 192006K103 ..(KA's) ANSWER 6.02 (1.00) All pressurizer heaters energiz (0.25) Primary pressure rises [0.25] and then-stabilizes.at the setpoint of
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the power operated relief valves [0.25]. 1 A single PORV will automatically open[0.25]. ,
REFERENCE MP3 NSSS Par Pressure and Level Instrumentation,.pp. 6-16- ' Objective-1987 RQ 2A-3 1986 RQ, Describe the plant response inithe event of'an instrume
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failure with.no operator respons l 010000K607 192001K102 ~ ..(KA's)
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l ANSWER 6.03 (0.50) Permits the feed regulating valve to provide fine control of feed flow; (makes the feed pump speed respond slowly during and after secondary plant transients).
REFERENCE MP3 1987 Requal Objectives, , Item 3B- MP3 "SGWLC" Lesson Plan Re O, pp. 18-1 K405 192002K112 192002K110 ..(KA's) ANSWER 6.04 (1.00) To protect the low pressure RHR piping [0.50) and preclude the possibility of uncontrolled RCS depressurization[0.25] to the RWST l [0.125] or containment sump [0.125]. REFERENCE l MP3 NSSS RHR, p. 2-4 Objective-1987 RQ 10C-1 005000K407 192007K104 ..(KA's) i ANSWER 6.05 (1.50) Transfer switch not in MAIN . 86 HBU backup lockout relay rese . 86 HP primary lockout relay rese . Start failure relay not energize . Control power available to stop circui . Shutdown relay not energize i 7 ., Mechanical trip circuit control power available, i Barring device relay not actuate [Any 6, @ 0.25 ea.) j REFERENCE MP3 1987 Requal Objectives, p. 7, Item 4A- MP3 " Diesel Gen. & Support Systems" Lesson Plan Rev. O, pp. 28-2 G007 193008K105 ..(KA's)
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ANSWER 6.06- (2.00) With control level indicating highLthe actual VCT level will drop [0.50] because (charging continues but) letdown is. diverte from the VCT [0.50]. The VCT.will eventually-be completely draine'd
[0.50] because the charging pump suction will not shift to the RWST- [0.50],
REFERENCE MP3 NSSS CVCS, pp. 8&9 . Objective-1986 RQ, Describe the plant response in the: event of an
' instrument failure with no' operator respons K605 004000K106 004000A301 004000A207 193009K107- ..(KA's)
ANSWER 6.07 (2.00)
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Co. r 3 Thermal shock would occur [0.5] to the spray piping [0.25] :nd the-eprey no::10 [0,25} because the piping downstream of the spray valves would cool to conta$nment ambient temperature [0 25] and then be subjected toj$50' nur F) water when spray flow started [0.25] og , l Pressurizer spray line low' temperature alarm [0.25], which is l set at 530 F [0.25]. . v,.r' To rhe spy r,o +e4 ECC1 j,eca,.sc. ri,e REEERENCE * * * I O 4 <>se 'ru ! & 7e Ymrx of 'r he
*"d ri."< Le e a p<wd 1 <w i g '.,r,;,,,,,.,,. n u,,, , j'arlo.6)
,,, ;
, MP3 "Psr & PRT" Lesson Plan RevI t[W8'. " " d 'a s/'y /4~ J ra < rec /BdJ L ! 010000K401 193010K101 ..(KA's) l ANSWER 6.08 (2.00) ' Train A inputs (PCV 455A) are auctioneered' low Th (WR) [0.33] and wide { range pressure (PT 405) [0.33]. Therefore, Train A PORV will not open
[0.33]. Train B inputs (PCV 456) are auctioneered low Tc (WR)-[0.33]-
and wide range pressure (PT 403) [0.33]. (Since COPPS is armed and-the Train B pressure setpoint is low due to the failed Tc instrument,) PORV 456 (Train B PORV) will open [0.33).
