IR 05000423/1986031
| ML20207S699 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/03/1987 |
| From: | Collins S, Keller R, Norris B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207S690 | List: |
| References | |
| 50-423-86-31OL, NUDOCS 8703200189 | |
| Download: ML20207S699 (140) | |
Text
{{#Wiki_filter:U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 86-31.(0L) FACILITY DOCKET NO. S0-423 FACILITY LICENSE NO. CPPR-113 i LICENSEE: Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06141-0270 FACILITY: Millstone Unit 3 EXAMINATION DATES: December 15-19, 1986 CHIEF EXAMINER: IY /4 %4-f7 B~ S. No ris, R actor Engineer (Examiner) Date' . REVIEWED BY: W
F0,f 1 'M' ff 7 R. M. Keller, Chidf, Projects Section 1C Date APPROVED BY: h/lh)d $d%1 6. J. Coirins,. Deputy Director, DRP Date SUMMARY: Five Senior Reactor Operator (SRO) candidates, seven Reactor Operator (RO) candidates, and one Instructor Certification (IC) candidate were examined during this period. All candidates received their licenses /certifica- , tion with the exception of one SRO candidate who failed the simulator portion of the operating examination.
8703200189 870310 PDR ADOCK 05000423 V .PDR - _ _ _ _ - _ _
. _. REPORT DETAILS i TYPE OF EXAMSi Replacement EXAM RESULTS: R0 SRO-Inst. Cert Pass / Fail Pass / Fail Pass / Fail ~ Written Exam-7/0 4/0 1/0 Oral Exam 7/0 5/0 1/0 Simulator Exam 7/0 4/1 1/0 Overall 7/0 4/1 1/0 CHIEF EXAMINER AT SITE: B. S. Norris, NRC OTHER EXAMINERS: D. H. Coe, NRC N. F. Dudley, NRC N. C. Jensen, EG&G T. J.-Kenny, NRC B. A. Picker, EG&G P. D. Swetland, NRC R. R. Temps, NRC .
'1.
The followire is a summary of generic knowledge or ability weaknesses which were .ted during the written and operating examinations. These , weaknesses are provided to allow an assessment of the effectiveness of the training program, and should be considered when developing future requali-fication programs.
Future examinations may audit these knowleges and abilities.
- The ability to use the Emergency Plan to classify accident condi-tions.
The knowledge of the amount of decay heat that is produced after a - reactor shutdown.
The knowledge of the method used~to cooldown the reactor coolant - after various changes in the steam dump system.
The knowledge of the effect of the Tavg signal.on various control - systems.
- The knowledge of the effects of an instrument failure on the Cold Overpressure Protection System.
- The knowledge of what actions should be-taken if a Response-Not-Obtained (RNO) cannot be satisfied during the performance of the Emergency Operating Procedures.
- . The knowledge of what conditions are necessary to remove safety tags from equipment.
2.
Personnel present at Exit Meeting: NRC B. S. Norris, Reactor Engineer (Examiner) N. F. Dudley, Lead Reactor Engineer (Examiner) T. J. Kenny, Senior Resident Inspecter, Salem P. D. Swetland, Senior Resident Inspector, Haddam Neck 'R. R. Temps, Reactor Engineer (Examiner) Facility K. L. Burton, Unit 3 Operations Supervisor H. F. Haynes, Manager, Operator Training R. F. Martin, Unit 3 Operator Training M. J. Moehlmann, ATS, Operator Training S. E. Scace, Station Superintendent, Millstone R. G. Stotts, Supervisor, Operator Training-i
3.
Summary of comments made at Exit Meeting: i It was noted by the Chief Examiner that most of the candidates performed very well in the simulator, both board manipulations and use of the proce-dures; especially noted was the effective and efficient use of the Emer-gency Operating Procedures (EOPs). Additionally, the candidates exhibited a good understanding and familiarity of the Piping and Instrument Diagrams (P& ids).
It was noted by the Chief Examiner that many of the SR0 candidates could not properly classify accident conditions using the E-Plan.
It was noted by the Chief Examiner that " Adverse Containment" is used in the interpretation of the E0Ps, but it is defined only in the background documents.
The facility stated that this would be incorporated into the next revision of the E0Ps.
. It was noted by the Chief Examiner that the cleanliness of the plant was much improved compared to September, 1985.
It was noted by the Chief Examiner that during one of the plant walk-throughs that two different revisions of E0P-3503 " Shutdown Outside Control Room" were in the plant.
In the Control Room Revision 3 (6/16/86) was being used, but at the Auxiliary Shutdown Panel Revision 2 (11/5/85) was in the binder. This item has been turned over to the MP3 Senior Resident Inspector for followup.
4.
