IR 05000423/1987016
| ML20235Z478 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/09/1987 |
| From: | Keller R, David Silk NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235Z473 | List: |
| References | |
| 50-423-87-16OL, NUDOCS 8710210166 | |
| Download: ML20235Z478 (56) | |
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:U S. NUCLEAR REGULATORY COMMISSION REGION I-
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' OPERATOR LICENSING. EXAMINATION REPORT
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' Examination Report No. 87-16 -(0L) -
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'. . Facility ; Docket' N Facility License No. NPF-49 Licensee: ' Northeast Utilities-f P. 0.~ Box'270-
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' Hartford;~CT 06141-0270 Facility: .Millston'e Unit 3-
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:ExaminationiDates: -August 18-20,:.1987 .
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Ch ef Examiner: D. M. Silk, Reactor. Engineer (Examiner) OR /0/1/D
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Approved-By: . W_
- 'R. M Ke 1er, Chief 77WR .Section, DRS ft/d-date
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l Summan: . Written',-and operating examinations were administered'to'six' I Senior. Reactor Operator candidates. One candidate failed the written'
: examination and two other candidates failed the operating examinatio m l
l 8710210166 071015 PDR ADOCK 0500042 V PDR j
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REPORT DETAILS i REPLACEMENT EXAM i EXAM RESULTS: SR0 Pass / Fail Written Examin: tion 5/1 Operating Examination 4/2 Overa'l 3/3 Chief Examiner at Site: D. Silk Other Examiners: R. Temps P. Isaksen (EG&G) 3 Parsonnel Present at Exit Interview NRC Personnel P. Temps, Reactor Engineer (Examiner) E. Raymond, Senior Resident Inspector Facil' ty Personnel
!.. Scace, Station Superintendent C. Clement, Superintendent J. Harris, Operations Supervisor B. Ruth, Manager, Operator Training R. Martin, Assistant Supervisor Operator Training j M. Moehlmann, Operator Training Summary of NRC Comments made at exit interview:
The NRC expressed appreciation for the licensee's assistance in expediting access to the plant and for providing facilities for the exclusive use of the examiners while at the site. The NRC reviewed the number and type of examinations conducted over the previous week and presented generic weaknesses observed during the operating examinations as well as some problems noted regarding simulator fidelity and limitations. Two problems regarding the simulator were discussed: 1) Discrepancies between the simulator malfunction book's description of system response for the - _ _ _ _ - _ _ - _ _ _
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feed regulating valve oscillation and. locked RCP rotor malfunctions and actual simulator response; 2) limitation as to the inability to fail only a single train of SI without inserting individual overrides for each , componen { l 5. Summary of generic weaknesses noted on the oral examinations: 1
- When -classifying events, some candidates were using either Unit 1 or !
Unit-2's EPIP With the liquid discharge monitor out of servica, some candidates were not fully aware of Technical Specifications requirements imposed to continue a liquid waste releas While discussing a failure of the refueling cavity seal, inconsistencies were noted among the candidates' responses in that all candidates were unsure of, and gave different interpretations to, step 5 of A0P 3572 regarding verification that a cavity drain line ! has not faile . Examination review: A review of the written examination was conducted immediately following the examination. A number of comments were resolved during the examination review. The NRC agreed to grade various questions based upon reasonable assumptions made by the candidates. Attachment 2 contains facility comments and documentation to support alternate responses and modifications to the answer ke Attachment 3 details NRC responses to the facility comment . Summary of generic weaknesses noted from the grading of written examinations: This information is being provided to document areas of weakness which should aid the licensee in upgrading replacement training programs. No reply is require Question N .06b The effect of a stuck rod, during a reactor trip, on calculated shutdown margi .10a&b The effect of the loss of natural circulation on core delta T and steam generator pressur .11b The effect that the Emergency Generator Load Sequencer has on plant equipment during multiple casualties.
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'7.08a Verifying that the turbine governor and throttle valves l are closed following a turbine trip to prevent turbine !
overspree .05b The technical specification in effect if the minimum crew composition cannot be me .08 Personnel who can issue and can .be issued security key . j Attachments: Written Examination and Answer Key (SRO) Facility Comments on Written Examinations made after Exam Review l NRC Resolution of Facility Comments
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l U. S. NUCLEAR REGULATORY COMMISSION 4 SENIOR REACTOR OPERATOR LICENSE EXAMINATION
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FACILI(Y: _MILLSTQNg_3_____________ 4 REACTOR TYPE: _PWR-Mgg4________________ l DATE ADMINISTERED _@ZZ99ZI@________________ EXAMINER: _SJLE2_p ________________
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CANDIDATE: __ k _ ____ _ ________ INSIBUGII9BS_I9_GGUDID6IE1 Una separate paper for the answer Write answers on one side onl ;
.Stcple question sheet on top of the answer sheet Points for each l quastion are indicated in parentheses after the question. The passing {
;grada requires at least 70% in each category and a final grado of at "
1 cast 80%. Examination papers will be picked up six (6) hours after the' examination start % OF 4TEGORY % OF CANDIDATE'S CATEGORY _YGLUE_ _IRIGL ___SGQBE___ _YGLUE__ ______________C6IE@gBY_____________
.2Ez99__ _2Et99 ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 29299__ _2E299 ____._______ ________ PLANT SYSTEMS DESIGN, CONTROL, j AND INSTRUMENTATION 1
1 2E199._ _2Ez99 ___________ ________ PROCEDURES - NORMAL, ABNORMAL, j EMERGENCY AND RADIOLOGICAL j CONTROL i l i 2Ez99__ _2Es99 ___________ ________ ADMINISTRATIVE PROCEDURES, ) CONDITIONS, AND LIMITATIONS l
29200 % Totals Final Grade -
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All work done on this examination is my ow I have neither given i nor received ai ___________________________________ Candidate's Signature I i l
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS uring the administration of this examination the following rules apply:
. . Cheating on the examination means an automatic denial of your application End could result i n 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 examination room to avoid even the appearance or possibility of cheatin .. Use black ink or dark pencil gely to facilitate legible reproduction . Print your n.ame in the blank provided on the cover sheet.of the examinatio . Fill i n the date on the cover sheet of the examination (i f necessary).
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.. .Use only the paper provided for answer ) . Print your name in the upper right-hand corner of the first page of ecch j nection of the answer shee . Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a gew page, write gely 90 egg gidg 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 thrgg lines between each answe ~. Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl t Use abbreviations only if they are commonly used in facility Litetgtut !
' 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 . Show all calculations, methods, or assumptions used to obtain an answer i to mathematical problems whether indicated in the question or no { Partial credit may be give 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 gunmingt 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 completing the examinatio This must be done after the examination has , been completed-l l l l l ,
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: l3.JWhen you complete-your' examination, you shall:
a.D . Assen,ble your examination an f ollows:
. (1) Exam questions on top.
' ' (2) Exam' aids - figures, tables, et ' y (3) . Answer 1pages including figures which are part of the answer, Turn in your copy of the examination.and all pages used to answer the examination questien c.- Turn in all' scrap paper and the balance of the paper that you did 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 in progress, ybur license may be denied or revoke ,
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2222IBE98Y_9ElNUCLEeB_E9 WEB _EL6NI 9EEBBII9N2_ELUIDD2iGNpi PAGE: :2 h .a ;4m:IBE809DYNedIGSE
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' ' : QUESTION : 5 ' 01-~ " ( 1 ~. 50 ) ;
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fAicentrifugal' charging' pump.is running with the flow; control valve, FCV p ,401214."in.mid positio Indicate how each parameter. will' change (Increase,
Dacrease',for Remaintthe Game) if FCV-121 is. fully opene t . . , p ;. Discharge flow b; . .Pumpidischargefpressure-upstream;of the discharge valve.
Tl ' , -: c .' Notor amps' L, dV 'Available.NPSH'to'pumpE "5 e. . ' Seal injectionfflow (Assume seal injectionsflowl control valv '
<. is'in manual).
QUESTION.:5.02 (2.00) JANSWER the fallowing.TRUElor FALS k fa.-Equilibrium xenon. concentration at 50% power is one halt of the-
. equilibrium xrnon concentration. at 100% powe , ,
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' b.f Xenon concentration initial'ly increases' af ter a reactor . trip. . c'. Fina1' samarium. concentration'after a trip is a function of the
' . previous power leve .
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' Fi nal xenon concentration after a trip is a function of.the previous power leve EQUESTION .5.03 (1.50) .How will;the following parameters change (Increase, Decrease, or No change) 'ifi cnefmain steam isolation valve closes with the plant at 25%-
Lloadi Assume all controls are'in' automatic and that no trip occurs.- .j l na. Affected loop steam generator level' (initial change only) -l b. Affected loop cold leg temperature _ j ' c'.- Unaffected loop steam. generator level (initial change only) _d.. Unaffected loop steam generator pressure n.~ Unaffected loop cold leg temperature-OUESTION 5.04 '(1.50)
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Euplain how the flow element venturi-type flow restrictor will act to
:li'mit AFW f low i f a break occurs downstream of the ventur .(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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>lGUESTION.1 5.05' ( 1'. 60 )~ <UThei reactor has the following characteristics,at 100% powers-Tave is 587 F
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. ;end;Tstm'is 543 F. Assume that'the plant is'chutdown and that some:. steam ' .gsnerator tubes are plugged ~soLthat'only 95% of the original. heat transfer Earealis'available. :If the-plant returns,to 100%; power.and Tave in again l597sF, determine the pressure of the steam ~1eaving the steam' generato State'all equations used'and assumptions.made and show all wor ~ QUESTION. 5.06 (2.00)-
L4 . . . . . La.-The plant is operating at 85% power with all systems in:automati' 'The ' operator; inadvertently aligns charging pump suction to the RWS Explain how'and why' shutdown margin'will be affected by this action?