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l REFERENCE MP3 1987 Requal Objectives, p. 28, Item 10A/B MP3 " Par. Pressure & Level" Lesson Plan'Rev. 0 . . pp . 11'1 K403 .193010K105' ..(KA's) ANSWER 6.09- (1.00) , Normal: Primary Grade Water System [0.2],. and RWST [0.2]. f Emergency: Fire Protection' Water System-[0.2] - Preferred [0.2], and i Service Water System [0.2]. REFERENCE- 'j l MP3 1987 Requal Objectives, p.'.30, Item 10C/5 { MP3 procedure.3305. pp. 9-1 )
'033000K404 033000K401 193010K106 ..(KA's) 1 i
ANSWER 6.10 (2.00) [cN&Nh Waste Neutralization Sump Monitor ondensatePolishingFacility)'
-d drains, Ssg Aezili2rr d:s ek rg conden :te f.Q,e Q'"-)rted-to-the-aeratad e os .settree '
y m ,.4 Turbine Building Floor Drains: 3 Turbine building sump effluent is ;l diverted to the turbine plant component cooling drain sump'.
(% a 7o )- . . .
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Liquid Waste Monitor: Liquid waste-effluent is isolated from the
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discharge cana i.wc onst) Regenerate Evaporator Monitorg(Condensate Polishing Facility:) Regenerate evaporator system effluent is diverted to'the regenerate i
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evaporator feed tan .
(s naws) Steam Generator Blowdown Monitor:o Steam generator blowdown is l isolated.
l [Each monitor: 0.25; each' function: 0.25; . total of four each require Functionco-,must match monitor.] 1 6. 11u ,;/h .y G, . de o s a re A.,; re (c.us .ga t n) ; k ,I:.,9 i Jen s te ;, ,/; ' <.- mi r , 7 t, e j REFERENCE w ~ red d o*'<> lo* r* rk' ' ~ * le s ? *~ * !d hy s ~ f ). 1 ' MP3 1987 Requal Objectives, p. 15, Item C-2B MP3 TS Table 3.3-1 MP3 BOP Lesson Plans, Rad Monitors, pp. 46-47 j Y3+ K401 193001K103 ..(KA's) , in ? ? 6cP Le>w.* five y lo ,h, are Dem * e 'sll t en, ff A I - al F
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ANSWER 7.01 (1.00) Containment temperature [0.25] >180 F.[0.25], containment: j radiation..[0.25] >10E5 R/Hr [0.25]. 1 REFERENCE ! MP3 1987 Requal Objectives, p. 30, Item 10C- No Facility Reference Identifie K312 193003K125 193003K117 ..(KA's) l ANSWER 7.02 (1.50) l No [0.5]. .An RWP is required if entry is to be made into a , ' posted neutron radiation area where neutron radiation >/= 2.5 mr/hr I exists [1.0]. .I REFERENCE
.a MP3 1986 Requal Objectives, HP Procedures Objectives, # MP3 SHP 4912, par . K103 193008K123 193008K122 ..(KA's)
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7.03 i ANSWER (2.00) l I RCS subcooling (based on core exit TC's) [0.20] >30 F [0.20] ('90 F for adv. cont.). Total feed flow to intact S/G's [0.20] >525 gpm [0.20), or NR level in at least'one intact S/G [0.20] >4% [0.20] (34% for adv. cont.).
(Both options -- i.e. feed flow & NR level -- required for full :
credit.) RCS pressure [0.20] stable [0.10] or increasing [0.10]. (Both options required for full credit.)
4. Par. level [0.20] >7% [0.20] (50% for adv. cont.). l l
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. * PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 28 .
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REFERENCE J MP3 1987 Requal Objectives, p. 2, Item 1B- ; MP3 EOP 35 E-1, " Loss of Reactor or Secondary Coolant" Rev. 1, Step l 000009K321 ..(KA's) i
l i ANSWER 7.04 (2.00) 1 . If total feed flow to S/G's cannot be maintained >525 gpm.[0.5] ) If WR level in any 3 S/G's is <39% (54% adv. cont.). [0.5] l If par. pressure >/= 2350 psi [0.b] ] Dryout of the S/G's will occur earlier (less time available for l establishing secondary heat sink or RCS feed and bleed);
- OR -
Causes RCS feed and bleed to be less effectiv [0.5]
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REFERENCE .Tw ea u > &<r i n (w rp r, g c5 , i MP3 1987 Requal Objectives, , Item 2B- MP3 EOP 35 FR- ! MP3 EOP Development Training Text HO EOP 35 FR-8, pp.21-2 ) 000054K304 003000K608 ..(KA's)
]
ANSWER 7.05 (1.00) % per minut [0.50] In no case should the load reduction be halted due to AFD going out of the target ban [0.50] REFERENCE MP3 1987 Requal Objectives, , Item 4A- MP3 OP 3204, p. 1 G001 004000K405 004000K123 004000K122 004000K108
..(KA's)
ANSWER 7.06 (1.50) If the RCP has been idle for an extended period (30 days). If RCP maintenance has been performe . If the loop has been draine [3, @ 0.5 ea.]