During the simulator portion of the examination, an evaluation was made of the operators' understanding of the Emergency Operating Procedures.
The simulator scenarios included: - Steam Generator Tube Rupture - Turbine Trip with no Reactor Trip Main Steam Rupture - Loss of Coolant Accident followed by a Loss of Offsite Power - Loss of All Feed - Control Rod Ejection - - Failed Fuel with a Steam Generator Tube Leak Feed Water Rupture Inside Containment - Earthquake followed by an RCS Cold Leg Rupture - Results of the evaluation are: a.
The operators were familiar with their responsibilities and with the required actions during the emergencies, both individually and as a team.
b.
The procedures were performed by the minimum staff identified in the Technical Specification c.
The operators did not physically interfere with each other while performing the E0Ps.
i d.
The operators did not duplicate actions unless required by procedure (i.e., independent verification).
e.
Where transitions were necessary_from one E0P to another E0P or to other procedures, precautions'were taken to ensure that all necessary steps, prerequisites and initial conditions were met or completed. The operators were knowledgeable about where to enter and exit the other procedures.
5.
The R0 and SR0 written examinations and answer keys are enclosed in Attachments 1 and 2; facility comments on the written examinations, and the NRC resolutions, are enclosed in Attachments 3 & 4.
As noted on Attachments 3&4, the system descriptions need to be updated on a continuing basis as plant modifications are made.
Attachments: 1.
Written Examination and Answer Key - R0 2.
Written Examinations and Answer Key - SRO 3.
Facility Comments on the R0 Written Examination and NRC Resolution 4.
Facility Comments on the SRO Written Examination and NRC Resolution i .. . . _____ ______
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S.
NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _MILLSTgNE_2.____ _ REACTOR TYPE: _PWR-WEg4 ______________ DATE ADMINISTERED: _@6fj_2/15________________ EXAMINER: _NgRRISz_Dz_$. _,_ _ _ CANDIDATE: dk ([ .[ INSISUCIIQN5_IQ_C6UDID61El Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question ee indicated in pe entheses after the question.
The passing grade req; ires at least 7t^ ". in each category and a final grade of at least 807.. Examination papers wili be picked up sit (6) hours after the examinatior starts.
~4 OF CATEGORY % OF CANDIDATE'S CATEGORY __y66UE_ _I0166 ___@ggEL___ _266UE__ ______________C6IEQQEX_,,___________ _2E199__ _2Et99 i.
PFINCIPLEO OF IJUCLEAR POWE7 ____._______ _ _ _ _ _... _. _ PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW _2Ez99-_ _25sgg ___________ ________ 2.
PLANT DESIGN INCLUDING EAFETY AND EMERGENCY SYSTEMS 2.3. 4~b _2Ez29 ____--_____ ________ C.
INSTRUMENTS AND CONTROLS _ _z____ _2Ezgg__ _g5 99 ___________ ________ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 19g199__ Totals ___________ Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature g
F e . NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination r oorc. to avoid even the appearance or possibility of cheating.
3.
Use black ink or dart pencil gnly to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (i f necessary).
6.
Use only the p ager provided for answers.
7.
Frint you nar,e i t.
the upper right-hent corner of the first page ci each secti on of the answer sheet.
6.
Consecutively number each cmsw4r sheet, write "End of Ccitegory __" as appropriate, start each category on a neti page, write gnly gn gn; sidg of the paper c.n c; write "Last Page' on the last answer sheet.
, 9.
Numt-er each e r. s w a tu catecc y and number, 4cr eMample.
1.4, 6.3.
10. Slip at lesnt tir y line b e t w: e r each answer.
11. Sepa ate a r t.wer sheets f r om pad End plate finished answer sheets face doon on your desk c-table.
12. Lise ab brevi at i cm 3 oni, ii t h e. are common 1y used in iaci1ity ;_1 t e r_ s t_u ce. 13. The point voluo for eacn questien :s indicatec in parentheses after the
question ena c+r o e-use* aw a qu i c c-f or the depth of answer required.
14. Show all calculeticos, methode, or assumptions used to obtain an answer to mathematic +1 problems whether indicated in the question or not.
15. Partial credit may be gi ven.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16.
If p arts of the examination are not clear as to intent, ask questions of the e >1a.3 Ln er_ only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or becn given assistance in completing the examsnation.
Th 2 s-must be done a+ter the es: c mi nati on has been complete ____ _ _ _ _ _ _ _ _ - - - - _ _ _ - _ _ _ _ _ _ _ ___ .. - _ _. _ _ _ - _ - _ - -- - - - - - - - - . . 1G. When you coa.plete > otr ex an;i ne ti on, you shc.11: a.
(4s sea.b l e your e:: ami na t i on se f oll OWL: (1) E:: am questicos on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages ir:cluding figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to a7swer the examination questions.
c.