- State all assumption . If=a' control. rod was stuck complet'ely out of the core on a reactor-trip, how'and.why would the CALCULATED SDM differ from'the case where all rods'are inserted'on'a trip? .-QUESTIO'N 5.07 (2.00) ~ HOW (Increase, Decrease or Remain the Game) and WHY would each of the following parameter changen affect DNBR7 /
a .' - Pressurizer Temperature Increases 5 degrees b.- Mass flow rate in the core Increases 10% c.- AFD increases to +10%
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QUESTION 5.08 (2.40) Compare and explain the difference in the reactivity worth of a rod that
, 'is i droppedLwhile.at power to the reactivity worth of the same rod stuck out while.all the other rods are inserte I a .
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.QUESTIONi,5.09- (2.80)
Listed' bel'ow are conditions-which-will affect a fuel. load 1/M plo For each1 condition listed, state if criticality will be UNDER or OVER predicted'and give'a brief' reason for your answe a.< A f uel : assembly in loaded ' near. the detecto 'r
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b.o.The' detector is too far-from'the newly'insta11ed' fue > , . . , The detector is too close to the sourc '
.- -d.-:The detector is too far from the sourc QUESTION 1 :5.10 (2.50) ,ah Describe HON-(Increase, Decrease, or Remain the'Same) and WHY core ' Delta--T wi11' respond to a loss of natural circulation' flow (caused' .
byc thepsteam dump valves f ailing ' closed) following'a reactor-trip
:from 100% powe (0. 75)'
b . -- Explain how Teold wide' range and steam generator pressure will-
: respond-to.a11oss of; natural circulation flow (caused by a loss of
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f eedLflow) - f ollowing a reactor trip f rom 100% power? (0.75) How would the operator' change the following parameters (Increase,
- Decrease, or.No Change) in order to ensure that natural circulation
. continue Consider each separatel (1.0)
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.1. Pressuri zer ' level at 5%-
2. RCS cooldown rate at 40 F/hr 3. Steam' generator level at 30% WR 4. RCSl pressure at 1900 psig ,
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l QUEST' ION 5.11 (2.20) a.- Of the coefficients that contribute to the power defect, which coefficient contributes most to the change in power defect over core life and what causes the change? I , Of the coefficients that contribute to power defect, which 1 I coefficient acts first to affect reactivity on a sudden power change and why does the coefficient react before the other i l I (***** . CATEGORY 05 CONTINUED ON NEXT PAGE *****) i
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-QUESTION 5.12 (3.00)
Consider the f ollowing plant conditions:
' MODE 3, BOL Boron concentration is 900 ppm All shutdown banks withdrawn Actual reactivity present in the core is minus 4% delta-K/K Source range indication of.100 CPS Differential boron worth is minus 10 pcm/ ppm A' boron dilution to 750 ppm increases the source range indication to L132 CP During the dilution, Xenon concentration has changed. How many PCM of reactivity did xenon contribute during the dilution? State all equations used and' assumptions made and show all wor I
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,- ; ; . QUESTION ' 6. 01; . (2. 00) - , :Whatiis'theifai1* position of the. f ollowing valves (Fail: closed, Fail'open, . Fail,_'as,is)(as caused'by a' loss'of contal ai ' ' R Feedwater .regul ating val ve'- (FCV-510) . '
i b. Seal : water ' flow controller ' (CHS-HCV-182)
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c. ; Chargingo flow control val ve.;(FCV-121)
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, Pressurizer spray valve (PCV-455B) RHR.HC. bypass flow control. valve (FCV-618)'
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TQUESTION- 6.02 . (2. 50)- m . . . i
.' '/The plant.is operating at 80%' power when'a'Thot RTD fails high'. -How,.if ct all . .- wi'11 this failure. affect:the following i.tems. : Consider each item 1 independent y.; Assume'no. operator action and all control systems .are in . automati ,
a. Rod insertion limit setpoint
::b. Charging flow- (initially)
i c. Control. rod-bank position Steam _' dump conteal system
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L e.-Overtemperature Delta-T trip setpoint OUESTION '6.03 -(2. 00)
:TheLplant is.at 100% when a small break LOCA occurs. What changen, i f :tany, occur in the main feedwater and the AFW systems in response to this : accident? Component numbers are not required. /}ss,@c W,k rS S 'en c, bl(
bepsfiht GONfy c b ibe ch6<g;n3 Sydem
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1,.. QUESTION,;6.~04 -(2.40)
. , i-l c . . Reactor Trip l Breaker RTA is' opened for RPS> testing.dneing Mode .1.- , .If1 Bypass. Breaker BYB is closed:priornto resetting RTA,owhat actions, if;any,,willfaccur? (1.0) ~b.c;A spurious reactor trip 1 occurs with-the plant at1100%~ power, however
' ' ReactorLTrip BreakerfRTB fails to open. What will be the effect, ifL an'y,,on eachlof'thalfollowing: ( 1. 4 )-
% t 1. ' Turbine Generator; * Feedwater isolation valves Steam. dumps!. . '4.- SI T reset ? (if' SI had occurred) ,
QUESTION l,6.05 (2.50) p,
...,, : C a . -. 'IF theilocalitemperature' controllers for the RPCCW heat exchangers are .
significant1y' increased, indicate how the-following'parametern will change (Increase,; Decrease, or. Remain the Same): (1.0) 1 '. . Thermal barrier flow 'VCT temperature: RPCCW letdown' heat exchanger flow
~ Gurge. tank level' IIn the event of a. thermal barrier leak, what three system design-features (both active and passive) will prevent the spread of reactor coolant throughout the RPCCW system? (0.75) What three system responses occur as a result of a low surge tank levol?- (0.75)
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-QUESTION. 6.06' (2.50)
Followingia large break LOCA, circle the valves on Figure 1 that will be: in a dif f erent position than shown f or cold leg recirculatio QUESTION 6.07 (1.60) s ; Assume a' normal plant cooldown is in effect with the RCS at 450 F and 650
: p ni g .- HOW and WHY will RCS pressure change if cold leg RTD. TE 413B, ' fails-low? (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) - _ _ - _ _ _ _ - _ _ - _ _-_
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D DbESTIONL'6.08' J(2.000 r jTh'e;plantcis':at 30% power.withJnormal system al'ignments and all; control p 2 systems are in automatic., . Explain the plant responses and f actors lead-D , ing:toLa reactor trip ifEfirst stage pressure transmitter, PT-505, fails
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- f l o w ( a t c B O L'.. ;iz . -QUESTION-.6,09 (2'00) . -a.- :DuringDa plant startup, .at less than 10EE-10 amps on the intermediate '
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' . range detectors', aEreactor< trip will occur du's to a loss of a single-120 VAC? instrument; bus. .. Assuming that all other equipment and instru-mentation is operable, what are two possible causes of the. trip? ' 'How, if at all , would ' the saf ety related 120 VAC system alignmerit s
change if'an, inverter experiences an overload? , 4
; OU'STION)E . 6.10 '. (2.50)
la. - For: EACH of' the .f ollowing two cases, state YES or NO as to whether the reactori.will-trip =in response to a simultaneous failure low of both
~ - -Intermediate Range channels. Justify each answe ) Reactor startup in pregress and at 5%ipowe .(0.75) d 2) The reactor'is at 20%' steady state power (0.7F) j l Assume the reactor is at 100% power when one Intermediate Range channel 'I ' f ails high,; f ollowed immediately by a reactor trip . (f rom other ut.uses) .
Wh'at 7 actiot1, if any , will have to be taken during emergency procedureu to ensure proper operation of the Nuclear Instrumentation System? (1.0)
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M OOESTION6i11 '
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Answer?thelfallowing questiens regarding'.the; Emergency Generator Load
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i, '.} Szquencer 3 (EGLS) : > 1
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;s.; .Following/a . loss of offsite' power (LOP) condition,Jgive the,tw conditions;when.the Safeguard Sequencer Start (SSS) .and; Manual
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Trip- " ! Block-(MTB)-are: reset?- If:.a3 LOP accident;is.in progress with the EGLS at time'30 seconds in La' LOP. sequence.and a subsequent CDA signal occurs, how'will the EGLS . . affect.the sequencing of loads required.by.the CDA signal?
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' accident condition during a test. sequence for Test i and.. Test 27' ,y
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OUESTION 7.01 (1.00) l Mntch the class of . fires listed below (A - D) with the materials involved (1 - 4).
CLASS OF. FIRE MATERIALS INVOLVED i ______ Alpha Flammable liquids, gases, or greases ______ ' Bravo Combustible metals ______ Charlie Ordinary combustibles (paper, wood, etc.)
I ______ Delta Energized electrical equi pment' DUESTION 7.02 (2.00) In or der to maintain the plant at 100% power, work must be performed inside the containment in a radiation field of 400 mrem /hr gamma and
~2.0 Rem /hr fast neutro The maintenance man selected is 24 years old and han a lifetime exposure through last quarter of 28 Rem on his NRC Form How long may the man be permitted to work in this area per ;
10CFR2O limits? ! QUESTION 7.03 (1.60) !
! An NRC inspector f rom Washington wants to go into containment while the plant is at 50% powe :" arder for this entry to be made who, if anyone, must accompany tem inspector? (0.6) Under-what condition, if any, can the 10 psi /hr repressurization rate ;
limit for containment atmosphere be exceeded? (0.5) j
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!- According to Technical Specifications, what action should be taken ,
after containment vacuum is broken while at power? ( ) ! t
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QUESTION ~17.04L (2.00) re.!VWhile/ supervising?a tefueling; operation,.the SRO-in containment-
' .ob'serveskthatia new;fue1Lassembly;was inserted-into'the core and 'leftEfree. standin Iszthis permissible? ( 0. 4') ~
l n 16.:.Whyfshould,a spent l fuel assembly, in'the'upender on the containment l side, be_inithe hacizontal' position priorfto' latching another. spent ~ fuel, assembly? (0.6) Jc . :During' refueling, under what condition, if any, can residual ~ heat removal' flow be. suspended? - ( 1. 0 ) -
::n QUESTION ~ 7.05 (3.00) ' . Answer theLfollowing questions regarding OP13201, Plant Heatups a. - The1 shutdown banks must be fully withdrawn whenever positive. react-- ~ " -
ivity 'is being added .by xenon concentration change Under what two conditions,Liffany, is the above statement ~not applicable? : ( 1. 0) 6.;JWhile'in'Mdde 4, at least one RCP shall be in operation. However, the RCPsimay' be .de-energized f or up to one hour provided two condit-ions are me What?are these two conditions? (1.0) During a startup after RCPs 1, 2,.and 3 have been started, it is de-cided'to.stop'RCP 2 due to excessive vibratio Why is.the loop 2
'L-ripray valve-(PCV-455C) closed? (0.5).