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'* PROCEDURES:- NORMAL; ABNORMAL,. EMERGENCY Pago'2R AND RADIOLOGICAL CONTROL .
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. REFERENCE q MP3 1987 Requal Objectives, p. 17, Item C-2E MP3 OP 3301D, p. 1 !
003000K614- 004000K603 ..(KA's) l l ANSWER 7.07 (2.50)- (N-18) : 85 REM [0 30).
l Total lifetime to date = 81.0 + 1.75 = 82.75 Re Total lifetime.available = ;85 --82.75 = 2.25 Rem [0.30]. i Total this quarter available'= 3 - 1.75 = 1.251 Rem [0.30]. l
.I Quarterly is more' restrictive than lifetime limit [0.10]. 1 l
0.50 Rem /Hr gamma + ( 045. Rad /Hr)(10 QF) neutron.: 0.95 Rem /Hr. dose. rate [0.30]. l l 1.25 Rem /0.95 Rem /Hr = (1.32 Hrs) = 1 Hour,'19' Minutes.[0.20]~. l
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! Rem whole body one time exposureL[0.50]. l Director of Site Emergency Operations (DSEO).[0.50]. ]1 i REFERENCE j
.l MP3 1986 Requal Objectives, HP Procedures Objectives, #2 & #4, j 10 CFR 20 j MP3 SHP-4902 , pp. 6-7 & 14-1 '
194001K103 007000K100 007000KiO3 ..(KA's)- I ANSWER 7.08 (0.50) To assure that the amount of heatsink postulated in accident analysis is available to mitigate an accident (while an evaluation of steam gen-erator level instruments is conducted). ~or Becese h.s;y h ade y s an rt, e & G. re fere,a a e i 3 w, ,w ), in erNr3 on i s ll ere rect 1; &- / eve /, MP3 OP 3204 Change 4 documentatio K401 '005000K104 005000K102 ..(KA's) l
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q ANSWER 7.09 (1.00)- 1 1 Terminate.the startu . Drive rods i . ' Commence boratio [3 @ 0.333'ea.]- g
]
REFERENCE . I MP3 1987 Requal' Objectives, p . L 21' , Item C-3E '] MP3 OP 3202, precaution 4.1 l
*** Include MP3 Figure 7.1'from OP 3202 " ROD BANK INSERTION LIMIT vs. THERMAL POWER,.FOUR LOOP i
. OPERATION" with' examinee's packag *** j 001000G001 006000K601 ..(KA's)- i
i f ANSWER 7.10 (2.00)' RCS subcooling'(based'on core exit'.TC) - >30 . SG pressures - Stable.or decreasin . RCS hot' leg (WR) temperatures - Stableior decreasin a . Core exit TCs - Stable or. decreasing. . .
' RCS cold leg (WR) temperatures .At saturation for SG pressur [5,- @ 0.4 ea.] ,
t REFERENCE MP3 1987 Requal Objectives, , Item 9A- MP3 EOP 35 ES-0.1, Step K101 006020K404 ..(KA's)
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ANSWER 8.01 (1.50) Reviewing and approving all discharge [0.75] . I Terminating the discharge permit, if he does not agree with the , method of discharg [0.75] ! REFERENCE MP3 1986 Requal Objectives, ACP 6.03 Objectives, # MP3 ACP 6.03 Rev. 6, Para. ) 068000G001 026020K401 026000K402 ..(KA's) l i I ANSWER 8.02 (2.00) i l Least severe (1) to most severe (4). j
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Classification Level State Posture Code l
-------------------- ------------------ \ Unusual Event Delta-One/ Delta-Two 3 Alert . Charlie-One ' Site Area Emergency Charlie-Two General Emergency Bravo / Alpha
[8 components, @ 0.125 ea.]