Turn in all scrap paper and the balante of the paper that you did not use for answering the questions.
d.
Leasc the e:. amination area, as dcfined by t h e-eneminer If a f t e."
. ieav;ngs you ee foand in tt i c a. = e. v.h i l e the e~ec.nstion is still n p"OgrEsh, yot } I CerWE rit ? y L. 9 c er.) ed ()' rE,. C. C.
_ ... Iz__EBINGIELEg_gE_UUg6E68_EOUEE_ELONI_OEEEgIIgN PAGE
z IUESd?DYEBUICg,_HEGI_IESUgEEB_6ND_E6UID_E6gy - OUESTION 1.01 (3.00) The reactor is at.15% power, equilibrium Xenon and Samarium, boron concentration is 1200 ppm, BOL, Control Bani: C rods are at 200 steps.
Using Figures 1.1 to 1.11: a.
What will be the final baron concentration if power is raised to 100%, ARO, steady state conditions? State which figures you use.
State all assumptionc and show all work.
(2.401 b.
How many gallons of pure water or boric acid will have to be added? (0.60) DUESTION 1.62 C.ON The Ur= : t 3 calculated Shutdown Margin is 10% delte f./I' acsuming the most , r reactive control rod is 1000 pcm.
The Source Ra r.o 1 count rate is 50 cps.
Show sl1 wer; and ut.ato a..y asna,ptitas.
Cons i der eacti situatior sepa ately.
, a.
What would t,e thc f i n e.: count rate i4 the shutdown banks are fully withcrewn? (-ssucc t"* thatdown bank rod worth is 5600 pcm.
b.
Wi + e t would be the finel cou,t rate if a 100 ppm d i l u t i or. of the RCS w e r t-performud? GUEDllON 1. 0 ". (T.75) A reactor at POL is critical at 10E(-9) amps.
Rods are withdrawn at 48 steps per minute for 10 seconds in bank sequence, a.
Calculate the SL'R i mmedi ately AFTER rod motion stops.
Assume differential rod worth is 10 pcm/ step, effective delayed neutron fraction is 0.006 and the effective precursor decay constant is 0.1 sec-1.
(1.25) b.
Compared to part (a) obove, HOW and WHY would the SUR change if the initial conditions were at EOL? No calculations are necessary.
(0.75) c.
Conpared with part (a) ebove, HOU and WHY would tho SU! change if it were j ust bef or e rod mo t i a.
is stopped' No calcul6tions are necessar y.
(0.75) (***14 CATEGOR1 01 CONTINUED CN NEXT PAGE
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t ISEEdODYNGdlCgt_SE61_lE6NSEEE_QND_E6QlD_ELQW - QUESTION 1.04 (2.50) a.
What is the suocooling margin of the plant if the f ollowing conditions exist: (1.00) That 587 F Tavg = 572 F Tcold = 557 F = Ppzr 2235 psig Psg 1033 psig = = b.
If power is raised from 50% to 100%, how will the subcooling margin change (increase, decrease, or remain the same)? Justify.
(0.75) c.
Which one of the following conditions would result in the smallest subcooling margir."
Assuma RCS pressure is the same for all cases.
Justi'y.
(0.75) 1.
Centrolled natural circulation cooldown immscistely following a reactor trip f rom loss of flow.
2.
Continued operation at 5% power.
3.
All RCFe'ere opc-r ed i ng e t n or n.al no-los: temperature after an e.:tendeu st.utdown.
. QUZETIDN 1.05 (2.251 If r eac t or power >= Increeser fsom 50% to 10J.r4, how wi.1 DI FFEREN11 AL rod wortr> chang.e (Increa=c. d ec r e a r,e, or remain the same) for the following cond i t i on L Justley your answer.
Consider each cese separately.
a.
Rod position and baron cor. centr a t i on sre held ccn st ant, temperature is allowed t c ct-t;e e a u, b.
Bo*on is held constant, rods are withdrawn from 150 steps to maintein temperature constant.
c.
Red position is constent, boron concentration is diluted to maintain temperature constant.
( t t' 4 4 4 CATEGORs 01 CONTINUED CN NEXT PAM (14*4) f
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- ISEBOOpyg8DIC53_ME6I_IB6BSEEB_8UD_E6UID_E60W OUECTION 1.06 (1.50) WHAT percentage of full power is produced by decay heat at the followino times after a reactor shutdown from 100% power? Assume equilibrium was reached prior to shutdown.
a.
One second after shutdown.
b.
One hour after shutdown.
c.
One de.y after shutdown.
CUEST I n:, 1. t C t2.0 ' 1he reactn i =, in Mcce 1., c'. E!.utdown banks are full out, and tr.e control banLL ace withdrauri to 18J stepu or. Bent.