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. During a lstartup, 'ediy are the atmospheric steam dump controllers in MANUAL' prior 1to making'any setpoint changes and then placed in AUTO after thejchanges are completed? (0.5)
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Zu__ESQQEQUBgg_ _UORMAL, ABNORMAL _EMEB@gNQY_AND t PAGE- 12 4,L bed 1969EIGGL_CQUIB96 i QUESTION ~ '7.06 -(3.00) q
. - l .Antwer the following questions regarding DP 3202, Reactor Startup: ;
i ' Prior.to commencing a reactor startup, what is the minimum nuclear instrumentation required to be operable? (0.75) WhatLthree individuals (by title) can grant permission to take the reactor critical? (0.75) ;
! What action, if any, should.be taken if criticality occurs below the !
rod insertion limit? (1.0) l What action, if any, should be taken if an inadvertent steam dump actuation occurs just after criticality is achieved and Tave decreases to 549 F7 (0.5)
.QUECTION 7.07 (2.80)
Answer the following questions regarding AOP 3554, RCP Trip or Seal Failures How would RCP seal water return flow and RCP lower bearing water temperature change as a result of a f ailure of a RCP's number one seal? (1.0) i 'What actions should be taken regarding RCS leakage, reactor protect- ! ion input, and reactor power?f % e c <g,/ Mfm he eu vera/ (1.8) )
QUESTION 7.00 (2.00) . l Answer the following questions regarding AOP 3550, Turbine / Generator Trips When tripping the turbine and generator, what three actions should be taken (in proper order) to prevent turbine overspeed? (1.4) i If the turbine does not trip following a reactor trip and will not manually trip, what action can be taken to trip the turbine? (0.6)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE ****+)
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l l QUESTION 7.09' (3.10) During an' emergency condition'the.STA reports the following: 1 -. Core Cooling - Orange Path Subtriticality - Yellow Path Integrity - Orange Path Heat Sink - Red Path 'In what order.should the above conditions be addressed? (1,0) Ten minutes later the STA reports that the above conditions still exist except that Core Cooling is'en a Red Path. What. actions should be taken and why? (1.0) With conditions 1 through 4 present, what optimal or functional'
. recovery procedure should the operator follow if a loss of offsite power occurs and both diesels fail to start? (0,5) After a Reactor Trip, what two circumstances initiate monitoring the Status' Trees? (0.6) 'OUESTION 7.10 (2.50)
Answer the following questions regarding E-3, Steam Generator Tube l
'
Rupture (SGTR): List the four ways of identifying a ruptured steam generator (1.0) Why would it be necessary to bypass the low-low Tavg interlock during a steam generator tube rupture? (0,5) Why would voiding in the upper head region not prevent SI termination? (1.0)
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PAGE 14 q LZs_$EBQGEQQSE@__UQBd@bz_G@NQBd6Lx_EdE8@E6GY_66Q' ji -BGQ1960Q1966_GQUIBQL-QUESTION. 7.11 ~(2.00) q p Anuwer the f ollowing questions regarding FR-H.1, Response to Loss of-Sucandary Heat Sink: j o What are the Red Path entry conditions'for FR-H,17 (1.0) When, if-at all, is it permissible to feed a faulted steam generator?
(0. 5) . Fol'1owing' bleed and feed operation, why.would the operator transition out of FR-H.1 if RWST level decreases to less than 520,000 gallons? .
(0.5)
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FQUESTION' 8.01 '(2.00) ,
, s.n ,
9Temporaryichanges to: procedures may be'made if'what three' provision yerenmet?
: QUESTION' O.02 ' (2. 00) -
i;The-concentration of^the boric acid solution i n the Refueling Water
; Storage Tankj(RWST) shall be verified once per 7 days in accordance, LwithfTechnical Specification 3. The chemist sampled the RWST j
<' on1the following schedule. '(A11 samples taken.atji200' hours.,).
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. April 1 ---' April-l8.--- April 16.--- April 24'-- ' April 31 s
j La. . EXPLAIN why the surveillance time interval' requirements'WERE o'r
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fWERE NOTLexceeded,on April.1 bk EXPLAIN why'the surveill'ance time interval. requirements WERE.or-
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WERE NOT exceeded;on April 2 ~
1 QUESTION 'B.03 (2.00)
'In.accordance with the Station Tagging procedure ACP-DA-2.06A, what iTHREE conditions muss be met in order to lift safety tags.from equip-ment, thus allowing temporary operation, such as a-retest or a , afunctional check? QUESTION. 8.04 (2.00)
! .The plant is operating at 75% power and.the latest leak rate data shows: 10.5 GPM - Total RCS leakage rate
~1.5 GPM - Leakage into the Pressurizer Relief Tank 1".2 GPM - Leakage into the Primary Drains Transfer Tank 1.5 GPM - Leakage through 3-RHS-MVO702C
. I" O.8 GPM - Total primary to secondary leakage N w ddrd M ca'.d R') [ 4.2 GPM - Leakage past RCP seals
.What RCS. leakage limits, if any, have been exceeded? Justify your answer by ref erencing the appropriate sections f rom the Technical Specifications l - and:by showing any calculation ,
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.o tt I QUESTION 8.05 (2.10) l l
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The unit is presently at 30% power and is increasing. Due to a combinat-ion of vacations, sick leave, and inclement weather, the oncoming shift will be one member short of the Technical Specifications minimum crew l composition (a reactor operator will be missing). l What should happen at shift turnover and what other administrative requirement should be checked in regards to meeting the minimum crew composition? (1.3) If the minimum crew composition can not be met, what Technical Spec- 1 ification action statement is in effect? (0.8) , i l l OOESTION 8.06 (2.20) j i The following concern " Station Bypass / Jumper Control" procedure, ACP-QA-2.06 If a Technical Specification change is required, or'an unreviewed j
'
safety question is found to exist, then approval is required from ____ (1) ____ AND ____(2) ____ before bypass / jumper installation can i
!
be authorize What condition allows the Shift Supervisor to grant an exception to performing a second verification of a jumper installation?
! Which TWO individuals, by job position / title, must complete and ;
sign the Assessment Section of the jumper-lifted lead-bypass control sheet? (Do not include the PORC review / approval signature.) Under what conditions can a jumper be installed on operable ) equipment WITHOUT using procedure ACP-QA-2.06B7 ! l QUESTION B . 0 7- (2.20) i
'
The plant is in Mode 3 with a routine startup in progress. During sur-veillance testing on one of the Diesel Generators it is determined that its day tank level switch is out of service and will not control its fuel oil transfer pumps. The Maintenance Supervisor assures the Shift Super-visor that repairs can be completed within 48 hour Explain why the startup SHOULD or SHOULD NOT procee ,
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0QUESTIONU8.08? .
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Arbdr the f f ollbwing questions regarding' ACP 7.04, C ont.r ol of Lock and
'
4 fKays Used .to Prcivide' Nuclear Related Security:: ,,
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< 1 4 - . . .. . . :-
, 4. ; lWhat:sixjplant'.personne1Eare authorized to.be-issued security keys ".
al
, . . .
. J'T . so . plant saf ety can be ' maintained ;during ' security . system if ailure?
* W/.(2.0)' [
jfh;; , Who istresponsible for authorizing the; issuance'of; security keys N tg9- .
;b(N'during.'an emergency condition? .
_ (0,5) '
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. (2. 50) . 1
{.OUESTION'8.09-
> .a[ Under:what circumstance is:it permissible to violate a-Technical ~
specification Limiting'. Condition 1of Operation.(LCO)? (2.0) b.ijf'Wh'o'should be notified'.if'an'LCO.isiintentionally violated?
' (0.5)-
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LOUESTION.~8.10: -( 2. 50 L .
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a; In .accordance w'i th . EPIP 4010A . '.'Shi f t Supervi ser", when must t the SS assume,;<the responsibilities of the Acting Director of Station
~ Emergency?Operationc? ,<
(0,5)
..bn -In accordance with EPIP 4010B "Ac ng('k)i rector of Station 'i Emergency Operations" 'what f our co'r/dli bMns must you consider when
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determining whether or not -to evac;uateT ~
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'OUESTION 8.11 *
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<, What' actions, BOTH operations?/'AND administrative,.are required b) the i
: Technical Specification f Reactor Coolant ~'Syphem pressure reaches 2775 ~
,
Lpsig1while in Mode 27 I clude. appropriate time limit ! ', ,j ,
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P(g 7 NEthCTOR COOLANT SYSTEM I hp~ , w ,, JAN 31 1986 !
~' 4 i ,c OPERAT10NAL' LEAKAG l LIMITING CONDITION FOR OPERATION i
3.4,' Reactor Coolant System leakage shall be limited to: I No PRESSURE BOUNDARY LEAKAGE, ' gpm UNIDENTIFIED LEAKAGE, I gpm total reactor-to-secondary leakage through all steam
, generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated 4 from the Reactor Coolant System, l gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, v ,40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2250 1 20 psia, and .5 gpa leakage per nominal inch of valve size up to a maximum of 5 gpa at a Reactor Coolant System pressure of 2250 1 20 psia from any Reactor Coolant System Pressure Isolation Valve specified in .,
g Table 3.4-1.