l . Used to determine which off-site protective actions to implemen [0,50] Used to determine when to man emergency operating centers. [0.50] REFERENCE MP3 1986 Requal Objectives, Emergency Plan Objectives, #1 & # MP3 EPIP 4701, Sect. 3.1.2 & Table 4701- A116 002000K109 002000K106 ..(KA's)
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AND LIMITATIONS ANSWER 8.03 (2.00) Work more than 16 hours straight [0.50], excluding shift relief / turnover time [0.166]. Work more than 16 hours-in a 24 hour' period'[0.50], excluding l shift relief / turnover time [0.166]. ] Work without a break of 8 hours between work periods [0.50],- I including shift relief / turnover time [0.166]. ! REFERENCE I I MP3 1986 Requal Objectives, ACP 1.19 Objectives, # ) MP3 ACP 1.19, Sect. 6.2 & Fig. H 194001A103 008010K401 ..(KA's) ANSWER 8.04 (1.50) i Authorized individuals responsible for direct operation of the uni I Authorized personnel who may be required to support or advise the I operatio l Resident NRC inspector REFERENCE MP3 1986 Requal Objectives, ACP 6.01 Objectives, # l ACP 6.01 p. 1 K105 008000K102 ..(KA's) ANSWER 8.05 (1.50) Proper operation and indication of Train B SI must be verified (before taking Train A out of service).
REFERENCE MP3 1986 Requal Objectives, ACP Objectives, " Removing Equipment From Service."
MP3 OP 3250, Sect. 6.1 K101 ..(KA's)
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AND LIMITATIONS ANSWER 8.06 (1.00) Spent fuel pool' water level.[0.50] shall be maintained >/:23 feet' over the top of'the irradiated fuel assemblies [0.50] (whenever
. irradiated fuel assemblies are in the storage pool).
REFERENCE MP3 1987 Requal Objectives, , Item 100-5 MP3 TS 3.9.11 & B3.9.1 G003 061000SG4 061000K111- 061000K105 061000K101
..(KA's)
ANSWER 8.07 (2.00) This ensures'that: 1. MTC is withiniits analyced temperature rang ~ 2. Trip instrumentation is within its normal temperature rang . The F-12 interlock is above its setpoin . The pressurizer is capable of being operable, with a bubbl . The reactor vessel is-above its minimum. nil-ductility 1 temperatur [Any 4, @ 0.25 ea.] ' . Within 15 minutes prior to achieving reactor criticalit [0.50] .I At least once per 30 minutes when the reactor is critical l
[0.166] and Tavg <561 F [0.166], with the Tavg-Tref'
Deviation Alarm not reset [0.166]. REFERENCE l MP3.1987 Requal Objectives, p. 7, Item 30- I MP3 TS 3.1.1.4, 4.1.1.4, & B3.1. G005 078000K301 ..(KA's) I i I f
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
- _ _ _ _ _ _ _ ._ . _ . _ _ _ _ _
.,, .. ' . ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 341 ,
AND LIMITATION ) ANSWER 8.08 (2.00)' Suspend all' operations involving positive reactivity changes (cooldown). . Suspend movement of= irradiated fue . Suspend 1 crane operation with loads over the fuel storageLpoo . Initiate corrective action to' restore the. required power sources
'
to' operable status as soon as.possibl [4,- @ 0.5 ea.-]l NOTE: No credit given if candidate states,." Suspend all operations involving core alterations", because the conditions,of th question stated no core component movement'is in. progress in the containmen REFERENCE MP3 1987 Requal Objectives, , Item C-3A MP3 TS Sect 3.8. G005 076000K403- 076000K402 076000K401 ..(KA's) ANSWER 8.09' (1.50) Immediately open the isolation valve [0.5]. ! . Be in (at least) hot. standby within 6 hours-[0.5), and ~ Reduce pressurizer pressure to <1000'psig within the ' following 6 hours [0.5]. REFERENCE i MP3 1987 Requal Objectives, ,. Item C-3A2 MP3 TS Sect. 3. G005 001010K603 ..(KA's) l (***** END OF CATEGORY 8 *****) l (********** END OF EXAMINATION **********) _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ .
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QUESTION .VALUE- -REFERENCE' 'i
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5 .' 01 ~ 1.00 ZZZ000000 ' l 5.02 ~1.00 ZZZ0000002 1 5.03 1.00- ;ZZZ0000003 'l 5.04 1.00 -ZZZ0000004
.5.05 5.06 1.00'
1.00 ZZZ0000005 ZZZ0000006
]] l 5.07 1.00 ZZZ0000007 4 5.08 1.00 ZZZ0000008 5.09 1.00- ZZZ0000009 5.10 1.00 ZZZ0000010 a 5.11- 1.00 ZZZ0000011 1 5.12- 1.00 ZZZ0000012 3'
5.13 -1.00 ZZZ0000013 3 5.14 1'.00 ZZZ0000014 1 5.15 1.00 ZZZ0000015 j ______ 15.00 6.01 2.00 ,ZZZ0000016 6.02 1.00 ZZZ0000017 : 6.03 0.50 ZZZ0000018 '
6.04 1.00 ZZZ000001 i 6.05 1.50 ZZZ0000020- 1 6.06 2.00 ZZZ0000021 6.07 2.00 ZZZ000002 ! 6.08 2.00 ZZZ0000023 6.09 1.00 ZZZ0000024' 6.10 2.00 ZZZ0000025 ______ J 15.00 7.01 1.00 ZZZ0000026 7.02 1.50 ZZZ0000027 7.03 2.00 ZZZ0000028 7.04 2.00 ZZZ0000029 7.05 1.00 ZZZ0000030 ; 7.06 1.50 ZZZ0000031 I 7.07 2.50 ZZZ0000032 -i 7.08 0.50- ZZZ0000033 7.09 1.00 ZZZ0000034 7.10 2.00 ZZZ0000035 ______ , _7 15.00 '
:
8.01 1.50 ZZZ0000036 I 8.02 2.00 .ZZZ0000037 8.03 2.00 ZZZ0000038 l ' 8.04 1.50 ZZZ0000039 8.05 1.50 ZZZ0000040 8.06 1.00 ZZZ0000041 8.07 2~.00 ZZZ0000042 8.08 2.00 ZZZ0000043 8.09- 1.50 ZZZ000004 ______
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NORTHEAST UTELITIES
, ._~._._,w_ o,ne, i o,,,ce, . seioen si,,et e,mn. conn.c,,co,
,
*",$[' [U"'e'j,'*"" .P.o BOX 270 ( ' 'd wmm uws nue c HARTFORD, CONNECTICUT 061410270 wasw %cso mec ~
(203) 665-5000 September 9, 1987 7 MP-10831 Re: NUREG-1021/ES-201/ para .i U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555- ~
l Reference: Facility' Operating License No. NPF-49 Docket No.-50-423 September 4, 1987 NRC License Requalification Examination Comments Gentlemen: i Attached is the compilation of comments on the' written requalification examinations administered'to Millstone Unit No. 3 license holders on September 4, 198 These comments were the result of a review of the examinations training staf conducted by members of the Millstone Unit N Included are both the comments discussed during the exam review meeting of September 4, 1987.plus additional comments resulting from reviews conducted subsequent to this meeting. Attendees at the September 4, f 1987 meeting were: j Stotts - Northeast Utilities I R. Temps - NRC ) M. Moehlmann - Northeast Utilities l i The exam reviews were conducted considering the following l Does the question elicit the correct response? 1 Is the key answer correct? Is there potential for additional correct responses? Is the question appropriate? - References are provided, where necessary, to substantiate the comment x
n - p I, Kjjj U.S. Nuclear,Reguletory l Commiesion 0,-j e Re's NUREG-1021/ES-201/ para H.1 dj ' ' * ,t, a Page 2 of 2
!) ;
Please contact Mr. Ron?qtotts, Supervisor, Operator Training, //, , Millstone Unit'No. 3e uf;th any questions concerning our
< ,M, , comment ;
'
,3 Yours truly,
!