- ,t .K/ PLACABILITY: M00ES 1, 2, 3, and f ACTION:
'- ' a, With any PRESSURE BOUNDARY LEAKAGE, be in at least NOT STANDBY !
C, within 6 hours and in COLD SHUTDOWN "ithin the following 30 hour j
' With any Reactor Coolant System leakage greater than any one of the 'above ?imits, excluding PRESSURE 80LNSARY LEAKAGE and leakage from i
e Neactor Coolant System Pressure Isolation Valves, reduce the leakage
' '4 4 i Yete to within limits within 4 hours or be in at least HDT STANDBY # ,
l< rithin the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
g( V A c. \ With ;any Reactor Coolant System Pressure Isolation Valve leakaoe
:p
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greater than the above limit, isolate the high pressure portion of ; the affected system from the low pressure portion within 4 hours by j I ' use of at least two closed manual or deactivated automatic valves, F or be in at least HOT STAND 8Y within the next 6 hours and in COLD
+ SHUTDOWN within the following 30 hour .t ' .} / ' *C/ '
- $ i ( 4-MILLSTONE - UNIT 3 3/4 4-22 r..,
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REACTOR COOLANT SYSTEM . gr 3 j ggg y ' OPERA 110NAL LLAKAGt
' ! SURVEILLANCE REQUIREMENTS _
i 4.4.6.2.1' Reactor Coolant System leakages shall be: demonstrated to be within
# -each of.the.above'11mits by: Monitoring the containment atmosphere'(gaseous or. particulate)
radioactivity monitor at.least once per 12 hours;
' ._ Monitoring the containment drain sump inventory and discharge at least once per 12 hours; Measurement of the CONTROLLED LEAKAGE to the reactor coolant pum seals when the Reactor Coolant System.pressore is 2250 1 20 psia at-least once per 31 days with the modulating valve fully open. The provisions of Specificai. ion 4.0.4 are not applicable for entry into
. MODE 3 or 4; Performance of a Reactor Coolant System water inventory balance at ,
least once per 72 hours; and Monitoring the Reactor Head Flange Leakoff System at least once'per 24 hour .4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within
.
its limit:
! At'least once per 18 months, Prior to entering MDDE 2 whenever the plant has been in COLD SHUTDOWN'for 72 hours or more and if leakage testing has not been performed in the previous 9 months, Prior to returning the valve to service following maintenance, repair or replacement work on the valve, Within 24 hours following valve actuation due to automatic or manual action or flow through the valve, and As outlined in the ASME Code, Section XI, paragraph IW-3427(b).
The provisions of Specification 4,0.4 are not applicable for entry into MODE 3 l l or ' I
.
9-MILLSTONE - UNIT 3 3/4 4-23 i
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l/IN .i I ; d TABLE 3.4-1 REACTOR COOLANT SYSTEM PRES 5URE ISOLATION VALVES VALVE NUMBER FUNCTION 3-SIL-V15 SI Tank 1A Discharge Isolation Valve 3-SIL-V17 SI Tank 18 Discharge Isolation Valve ,. - 3-SIL-V19 SI Tank 1C Discharge Isolation Valve 3-SIL-V21 SI Tank 10 Discharge Isolation Valve 3-SIL-V26 RHR/SI.to RCS Loop 2, Hot Leg 3-SIL-V27 SIN to RCS Loop 2. Hot Leg 3-SIL-V28 RHR/SI to RCS Loop 4, Not Leg 3-SIL-V29 SIH to RCS Loop 4 Hot Leg 3-SIL-V984 RW!/SI to RCS toop 4, Cold Leg 3-SIL-V985 RHR/SI to RCS Loop 3, Cold Leg 3-SIL-V986 RHR/SI to RCS Loop 2, Cold Leg 3-SIL-V987 RHR/SI to RCS Loop 1 Cold Leg 3-SIH-V5 SIH to RCS Cold Legs 3-SIH-V110 SIH to RCS Loop 1, Hot Leg ' 3-SIH-V112 SIH to RCS Loop 3, Hot Leg j 3-RCS-V26 SIH to RCS Loop 1, Hot Leg 3-RCS-V29 SIH to RCS Loop 1, Cold Leg 3-RCS-V30 SIL to RCS Loop 1, Cold Lag 1 3-RCS-V69 RHR/SI to RCS Loop 2, Not Leg 3-RCS-V70 SIH to RCS Loop 2, Cold Leg 3-RCS-V71 SIL to RCS toop 2, Cold Leg _ 3-RCS-V102 S!H to RCS Loop _3, Hot Leg 3-RCS-V106 SIH to RCS Loop 3, Cold Leg , 3-RCS-V107 SIL to RCS Loop 3, Cold Leg i 3-RCS-V142 RHR/SI to RCS Loop 4, Hot Leg 3-RCS-V145 S!H to RCS Loop 4, Cold Leg 3-RCS-V146 SIL to RCS Leap 4, Cold -Leg - 3-RHS-fW8701C RCS Leap 1, Not Leg-to AMR 3-RHS-MWS702C #CS Loop 4, Not Leg to AHR 3-RHS-9W8701A RCS Loop 1, Not Leg to RHR 3-RHS-9W87028 RCS Loop 4, Hot Leg to RHR
l i jY MILLSTONE - UNIT 3 3/4 4-24 _ _ - _ _ _ _ - _ _ _ f % ca va 5/t- Cycle-efficiency o (Netdex-q > out)/(Energy la)
w o og s o V,t + 1/2 at E o ac 2
KE = 1/2 av , , (y( , y o)ft 4 , xy 4 , Ao,-a t PE = 898 V = V, +-at w = e/t A= f an2/t1/2 = 0.693/t1/2 1/2'If * EII1/7)II))
-
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[(t1/2) * (*b)3 4 = 931 m
*
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' Q . = mCp at Q = UAat I = I e~"* !
n Pwr = Wfah I = 1, 10~*/U L TVL = 1.3/u ' sur(t) HVL = -0.693/u P = P*10 p , p ,t/ o
$UR = 26.06/T SCR = S/(1 - K,g)
CR x = S/(1 - K,ffx) SUR e 26o/t* + (a - o)T CR)(1 - K,ff)) G 2II ~ keff2)
-
T = ( t*/s ) + ((a - o ) /lo ] M = 1/(1 - K,g) = CR /CR g , T = 1/(e - s) M = (1 - K,go)/(1 - K,ff)) T = (8 - e)/(lo) SDM = (1 - K,g)/K,g a = (K,g-l)/K,ff = Meff/K,ff t' = 10-5 seconds l
- T = 0.1 seconds-I l
o = [(&*/(T K,g)] + [f,ff (1 / + 17)] =Id ld; 2 ,2 2 { P = (t+V)/(3 x 1010) IId gd jj 22 I = oN -
R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (feet) Water Parameters Miscellaneous Conversions j l 1 gal. = 8.345 lb curie = 3.7 x 1010dps l 1 ga:. = 3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 ga I hp = 2.54 x 10 Btu /hr Density = 62.4 lbm/ft3 1 av = 3.41 x 1 Stu/hr
)
Density = 1 gm/cm3 lin = 2.54 cm Heat of vaporization = 970 Stu/lom *F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. H BTU = 778 ft-lbf
'l ft. H 07 = 0.4335 lbf/i p' "&y,- n . . :. .. . . .
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.iDi__INE98Y'OF NUCLEAR'PQWER_ELANT OPERATION 2 _FLgJpgi_ANp PAGE. 118 M~* =INEBdQD%N6dIGSc lJ , . .
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4 ANSWERS -- MILLSTONE ' 3
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o JANSWER 5.011 ( 1.50) if ~ fX ' a." [If1 crease
< ' b' . : Decrease '
Lch = Increase- Decrease- Decreasef ;CO.3' pts each].
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:RE'FERENCEr q .Rx Thc.'HTFF,.2 of?2,: Plant,. processes, pgsL6-8, 16-19-Components :191004.' Pumps': 1. 05 ' 2. 3/2. 4!
1.06- 3.2/ ~
* '
-1.07 2.9/ ,
EO '-- LHTFF&T 25b ,30
. ;
V191004K1071 191004K106 191004K105 ...(KA'S)
,
JANSWER- 5.0 (2.00)~
' a.f Fid se
.b.'True c.:Truei d. False- CO.5 pts each] !
<
REFERENCE ')' l R>b JTh HTFF , .1 of 2, Xe & Sm, pgs 12-30
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3.l'.001 '000 'K'5.33 3.2/ E3.1'001.000 K 5.34 2.1/ .1L001 000. K-5.35 2.'1/ q EOJ- Rx Th 59b,93'
;. OO1000K535 ; -
OO1000K534 OO1000K533 ...(KA'S) v
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' ANSWERS;-- MILLSTONE 2L -87/08/15-SILK, ' ANSWER- 5.03 (1.50)
c. - . Decrease . Increase Increase Decrease
'e. . Decrease: E0.3 pts each3 REFERENCE Rx Th HTFF, 2 of 2, RCS heat removal, pgs 34-42 1 of 2, Heat transfer processes, pgs 19-22 j.3.2 002 000.K 5.01 3.1/ .09 3."/ .11- 4.0/ EO -- HTFF&T 72 ,
l OO2OOOK511 OO2OOOK509 OO2OOOK501 ...(KA's) ANSWER 5.04 -(1.50) Fluid velocity through the venturi increases E0,3] which causes pressure to decrease to the saturation pressure and cavitation occurs.EO.6] which lower the dp across the venturi EO.33 and thus reduces flow since flow
'is proportional to the dp EO.3 REFERENCE Rx Th HTFF, 2 of 2, Plant processes, pgs 11-19 '193006 Fluid statics and dynamics K 1.11 - 3.1/ ,
K 1.15 3.1 '3. 2 3.5 061 000 1 5.05 2.7/ EO - HTFF&T 44 4 193OO6K115 061000K505 193OO6K111 ...(KA'S) I I
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Dz_ THEORY _gE_UUGLg68_EgWEB_EL6NT_gEgRATJgN2_ELUJpg3_6ND PAGE 20
* . IBEBd9DXNedlGS ANSWERS -- MILLSTONE 3 -87/08/18-SILK, D.