" NORTHEAST NUCLEAR ENERGY COMPANY l
'
, 30lutf E Ete-si ' '
, St'ephen E. Scace
! l ', i, Station Superintendent
> , Millstone Nuclear Power Station
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SES/RFM:jas '
,,' l' Attachment: Reactor Operator and l ,
Senior Recr.itor Operator Exam ; I f Comments and applicable references ec: Collins, BrarchChief, Region I
: R. : . Temps ,: Operatdr Licensing Branch, Region I ,
Ruth,'Itar.pger Operator Training
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REACTOR' OPERATOR EXAM
' PLANT DESIGN INCLUDING SAFETY AND EMERGENCN SYSTEMS -l H
l 2.07 Agreed to change the WasteLNeutralization Sump _ Monitor'-
. Condensate Polishing Facility automatic' action 1from q " Auxiliary Condensate is1 diverted to the aerated drains"'
to " Discharge.of sump contents is termin'ated" '
(Reference, OP.3336Di Rev 1, Pg 43'Section'8.9)'
2.07 Agreed to. include an additional. acceptable answer .i stating " Auxiliary Condensate Monitor:' Diverts.
, Auxiliary Building auxiliary condensate to the Auxiliary , i
' Building Sump" ( Re f e re n ce~, P& ID EM-135C Area M6) y 2.09 Agreed to change part 2. of this answer from " Unblocks the low pressure SI signal" to " Unblocks steamline low 3 pressure SI and blocks' main steam line' isolation on high steam pressure rate of change" (Reference, Westinghouse MP3 Functional Diagrams [ Logics] Sheets 16 & 7) 1 INSTRUMENTS AND CONTROLS l 3.05 Agreed to add "(CS)" as amplifying information to part l 1. of this answer i 3.05 Agreed to add "(Rod Control /C-16 selectable)" as amplifying information to part 3. of this answer .
.i
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3.05 Agreed to include an additional 1 acceptable answer stating " Steam dumps (Tref) 1 ad reject' controller" l
(Reference, NSSS Vol. 5, I&C Failures, pg 35 section
6.1) '
!
3.0 Agreed to change answer from "T c" to "T g" (Reference, NSSS Vol. 4, PZR' press and level, pg 14) I
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.. . SENIOR' REACTOR OPERATOR EXAM' )
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1 THEORY OF NUCLEAR POWER PLANT' OPERATION, FLUIDS AND THERMODYNAMICS' !
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1 5.03 . Agreed to grade this question on a case by. case basis? depending'on assumptions made by the examinee PLANT SYSTEMS DESIGN,. CONTROL AND INSTRUMENTATION 6.10 Agreed to change the Waste Neutralization Sump. Monitor-
- Condensate' Polishing Facility automatic. action from " Auxiliary. Condensate'is diverted to the aerated drains" to"" Discharge of sump contents is terminated" L (Reference, OP 33367D, Rev 1, pg 43 Section 8.9)
L 6.10 Agreed to include an additional acceptable answer stating " Auxiliary Condensate Monitor: Diverts Auxiliary Building auxiliary condensate to the Auxiliary Building Sump" (Reference, P& ID EM-135C Area M6) ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS 8.09.b. Agreed to take into consideration the. fact that the knowledge required to answer this question is not-required knowledge per the 1987.Requalification Objective Millstone Unit 3 Technical. Specification required memorization policies do.not include'6 hour action statement (Reference, 1987 Licensed Operator Requalification Objectives pg 18 -and Ron Stotts letter OT3-87-030, to K. L. Burton, dated February 25, 1987)
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.OP 3336D Page 43- 9
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Subsequent 'l l Notify th'e Chemistry Departmen i When the-Chemistry Department has completed sampling Wastet ) j Neutralization Sump 3CND-TK11,' remove it from 7 recirculation as -per Section 7.1 step 7. .i
:) 8.9 WASTE NEUT SUMP DIS RADIATION HI CDX 2-4 *
Initiating Device Setpoint j 3CND-RIYO7- .1 Sample 2X li r Backg'ound/Analysisu
]
;
-Action
'
Auto i Vaste Neutralization Sump to Circulating Water Discharge a Tunnel 3CND-A0V245. CLOSE Initial ' 1
' If Waste Neutralization Sump 3CND-TK10 (3CND-TK11) l contents are being pumped to the Circulating Water Discharge Tunnel:. OPEN Tank 10 (Tank 11) Waste Neutralization Sump Recirculation 3CND-A0V298A.(3CND-A0V2988) at CDX CLOSE Tank 10 (Tank 11) Waste Neutralization Sump .
Discharge 3CND-A0V244A (3CND-A0V244B) at CD i Note the final reading on Waste Neutralization Sump ' Discharge Flow Quantity Indicator 3CND-FQI246 (CDX)' Subsequent Notify the Chemistry Department When the Chemistry Department has completed sampling Waste Neutralization Sump 3CND-TK10 (3CND-TK11) if it was previously lined up for pumping the contents'to the !
:
Circulating Water Discharge Tunnel, either i Restore the lineup to the Circulating Water Tunnel if radiation levels are within specification a's pe : Section 7.1 step 7. Pump the contents to the Regenerant Evaporator Feed
. Tanks as per Section 7.1 step 7. .i
)
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Consider first channel 50 The output from PT 505 is used to j perform the following functions: j o Block automatic rod withdrawal when PT 505 senses that percent turbine load has dropped below 15 percent (if selected by PS 505Z switch on MB7) o Sent to.the reactor protection system to develop the P-13 signa If both PT 505 and PT 506 sense that percent turbine load is below 10 percent, a signal is sent to P-7 (along with the 3 of 4 PR < P-10) to block the "at power"' trips, o Generate a reference temperature signal, Tref, (if selected by PS 505Z switch on MB7) for usage by the automatic rod j control system, temperature error circuit (Tref-Tavg) l o Generate a reference temperature signal, Tref, for usage in l[ _ the steam dump syste o Be compared to auctioneered high nuclear power for use in the automatic rod control rate of change of power mismatch circuit (if selected by PS 505Z switch on MB7) 6. PT 505 Instrument Failure High consider first a high failure of PT 505. PT 505 provides an input to the Tref circuit for use in both the rod control and steam dump system The Tref circuit, however, is designed with I a high limit on its output of 587 F - regardless of the magnitude of PT 505 impulse pressure. With the reactor plant already at 100-percent power, the rod control system should be maintaining auctioneered high Tavg very close to 587 Thus, no error i-35- ,
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Nb36 \f o l . . N y N % . VB switches' associated with the PORVs and block valves, there are two " arm / block" switches,. train A and train B, on MB4. The switches are in the " block" position during normal operating conditions and are placed'in " arm" when the plant is' cooled dow The redundant train A and train B pressure protection. circuits
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are identical;-only the source of.their pressure.and temperature inputs differ.y Temperature signals from each of the four1 hot leg wide-range RTDs (TE 413A, 423A, 433A, and 443A) are compared' . The lowest temperature is auctioneered out and used for train A actuation of PCV 455A; the' lowest reading cold leg RTD (TE 413B,. 423B, 433B, 443B) passes through a similar auctioneering unit and is used for train B actuation of PCV 45 The two-auctioneered outputs enter their'own programming circuit The programmers purpose is to deve:.op a PORV pressure actuation 'setpoint as a function of RCS input temperatur Because of the nature of the brittle fracture phenomenon, the protection setpoint automatically lower.1 as wide-range temperature lowers. The l characteristics of tha pressure versus temperature program are , directly related to the analyzed vessel fracture limitation Actual reactor coolant system pressure, as measured by the wide-range transmitter PT 405 and wide-range transmitter PT-403-for train A and train B respectively, is compared to the programmed setpoin As actual plant prassure rises, indicating
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that an unsafe condition is being approached, the following cold overpressure system events occur:
o When pressure comes to within approximately 30 psi of the current protection setpoint, a control board alarm alerts the operator to find and correct the source of the proble o When pressure reaches the setpoint, the " armed" PORVs open and relieve to the pressurizer relief tan The cold overpressurization protection system is placed in service on a plant cooldown. The train A and train B " arm / block"
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switches are selected to " arm". During heatup, the system is-14-
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.u C-3A2A
List the ' requirements for accumulator operability and" state-whether,;one' inoperable accumulator- involves a one hour- a action statement' (Modes.1, 2, a.3).
l C-3A2E' 'l
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l Describe the. actions required f or one ' accumulator inoperable; ! due to the isolation valve being-closed-(Modesci,12,'83)=. - C-3A3
' State whether the accumulator ' isolation. valves have Tech' I Specs associated device with their thermal overload protection j C-3A4
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Describe the requirements for accumulator isolation valve A.C. circuit operability inside containment and state L whether noncompliance involves a one hour action' statement I (Modes 1, 2, 3 & 4).