l l I ANSWER 5.05 (1.60) Q = U1 A1 (Tave - Tstml) = U2 A2 (Tave - Tstm2) CO.53 U1 = U2 A2 = .95 A1 (Tave - Tutml) = .95 (Tave - Tstm2)
-(587 - 543) = .95 (587 - Tutm2)
587 - 44/.95 = Tstm = 540.7 F CO.53 From steam tables Pstm = 967 psia CO.63 REFERENCE Rx Th HTFF, 1 of 2, Heat transfer processes, pgs 19-22 ( 3.5 041 020 K 5.02 EO - HTFF&T 54,61,62 041020K502 ...(KA'S) ANSWER 5.06 (2.00) I' . w - - L i t r . It ci^c '": .c-2,m : . '; . Thu Ju mi wuwe lii, ;-t;;;L, by Liie-trar vii will Lu equal Lu Ll u i r, m i ame ... . wouLivity f i vni i;TC C.c: eft-T uw L - L r t r+ I t increases E0.43 since RCS boron concentration is increased E-GrM (n. 63 (.9) m/s t r e f u rfW o * + c f the WC) M Its decreases CO.43 because the negative, rod worth was not inserted in to the core CO.6 REFERENCE Rx Th HTFF, 1 of 2, Reactor operations, pgs 33-35 OP 3209B 3.1 001 010 K 5.35 3.3/ K 5.19 3.5/ EO - Rx Th 96b OO1010K535 OO4000K519 ...(KA'S) _ _ _
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ANSWERS'- TMILLSTONE: '
-87/08/18-SILK,.D;. c '
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' ANSWER J5.07! (2.00) : <
m y J a.' IncreasesfCO.23. .AsLPZR. temperature rises, so.does. pressure.,'henc I tthe marginzto' saturation. increases'CO.463.
'i sbb LIncreases2[0.23. >The Delta'T ac'ross the core will be' lower'.tofpro- , duce lthe same power.-so:Th willl decrease and the.m'arginito saturation "
'i ncreases' C O. '46 3. .
ll'? Ici .DecreasesLCOL23. . Since!more power is being produced . in1the top half'
'of tthel core (and :since head . loss increases with core height) the. margi to! saturation decreases:: EO. 463.
l2 '
, REFERENCE
'
; RxETh' HTFF, 2'of 2,. Boiling, processes,'pgs 24-25 3.'4 003'000 K 5.01' 3.3/3.9-EO'-oHTFFLTl50
.OO3OOOK501' ...(KA'S)
'
! . ANSWER' 5 ' O'S t (2.40) L
?The-stuck rod would be-worth more CO.6 Reactivity worth is proportional:toothe relative flux squared EO.6 For a dropped rod, i' the-flux' is' depressed. adjacent-to it CO.63 whereas if the-same rod was:stuckLout,.while the-others were inserted, it would-be exposed to a. much' higher -flux' than the flux .in the ' rest of ' the core EO.6 REFERENCE RH Th'HTFF,-1'of'2, Neutron poisons, pgs 6-10 ~ 3.11OOO;OO3:EKL1.03 '3.5/ LOO 5 EK 1.05 3.3/ . ; . ' EO -:76a,76b,78,81 :OOOOO5K105 OOOOO3K103 ...(KA'S)
l l
\
!
..
L l- .. Asum-a---- a-a----L.-=--- ----
y'. ..
- $J ,
< DDi__IBE9By: OF NygLg68_EQWEB_EL6NI_gggBBI]QNi_E(yJp33_6Np PAGE . ' 2: t=F sit!EBUQ9XUBU1GB m [ ANSWERS.- YMILL' STONE'3 - 8 7 / O S / 1 8 - S I L K ',J D .
'
m'. t LANSWER '5.09' (2.80) ca.1(Loading (a fuelcassembly into the core.close to a neutron: detector-
' increases the fraction of neutrons in the core reaching the, detector).
N ;The1 detector. count' rate.. increases more than the core neutron population
, " increases ~ [0. 4 3. L Cri ti cal i ty' i s ' under.. 'predi cted1 C O. 3 ;m -
b. The' detector.will'not,see neutrons.until.:there are a great. numbe CO.4F Criticality.is overLpredicted. [0.33 c.'The initial count rate is too high.and the detector is insensitive to1 core changes. CO.43_ Criticality is.over predicted. [0.33 Id. The det'ector will notosee neutrons until there are a greater number CO.43'. Criticality'is over predicted. [0.33
,
REFERENCE-Rx Th HTFF, l'of.2, ~ Reactor operations, 19-23 3.1"001 010 K 5.16 2.9/3.5-
.EO.- Rx Th 97,98 '
OO1010K516 ...(KA'S) ANSWERL 5.10 (2.50)
~ Delta T increases CO.253 as That increases (as boiling occurs in core) and Tcold remains relatively constant CO.53 Psteam will'dec'rease as boiloff occurs in the SGs CO.43 while Tcold remains relatively constant due to stagnant RCS conditions EO.35 .. Increase "- - - - - Mc < kny , Increase <4 . No Change [0.25 pts each] , ' t e Th re n t REFERENCE RxLTh HTFF, 2 of 2, RCS heat removal, pgs 25-28 l .3.4 000 015 EK 1.01 4.4/ :ED:- HTFF&T 64,66,67 OOOO15K101 ...(KA'S) e
_ - - _ - _ _ _ _ _ _ _ - _ - - _ _ _ _ - _
_ ___ - _ _ ;
'
3 .5 .d , ~ J "llf'i,_"INEQBY_QE_NUGLE8B_EQWEB_EEGNI_QEEBGI1ONx_ELu1QSi_6NQ ' PAGE ~23
'
6,.1 x .2 w,IWEBdQDYNed1GE
,, _ . . . . .., m .. .. m .
, i. ANSWERS --- MILLSTONET 3 ,..
-87/08/lcl-SILK,.D.-
l _;.
,
n- <
,, ,
j:.]J
* ,' '
, EMNSWERi '5.l1! (2.20) o ? . . . . .,
.fah ModeratorDTemperature Coefficient (MTC) [0.43.due to an. increase- .y .(more; negative) in:MTC as bcron concentration.is reduced over core life'[0.7 .
bhLDoppler.(FTC) . C O. 4 Fuel' temperature changes before the,other J .+ para' meters change'CO.7 ;-: ' >>
'= REFERENCE X Rx;-Th HTFF, 1 of-24. Reactivity coefficients and defects,.pgs.6-24- :321 LOO 1 OUO:k;5.49 :3.4/3.7.- , LEO.- RxWTh'59a,59c,82c,82d,82 JOO1000K549 ... (KA'S)
iANSWER 5.12 (3.00) Keff1:=-1/((1-(-0.04).) = O.9615 CO,33 Keff2 =-0.9709 C0.73
'
100(1-~- .9615) =~'132(1 - Keff2) , rho 2'=-(O.9709 -1)/O.9709 = -0 03 EO.33-delta rho'= rho 2 - rhol ='-0.'03 -(-0.04),= 0.01 = 1000 pcm CO.63
'
.l
.h . Boron delta rho =--150.. ppm x -10 pcm/ ppm = 1500'pcm CO.53 .;
'
Xenon delta rho = 1000pcm 1500 pcm = -500 pcm [0.63 REFERENCE j
... '-Rx?Th HTFF, 1..of 2, Reactor onerations,.pgs 8-16 j e i . 3.1 001.000 K 5.28 3.5/ ^
EO -[ Rx Th 53b ,53d ,62,63 ~
) .OO1000K528- ... (KA'S) l l
,
t t
. . - , . .
..
.
]
y; - - - - - - . - _ _ _ . . _.
". '.; i . . . , l ki__ELGNI_EYSIEMS_p_E@l@Mx_QQNTROL3 ;AND INSTRUMENTATION PAGE 24 l . , I
- ' ANSWERS:-- MILLSTONE 3 -87/OG/18-SILK, D.- ' j .
i
' ,
i
, ANSWE .01 (2.00) j -
i
'a.: Fail' closed I b. Fail open j c . -- Fai l open {
:d. Fail closed e. Fail open .s ?
-f.. Fail open EO.33 pts each].
REFERENCE
'I AOP 3562, Loss of instrument air, pg 2 "3.8 078 000 K 3.02 3.4/ l EO - Intro to I&C 7 l ILC. failure analysis 1 I 078000K302 ...(KA'S) . ANSWER 6.02 (2.50) ,
l a.' Raises the limit CO.53 (because high dT indicates a higher power) Increases CO.53 (to' raise pressurizer level to-100% program, because of.the' higher Tave) 4 s c.. Rods move in C O . 5.] (because of the Auct. Tave/ Tref mismatch) 1 No effect E0.53 (the demand signal is present (Tave/ Tref) but ] there is no arming signal) , l
.e. Decreases setpoint [0.53 (because high dT indicates reduction in DNBR)
REFERENCE MP3, NSSS vol 3, Temp Ind., pgs 6,7,8,14,15,16 l;
!
3.9 016 000 K 3.02 3.4/ l K 3.03 3.0/ l K 3.01 3.4/ l l EO - Temperature indicating system 4,5,6 I I&C failure analysis 1 i l 016000K302- 016000K301 016000K303 ...(KA'S)
'
l l l l -_:__
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~- ~ -
'
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--- 7]
PAGE '25 1 "[n$ii268 NILS.YSIEdS_
. . .
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DES 10LI99dI5961_GN9_INSIB95gNIGIJ9d
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4.
I ': D ANSWERS 0--E' MILLSTONE . /08/18-SILK, D. : in s * II:
)7 u
s
. ' , ,f ypl; , #
ANSWERL is. 03 ! + ( 2.' 00 )
, , >t M- . 1 [e lMFW jumb.s~tr'ip' ,. .,ml 6ppyr '
o
;
Feedwatersregulatingjvalves clos : Maini f eedwater Lisolation cvalves - close
*
< ,
. !
'
MDAFW-pumps start: -l
:MDAFW:.pumpsi dicharge cross-connect valves open ~
-
;
TDAFW; pump lsteampsupply:, valves ope ~[0.33 pts each3
' '
tREFERENCE' ,
,
MMP3,. BOP:vol . 2, Feedwater,,pgs 5,8,9,12 T.BOPfvol.~- 3,3 AFW , > pg 5 23.5'059.:000 K.4.16:- '.321/3.'2
~
K 4.19'- 3.2/ , 061'000 K 4.02 24.5/4.6-
--
EDI-: MFW system 31_ 059000K416' '059000K419 061000K402' ...(KA'S) ANSWER 6'. 04 ' (2.40) sa.: . Doth bypass breakers will ' open resulting -in a reactor. trip , [1.~ O3
'
;b. . 1. TG-trips 2. Val ves - cl ose - !
. 3. Steam dumps wi.11' operate in the load re.iection mode
-
_4.- . No'effect CO.35 pts eachJ
.i t
i
'
REFERENC MP3, NSSS vol 5, RPSAS, pgs'20,21,65, Fig. RA-11 f NSSS..vol~3, Steam dump, pgs 10,11,19, Fig. SD-12 3.'11001:000 K'i~.05 4.5/ ~.K 2.02' 3.6/3.7-
K 6.03 3.7/ l i EO .Rx' protection L. safeguards actuation 7,8,15
~OO1000K603- OO1000K202 OO1000K105 ...(KA'S) .l
)
l, I
! /
-- - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
bz_iEL8SI_SYSIEUS_DESION4_G9 NIB 064_8ND_INSIB9dENISIJgN PAGE 26
* ANSWERS -- MILLSTON /08/18-SILK, ANSWER 6.05 (2.50) Remain the Same L 2 Increase
- 3 Increase l 4 Remain the Same CO.25 pts each3 Hi Hi flow signal CO.23 closes the return valve (AOV-178) [0.33 and as pressure increases a check valve on the RPCCW flow inlet seats CO.253 Isolates the non-safety train headers Splits. flow to the RCP's Isolates CAR cooling coils and neutron shield tank coolers E0.25 pts each]
REFERENCE MP3, NSSS vol 3, RPCCW, pgs 2,4,5,13,14,21 NSCS vol 1, RCP, pgs 4,21, Fig. RCP-15 3.10 008 000 K 1.02 3.3/ K 3.01 3.4/ EO - RPCCW 1,2a COBOOOK102 OOOOOOK301 ...(KA'S) ANSWER 6.06 (2.50) See Figure la EO.15 pts each valve] REFERENCE MP3, NSSS vol 2, ECCS, pgs 82, 83, Fig EC-44,48 3.2 006 000 K 4.06 3.9/ K 1.05 3.3/ K 4.03 3.4/ EO - ECCS 1,4,5 OO6000K406 OO6020K105 OO603OK403 ...(KA'S)
. _ _ _ _ _ _ _
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_ _ l:: _ _ _ _ _ _ _ ,: [ .a, _ _ - _a5iy UV^ ti
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _______ . 61__eL6MI_gygIgdg_pESJQL_CONIBL_6MD_JUSIBLJUENISIJQU PAGE 27 , -
.. .- * ANSWERS'-- MILLSTONE 3 -87/08/18-SILK, .
ANSWER 6.07 (1.60)
.
With COPS armed, the block valves will be open [0.63 so when TE 413B fails low the Train B PORV (456) will open CO.53 and thus RCS pressure will decrease CO.5 ' p C l REFERENCE 1 i OP 3208, pgs 25,26 l MP3, NSSS vol 4, pgs 13-15 3.3 010 000 K 4.03 3.8/ K 1.03 3.6/ EO - PZR & PRT system 3,4,5 I&C failure analysis 1 p 010000K103 010000K404 ...(KA*S)
.
ANSWER 6.08 (2.00) CuiiLr vi , uda vu tiidi un C Juc to th; tc;preture cr r- c :d pc r -i - maluii addiim puniLivu iecultsiLy CV.53. Wi t!. m smal1 tiTC , re cter pco wMi ituu iopidly ; .. i J u r u n e.- c e.;hcct - c ; u l t :. ;n c h.gh n c e r an f flus Lrip CO.3 REFERENCE MP3, NSSS vol 5, I&C f ailure analysis, pg 43 3.9 016 000 K 3.01 3.4/ K 4.03 2.8/ A 2.01 3.0/ EO - Main steam system 3d I&C failure analysis 1 016000A202 016000K301 016000K403 ...(KA'S)
/c prJwr0 Confr) co)$ '.%crb (CA} due fc s ('o~k !'DbmAh sig n a l [03.]
s ,,qs-eccusay wichv, - n ., ~ h. 6 +<- pe., w c s doc <c. w E n
~[195 (Cthe wbsk ShtstNb ud mm c f t:( om @ p m ,,n te f e m (0. Q
{cither crmors cev9)c, e) en .. c sa h, c. .g b,,,,, )
_. . _ _ _ _
. -. , . .
g s . Em t . g' V61__ELBUI;SYSIEMS_DESIQU4_G9 NIB 964 6Np21NSIB90gNIBIJg _
-
PAGE;;28 @ g" ANSWERS' ll !4 ..'MILLSTONE 31 -87/OD/18-SILK,1D.,:' ,
, 'k -t ,
l[m-i
' . )
1 I /I, I
{. Al , e I
s,. p A 9 ANSWER:- L J6. 09; (2.00).
e 2 akSourcerange.[O'.53'orL. intermediate. range'highfluxtrip,to.53'-
'
', - b.1The - static ' transf er switch will., automatically! transf er ..to receive ' ;;- -- power l.from the: bypass regulator :[1.03 ' ' ~
-
l{:. .
! "; REFERENCE .
o i
, _(AOP 3564P Loss:ofione protective system channel,:pgJ3 " IMP 3,; BOP.1, 120VAC,.pg 3 ; 3.7.062.00'O.K 2.01 3.3/ , K.4.10 3,1/3.5- - 3.910121000:K 2.01- '3.3/ 'IEO i- 120.VAC -Instrument Power System 3 ; .062062OOOK- OOK410- 2012010620 ...(KA'S) " ANSWER- 6'.10 (2.50)
Ea. ' - 1 ) YES-[0.253...P-6 drops out'(because IR < 10 E-10. amps) [0.25] and
.
.SR:NI reenergizes causing a level. trip signal [0.25]
2)'.NO'[0.253. P-10 is=in effect CO.25] which'will prevent the SR' from reenergizing EO.253.(even'though P-6 has dropped out) b.cNianually' reinstate the SR detectors with the SR Block-Reset switches
'
, , [1.03
. REFERENC MP3, NSSS vol-3, Excore NI, pgs 18-21
~
- 3.9 015 000 K1.01- 4.1/ K3.01- 3.9/4'.3 K4.01 3.1/ K4.07 3.7/ A2.02 3.1/ EO "Excore NIS 3,6,7
- Its.C failure analysis 1
.
! l.:015000A202
, . _ _ _ _ ~ _ - _ _ _ _ . - - _ - . - . _ . . ~ _ -_ _ _ - - . _ . , - _ _ . . - _ . _ - - _ _ _ . _ _ _ _ .
gmy ; , - ,
- ~
gi- a; y7 y- 31 m . . . .. . . . . .
.
. .
.l
# & dfL6NI[.. .$YEIEd@_QgQ1QNx_GQNIBQLt_68Q_lN@l8QdENI611QN PAGE 29? y nn. -
([m'AN'5WERS >> r0 , L --J f11 LLSTONE .l3 ,
-97/08/18-!'1LK,- .) 1 6 _ .
..
&.: < r l: n 1; <
jrb . ., ,, . . , . .
?..iANSWER > 6.11' =(3.00) ;;l v ' (. ~ '
e..When theLEGLS is reset to.53 OR ai change in; operating cor.ditions occut*
, :m . .. .. . ... . .
CO.53
.b.t.;The;EGLS wil1; sequence on equipment required.by theLnew candition with- .
k P "out! stripping',the running equipment:already running: [0.53.and trip equipment not normally. sequenced:en du' ring a'CDA/ LOP, accident CO.53.
W2-{c.-Durfn_g/ Test 1,the EGLS will reset to normal:and carry out its designed function'whereso during Test 2 EGLS-will not reset. C1.OJ
. .
.i
. REFERENCE '
.MP3,jDOP 1,/ Diesel generator . sequencer, pgs 2-5,. Fig EGS-1:
L3.7~064.000 K'4.-10, 3.5/4.OJ
'. 'KJ4.11- 3.5/4.O: -K 3.01 3.8/4.1: . k0 --'EGLS 3,4,5,6 '064000K301 064000K410 .064000K411 ...(KA'S)
9
.
.I
- _ -_
_ - - _ - . -
,I <
c, o . :: - 1 * .. .
.. .-
ZiL E899EDV8gg ~NORd8L1,,6pNQBMBLi_EdgBQgNgy_8ND-
-
PAGE .30
f!M;86D1969@1966G9dIB96E l m .+ E LANSWERSI--lMILLGTONE-3, '
*
-87/08/18-SILK, D.
6,'. w '
<
t 9- c ' j" g '
.
-
,
'
l- ,t', F t* - ANSWER 1 ;7.01 1 ( 1. 00 ) ? ? LA. $ , B .} J.1'
'
& C. l'4"
* '
l > -. D. :. J 2 ' 'CO.25 pts each3l- [TREFERENCEI LNRCL Read 5 and -- Si gn : Pragram , ..pg . ~15
' -(ACP-8.02) (OP-3256)-
M l Plant' Wide GenericE194001~ K 1.16' -3.5/ ;EO - ACP-QA-2.05 194001K116 ...(KA'S)
: ANSWER 7.02: - (2.00)
s 5(24-18)~= 30 Rem;. Lifetin.e limit = 30 - 28 = 2 Rem [0.53
' With Form 4. on- file' he is permitted 3 Rem /Qtr - -Lifetime limit is more restrictive [0.53 iO.4-Rem /hr + * 2A egr/hr x_ ~W= 2.4 Rem /hr [0.53 2.O' Rem /2.4 Rem /hr = 0.83 hrs =-50 minutes CO.53 : REFERENCE'
410CFR20.201 10CFR20.202 q
. .