! i d i C-3ASA i
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Describe the' minimum. shift crew composition requirements for , all modes of operation and include any
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exceptions; or provisions to these requirement . C-3A5B - Describe the. action' required for noncompliance with the minimum shift crew composition Tech Spe i 18 Page: 18 Eof 75
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February 25, 1987 ,. OT3-87-030 i.
i ! TO: Ken Burton { MP3 0 ra ng ppervisor aAW
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FROM: Ron Stotts ! MP3 Training Supervisor SUBJECT: ' Application of Tech' Specs Objectives Attached are the general objectives for Tech Specs knowledge as approved by TPCC in September '8 You and I discussed as early as 1984 the expectations of. station management in terms of required operator knowledge of Tech Spec I recommend that we revisit this important area, and that-Licensed Operators should achieve the following minimum level of Tech Specs understanding: 1) Know what items / components are covered by some LCO or another (e.g., if a name, a system, or component., the student must recognize that there is something in Tech Specs about it) 2) Be able to locate anything in Tech Specs in a minimal { amount of time (I suggest 5 minutes to look anything up, read it, and tell me what is required) 3) Operators should commit to memory any action required by Tech Specs that the time frame for initiating action precludes looking it up in the boo While this last item may appear to be associated with #2 (above), the issue is not readily disposed. The time. frame for demonstrating the ability to use and interpret the Specs assumes that this is the only task required, no other concurrent duties. For this reason, I believe we must arrive at guidance for Item #3 by one of the these two methods: analyze situations that could lead to entry of each LCO action and determine whether or not there is sufficient time to allow reference to the book prior to taking action, or OS70 REV 3 83 BN1
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Application of-Tech Spec; Objective i OT3-87-030 j
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February 25, 1987 Page 2 of 2 Assign a general cut-off time (as we have, traditionally done).within which we' assume.that i reference.to Tech Specs would not be convenient for any situation. To date, we.have established-this time to be less than one hou I believe.we should implement the. method of the' general cut-of f . time, and perhaps ' devote attention to the other more detailed approach at a later date.' -If this..isLacceptable,1I further propose the:following modification to'our existing a criteria:
1) Operators with RO licenses should not be authorized to (or responsible-for) take actions directed by Tech , Specs from memory without direction by the'SRO in the- ' control room (i.e., SCO or SS). For-those-required actions that are "immediate", the RO should'probabl j have a good. understanding of what the SRO will be directing to be done; hence, to ensure our graduates of initial training have this. solid base, they should be required to memorize action statements of 15 minutes or les ) Operators with Senior Reactor Operator..licensep should commit to memory) as we had previously defined, actions to be taken in less'than 1 hou ~ If you concur with this proposal, I will initiate the required changes to our existing Tech Specs learning objectives-and present them for our joint approval as soon as possibl RGS/ tap c: R. Martin M. Moehlmann 1 File 4.3. i i i L,_ _, _ . _
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, , EM C LOWR Tr_ . k L , .Aaadi.e.l ra n.c c L o p s 4 h/,, #,, Eu,,0,,,, /,,,, <. A n s tou_ g,ys The following changes were made as a result of final review of the examination Answer 2.07/6.10 Add the following--radiation monitors to.the indicated part.of'the answer': "(CND-RE07)" "(DAS-RE50)" "(LWS-RE70)" "(LWC-RE65)" "( SSR-RE08)'! .
l "(CNA-RE47)" Comment: Acceptable alternate to noun name for radiation monitor Answer 4.07/7.04 Added the following: , l
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Increases heat input into RCS."- Comment: An acceptable answer to the questio Answer 6.07 Rewrite a'nswer.as follows: ' Thermal shock would occur [0.5] as follows: To the spray piping [0.5] because the p'iping downstream of the spray valve l- would cool to containment ambient temperature [0.25] and then be subjected to hot (550-F) water when spray flow started [0.25]. ] QR To the spray nozzle [0.5] because the nozzle would rise to the temperature ) of the pressurizer [0.25] and then be exposed to cool'(containment ambient .] and then RCS, 550 F) water when spray flow started [0.25]. 1 I Comment: Answer was incomplete, j l
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.i Because design' inadequacies in the SG. reference legs'can' result'in l errors in indicated SG level." !
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