194001 P1 ant' Wide Generic K 1.03 2.8/ { K 1.04 3.3/ ) i EO -.HP proceduren 1 J
-194001K103 194001K104 ...(KA'S) j - :_ : __ =-
, , _ . Z___EB99ED9BES_:_N9Bdob2_8BN9Bdeb2_EdgBGEN9Y_6ND PAGE 31: f,; .B09196991996_99 NIB 96 l
~ANOWERS -- Mh LSTONE'3' -87/08/18-SILK,. l
' ANSWER 7.03 (1.60)
. An HP tech CO.33 and another-permanent staff person.CO.33 I l
i Emergency repressurization for personnel safety or fire brigade entry CO.53 Restore CNMT integrity within 1 hr (OR-be in HSB within 6 hrs) CO.53
'
s REFERENCE AOP 3568, Emergency Dreaking of CNMT Vacuum, pgs 2,'3 OP 3212, CNMT Entry, pgs 5,6
,
3.6 103 000 A 2.05 2.9/39 ; G 15 3.8/ G1 3.5/ G5 3.3/ EO - AOP 3565 1 ACP-7.05A i 103OOOA205
. ANSWER 7.04 (2.00) NO [O.43 To avoid having more than one spent fuel assembly out of the core in a vertical position during a potential loss of cavity seal event [0.63 Flow can be suspended for up to i hr per 8 hr period CO.53 during per-formance of core alterations in the vicinity of the reactor pressure vessel hot legs CO.5 REFERENCE OP 3210B, Refueling Operations, pgs 5,6,7
)
j l 3.'11'034000 K 1.01 2.5/ l K 1.02 2.5/ G 10 2.7/ i G 13 2.7/ j EO - Fuol handling system 5,6 034000G10 034000G13 034000K101 034000K102 ...(KA'S) l - _ - _ _
, . , _ . _ . . . . - _ _
E l' 1.4 9f / f . ,, , . JU , _l
'
' LZi__EB9CED9BES_:_NQBd861_8pNQBdSL3 EMERGENCY ANDl !PAGE 32 .. lipLBBDI96991986_G9dIB9 <
.] 1 m . . . ~ ' . . . . ! ANSWERS 2--iMILLSTON f-87/OB/18-SILK,' ,
ANSWER 7.05' ( .00).
- a. . 1h The1RCS~has been borated;to CSD concentration
' The RCS'has been,barated to. hot Xe-free' concentration and is being^ maintained at no-load Tavg CO.5 pts'each]
l, - - " ;1. No. baron'. dilution >isfaccurring; t2.; Core. outlet-; temperature is'at least 10 F'subcooled' CO.5'ptsLeach3- jTo: prevent)"short cycling ' spray flow f rom loop .1 * CO.53 - To prevent causing' steam pressure transients that could result.in S CO.5] s: ( / REFERENCE: OP 3201,/P1 antHeatup pgs< 10,11,14,15,28 3.1 001 050 G'10 -3.3/3.5-3.3 010 000-GL10 -3.3/3.6' 3.9'012 000 K 1.07- 3.2/ .4 OO3'OOO G 10 ,3.3/ . EO - G01-01-COO 3,15,16
.OO1050G10 003000G10 010000G10 012OOOK107 ...(KA'S)
i
- . .2 - - . . ._
e , Zi:1EB99EDUBES__NQBdSL2_8BNQBd862_EdgBGEN9Y_OND PAGE 33 88D196991986_99 NIB 96 ~ l-
' ANSWERS -- MILLSTONE 3 -87/08/18-SILK, !
' ANSWER 7.06 (3.00) Two SRs, Two irs, Three prs CO.25 pts each] Station superintendent, Unit superintendent, and Operations superviser
[0.25 pts each]
> Terminate startup, drive in rods, commence boration (immediate borate)
[0.33 pts each] .
I Return Tave within limits within 15 minutes or i put plant.in HSB in 15 minutes CO.5] REFERENCE OP 3202, Reactor Startup, pgs 3,4,6,7 , I
. G 2 3.2/ ! G 11 3.1/3.8- G 5 3.4/ .1 000 024 EK 3.01 4.1/ EO GO2-01-COO 4,7,14 OOOO24E301 OO1050G2 003000G5 015020G11 ...(KA'S)
ANSWER 7.07 (2.80) Return flow increanos (above 5 GPM) CO.5] Lower bearing water temperature inc eases (approaching 230 F) [0.53 I I Close the affected RCP Seal Leakoff Isolation Valve (AV-8141X) ) Defeat Loop 1 Temperature Inputs (Delta-T, Tave) i Reduce power to < 37.5% (P-8) CO.6 pts each] j REFERENCE AOP 3554, RCP Trip or Seal Failure, pgs 2,3 l 3.4 003 000 K 1.03 3.3/ K 6.02 2.7/ EK 2.07 2.9/ EK 3.03 3.7/ EK 1.05 2.8/ EO - AOP 3554 1,2,6 OOOO17E105 OOOO17E207 OOOO17E303 OO3OOOK103 OO3OOOK602
- .a
g s--
. -
; - _ u , .. i ..
., .
~'b2L EBQGED9BES = N9Bd6Li 9EU9Bd8LLEDEB9ENGY 0ND
. _ _ ._ PAGE 34(
5,o _.BeDIRL991G86_G9
:
NIB 96 f ANSWERSL-- . MILLSTONE 3. ^ ,
-87/OS/18-SILK, %g, R - ,
lbNSWER 7.08 -(2.00) i.
lt . '", '
:E [1.JTrip turbin > > 2.. Verify.turbinefvalves closed-
_
:370 pen generator' output' breaker ,CO 4 opts each, O.2.for order . .
<
. .
:6.f Place EHC; pumps-in-' Pull-TO-Loc C O. 6 J '
l ' REFERENC " l-AOP 3550, Turbine / Generator Trip, pg-3
'
3.5;045L000 K 4.02 :2.5/ ;
.K 4.13' 2.6/ .3.14 000 007 EA 1.07 4.3/ .
d EO -1AOP 3550 2,3 000007E107- 045000K402 045000K413 ...(KA'S)
$ANSWER- -7.09 ( 3.10 ) . ',1 ,3 ,21 .[1.03 , The: Core. Cooling Red Path should be immediately addressed C O. 5 3 becausepit is of. higher. priority than the Heat Sink Red Path CO.5 ECA- CO.5] -d.- When directed :to in E-O CO.33 or when transf erring out of E-O CO.33 REFERENCE '(Usage Rules)
LECA-0.O pg-3 194001 Plant-Wide Generic A 1.02 4.1/ .EO - EOP development 6,7 194001A102' ...(KA'S) __ i
' ~,
_ __ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ - _ _ - _ _ _ _ -
'
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, i
, Eu;- . . .
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' ANSWER' 7.- 10 , 12.50):
an Unexpected'i'ncrease:in any SG.NR. leve l M ' Highfradiation'from any,5G sample-
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;Hightradiation'from any; steam line-
'High? radiation; f rom'.any SG blowdown J1ine [0.25, pts each3 To maintainisteam' dump operation'during RCS cooldown
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lb.- [0.53: ) i c. - :SI must bel terminated when-termination criteria'are' satisfied to
. prevent'overffiling the. ruptured SGfE1.03 REFhRENCEL LE-3,spgs 4,10.,15 313.000=037 EK 3.074.2/4.4-G 12 3.5/ o-EO.-'E-3lSGTR 2,3,6
, , 000037E307 OOOO37G12 ...(KA'S)
.v-ANSWER 7. iii .( 2. 00 )- Feedwater-flow <iS25 GPM SG11evel'< 34% [0.5 pts each] < If~a'non-faulted:SG is not available [0.5]
l _c. .'To cool the core using. cold leg recirculation [0.5]
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l REF'RENC E LEOP F-0.3-
;FR-H.1, pgs 3,9 13.5 OOOO54:EK 3.04- 4.4/ G 11 3.4/3.3 L .
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1ANSUER 8.01: (2.00).
, =The intent.of the. original ~ procedure is not; altered CO. 6 3 - The change'is approved by two marber: c' c .; ;
' , et 1 ::t nc n'~ ? r- 5:Id: : CDC 1 munwe ~
the w.<# p12-Ihu?erv men [c.E606 catfn,s athiT, ylie ahJ vmf_ fyr e4 Mt.m sn<ll & the en c hfy SyH sqervy, . .. The' change i s ' reviewed by PORC/SORC and approved by the Unit Super-f: -irdendent/ Station.. Superintendent within 14 days of i mp) ementati on C O. 7 3.:
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,' REFERENCE
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TS pgs 6-16,6-17 ') .ACP-QA-3.02, pg 3 ;i Plant-Wide-Generics 194001 A 101 3.3/3.4~
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'EO - TS.'17 q 194001A116 ...(KA'S) 1 ANSWER O,02 (2.00) Interval requirement not exceeded.00.5 Eight days.does not exceed 1.25 times the specified interval CO.5 !
b.- ' Interval requirement exceeded C O. 5 3. The last 3 consecutive ; intervals exceed 3.25 times the specified interval CO.5]. REFERENCE i TS 4.02, pgs 3/4 0-1,'02 .l TS 3.5.4, pg.3/4 5-9 l 3.2 006'050 G 5 3.5/ l
'EO - ACP-DA-9.02 5 TS 7 i OO6050G5' ...(KA'S)
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. g.j ii 1Thetwork 1s. coordinated between the'SS/SCD/SRO and the (1;. ' 'fJob Supervisor (sf CO.~661 l d/2..'Anyjworkers protectedib'y; additional" tags / clearances, (that.would be .
' affected byj-operation of the equipment), shall be tdvised andHremain in the'. clear'before the tags are lifted EO.673:
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' 3 . .- -(The lif tied Itags' shall . be1 maintained with the work- documentation
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-in the custody: of the .SS/SCO -(until the operational 5 evolution:
Lis completed)'.CO.663
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tREFERENCE ACP-DA-2.06A,-pgsj16-1 Plant-Wide. Generic l 194001 K.1.02 3.7/ i-Y 'EO - ACP-QA-2.06A.L56
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' i ANSWER 8.~ 0 4 -(2.00) IU'N IDENTIFIED Leakage limits ei<ceeded 3.4.6. [1.03 30.5 --(1'.5+1'.2+1'.5+0.8+4.2) = 1.3 gpm which is > 1.0 gpm [1.03 REFERENCE
.TSl3.4. l3.2lOO2:020 Sys gen .6/4.1'
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% ,, pea a 4 +;n nas cand % hcab fanM Wh, a c.ft on is ' M<n i c . resfo rp 5 N C ""U /c Mi4 ^ ' ANSWER- 8.05'- (2.10)- m;4; mum r cpe,ne.f g,g Ar, RO f. u... t h m r ,- m v . w m m . lu iL m l; iumain vu July EU.633 iau L w ur k .. 4. aur 1..-it; .Tuat bm m m, i m J m, ma ; b . 6 .,3 gg r ! :: dict: ac t i cr, 0.3: .;.m m L L,m Luku i Lv 1 wu Lui u Lhu mi n i n,um uniit w,
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b. Tb 3. 03 [c'1] REFERENCE ACP 1.19'pg 5 TS pg'.6-5 , i Plant-Wide' Generic 194001 A 1.03 2.5/ EO - ACP-6.12 7 ACP-1,19 1-TS 14 194001A103- ...(KA'S)- ANSWER- 8.06 (2.20) a
'. NRC NRB CO.30 pts each] If the second verification would result in significant radiation exposur CO.53 SS and Duty Officer CO.3 pts each]
d.- If identified and controlled in another approved procedure CO.53 REFERENCE ACP-QA-2.06B,'pgs 7-10, 1 l Plant-Wide Generic 194001 K 1.02 3.7/ EO - ACP-OA 2.06B 1,5,7,16 l 194001K102 ...(KA'S) j l I 2______ _ _ _ _ _ _ _ _ _ _
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lacs all LCO's: are met- f or that cperating mode . withctut reliance on an: cction statement.EO.73,-therefcro the startup;SHOULD NOT; proceed..[0.53.,
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: ANSWER > 8.081 (2.50) , dr '
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LOperatingLShift Supepvisor l, b'
-Superv sing Control Operator Plant Equipmpnt Operator ',') ,'
Reactor Operator l'l' CO.33 pts each3 * . s'
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ANSWERC r- MILLSTONE 3 -87/08/10-SILK, .py .n > Di'.4 ANSWER _8 09 (2.50) . During an emergency to protect public health and safety C1.O] when l ncoother action consistent with TS can provide adequate or equivalent i protection is immediately apparent [1.O] or k4m pjiggn fg gog, 4, b.1 The NRC should be notified CO.5] 4*bt on occ;dmp *
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i REFERENCE 10CFR50.5% G o f' %E * n] Vs. Tcd (a s 1%,13 10CFR50.73 TS pg 6-15 Plant-Wide Generics 194001 A 1.16 3.1/ EO ACP-QAI)O.01 001 194001A116 ...(KA'S)
ANSWER ' O.10 (2.50) If the emergency condition is classified as an Alert, Site Area Emergency, or General Emergenc CO.53 [. Will assembly help or hinder the situation? Personnel done rates exceeding limits (limits: ' >10 mrem /hr or I-131 levels > 10 x MPC) ;r / Personnel exposure exceeding limits (limits: >500 mr em whole body or 500 I-131 MPC hours)
4.' Actual or potential personnel hazards exis CO.5 pts each] REFERENCE /'/2 EPIP AO)'2 0 A ', pg D; EPIP 4010B, pg Plant-Wide Generic 194001 A 1.16 3.1/ EO - EPIP 22 194001A116 >
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7 De.in hot standby.r,0,73 l
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[0.53 j Q[]Nofcit)/ the NRC in .1 hourNotif y the Vicke President-Nuclear Operations and j Chat ntian ;
Saf r+ty Rep.,cf NRB within edt prepared and 24submitted hours CO.53 (in 14 days)CO.53 ;
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q.} O e A TTMHt1EUT ]- NORTHEAST UTILITIES con on,c...s,w nsy,,,.e,n,n conn c,,co, p 9 1*U.2IS,'2.70'~C P.o Box 270 ( '
* [' $'d,$.,7,*[$.,,, ,,o. ., H ARTFORD, CONNECTICUT 06141-0270 E o,..c %. . .... c- ' ""
August 20, 1987
,MP-10745 ,
Re: NUREG-1021/ES-201/ para I
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U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555
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s Reference: Facility Operating License No. NPF-49 Docket No. 50-423 August 18, 1987 NRC License Examination Comments Gentlemen: Attached is the compilation of comments on the written examinations administered to Millstone Unit No. 3 license candidates on August 18, 198 These comments were the result of a review of the examinations conducted by members of the Millstone Unit No. 5 training staf Included are both the comments discussed during the exam review meeting of August 18, 1987 plus additional comments resulting from reviews conducted subsequent to this meeting. Attendees at the August 18, 1987 meeting were: D. Silk - NRC R. Martin - Northeast Utilitics R. Temps - NRC M. Moehlmann - Northeast Utilities The exam reviews were conductad considering the following: Does the question elicit the correct response? 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 !
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'Pleasescontact Mr.-Ron Stotts, Supervisor, Operator Training, . . Millstone Unit:No. 3, with any' questions concerning our ll . comments..
( Yours truly, NORTHEAST NUCLEAR ENERGY COMPANY M 0_- St phen E. Scace Station Superintendent Millstone Nuclear Power Station SES/MJMajac Attachment: Senior Reactor Operator Exam Comments' and applicable references Lcc S. Collins, Branch Chief, Region'I B. W. Ruth, Manager Operator Training i l l i i
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H 1 SENIOR-REACTOR OPERATOREEXAM ' L THEORY ?OF NUCLEAR POWER PLANT- OPERATION, FLUIDS AND b ,4 THERMODYNAMICS-
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D e5.06 Agreedithat;the reason'for shutdown margin Lincreasingimay also:be that the control. rods' move ! further.out resulting'in'more. negative t;eactivity tinserted!on:a; reactor trip (Reference OP3209B-1) c5.07c' Agreedith'at aLdiscussion'of head loss-is not required for; full credit on this questio (Reference Reactor Theory HTFF 2 of 2,< Boiling Process,vPg 25) .
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5.10.bL . Agreed that~the. answer key addresses;the long term
- responseLand consideration will be given,'on a case by case-basis, for'an explanation,which includes Psteam initially increasing'due to Thot increasing on a loss of feedwater to the steam generato .1 Agreed'to change 1part 2 to "NO. CHANGE" and' accept both'"NO CHANGE" and " INCREASE" as acceptable-answers'for part 4 due to the guidance provided to
'the: operators in step 10.of EOP - 35 ES- ' PLANT SYSTEMS DESIGN, CONTROL AND INSTRUMENTATION 6.0 ' Agreed that. answer should be rods move'in due.to NI/ Pimp mismatch resulting in reactor trip on low PZR: level. (Reference-Instrument Failure Analysis Text pp 45-46)
7 PROCEDURES - NORMAL, ABNORMA L , EMERGENCY AND-RADIOLOGICAL CONTROL 7.03a Agreed to accept any two permanent plant people to accompany the NRC' inspector into containmen ( Re f erence . OP3 212-l') l
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le 7.10c Agreed that:an acceptable alternate answer may be as L llonguas the SI termination criteria in COP 35 E-3 L are met the-core is subcooled,,regardless.of.the ! conditionsin'the reactor 1vecsel' head, and.SI may be l : terminated.- (Reference EOP'35 E-3 step 20) ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS 8.01 AgreedEthat acceptable answer may include comment-that concurrence of two licensed' Senior Reactor Operators (SRO's) onerof whom'is the onshift Shift Supervisor is necessary to implement a non-intent chang (Reference'- ACP'3.'02 p. 30)
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8.05: ' Agreed;that? acceptable answer for (a) may be. shift crew complement may'be one less than minimum required for a period of time not to exceed 2 hours provided immediate action taken to restore shift
. crew.to within minimum requirements and that if this answer given for.(a), acceptable answer for (b) may be~in accordance with Tech Spec. .3.0 ~
Also~ agreed that reference to ACP 1.19 in answer key is not applicable to this situatio (Reference - Technical Specifications, Table 6.2-1) 8.08b Agreed to accept Security' Shift Supervisor as
' alternative answer-(Reference-ACP.7.04 p.3)
8.09 Agreed that acceptable answer is whenever following guidance in emergency operating procedures, as question inquires only "under what. circumstances" is ' it permissible to violate LCO, and does not request any further explanatio (Reference-EOP Format and Use Text pp 12-13)
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ATTACHMENT 3 NRC RESPONSE TO FACILITY COMMENTS 5.06a Accept comment 5.07c Accept comment 5.10b Will consider on a case by case basi ! 5.10c Accept comment 6.08 Accept comment 7.03a Accept comment 7.10c Accept comment 8.01 The correct answer will include the statement that concurrence of two licensed Senior Reactor Operators, one of whom is the on shift shift Supervisor is necessary to implement a non-intent procedure chang .05 Accept comment 8.08 The correct answer will include Security Shift Supervisor in order to receive full credi .09 Accept comment l l
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