C.
Using the belc. information e.id F1gcr e !.12, d e t e r r.;; n e the criticai barcn conceottation by e.
1/M plot.
Sha sll wari: bor o. : -co entration Lourt Rete 15;' ppa 100 cp.> tU 125 13:9 1 t,5 1 Of_ 0 240 110: 445 10 :" > 333C QUESTION 1.00 f r. ->:)) Fiquee 1.13 show: e s y s t e r.
.or ce tnd t. h t-e characteristic curvem for tv, centrifugal pt sz.
Eact. pung is eenpped wi t!i e c. h e c k. valve on its d i t,c h a r g e. De i c aii r.e the s ys t en. pressure and the flow rete for: a.
The two pumps in PARALLEL.
b.
The two pumps in SERIES.
QUESTION 1.09 (3.00> Thr-re are THREE primary per c meter s that affect DND and can be controlled by the Rea c t o: Operator.
Assuc.ing four-loop operaticn, brtcfly cuplaire h u r, an increats in each of tt e p a r e;.u.e t e r s affects D NF: if the other two p a r &.:.-5 t er t remain constant.
(*#t** CATEGOR t' 1 CCU T I NUED Of.' f1E ri PAGE *8144)
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ItEEd991E5diggi_HEBI_IEBMEEES_6Sp_ ELL'Ip_ELgg -- GUESTICN 1.20 C. OC i The plant ic being cccled by Natural Circulation.
As the fission product dccay hcet decreesen.
Describe HCW and WHY the below parameters will changa.
Assume that E/G pressure is being maintained at 1000 psia.
l l a.
Tccld
b.
Th t I c.
Cerc'dcits T
Lac; tranrit tina ! l l I ( * * * *.R END OF CATEGOEv 0;
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. !1 QUESTION 2.01 (3.00) ! 'a.. List the automatic start conditions'for the Motor Driven and Turbine ! Driven Auxiliary Feedwater Pumps.
Include the logic where applicable.
(1.50) b.
' List, in order of. priority, the THREE sources of water-available to the AFW pumps.
(1.50) i OUESTION 2.02 (2.40) r a.
What FCUf-: conditict's will automatically close the steem generator blowdown cente i nme.,L i sol a ti or valves (CTV 2 2 s '> S (1.60) i b.
Whet TWO conditican will cutomatically closh the blowdown flow control vel s ea ( H'. ' 20c)^ (0.80) ' QUESTICN 2,07 (2.00) ? For each case b e i r-. esplein the essultant method of reactor cool an t s ys t e ct-temparaturc ccnttc a r.d indicete the approximate final RC5 Tavg.
Assume at: sy s t ea.s a r e.
norms? c:; c e p t es stetet and that no operator action is taken.
Consider each case separately, a.
The norme' set poi:.t on tbh steam dump system is reduced by ES psi while :n Hot Stand;' y anciting reactor startup.
b.
The Train A utsc, dump i nter l oci switch is taken to "OFF" while s t ata l e at 0% ree-tor pover.
c.
Reactor power is 100%; the pressure setpoint on the steam dump system is raised to 1200 psig and steam pressure control is selected at the , ' same time a reactor trip occurs.
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. QUESTION 2.04 (0.00) Given the below conditions following a Loss of Coolant Accident, which 4160 vac pumps will be running? Be specific.
RWST level: 200,000 gallons Containment pressure: 18 psia RCS pressure: 850 psia Time since accident: 1 hour Loss of all offsite electrical power B Diesel Generator out-of-service DUESTION 2.05 (2.80) a.
State whethcr the Lelow l oa d s a r c,- nor r.,s i l y supplied by RFCCW 1 rain A, Train B, or Both.
(1.60) RHf heat exchanger Reactor coolant pumps E >: c e ss l e t doo.n heat enchanger Letdowr hest exchanger Boron retove / ssstem Chsrgir.g p u o,p coaling sucge tant Saeet i oj e: L i on cooling surgo t er.i Fuei paal coole-6.
For the above loads, state which subheader is the n or r.ia l supply - Containment. Safety-Related, or Nonsafety-Related.
(1.2O) CUESTIOfJ 2.06 (3.25' a.
Describe the flowpath used to fill Accumulato-Tant-A when RCS pressure is 1500 psia and RCS temperature is 300 F.
(1.25) b.
What are FOUR basis for the nitrogen volume and pressure limits of the Accumulators? (2.00) Ii4513 CATEGORY O2 cot!TINUED O';fEyT PA3r 4*A44'
_.
.. . 21__E66NI_ des 193_INCLUQ1NQ_g6EEIy_6BD_EDES9ENCY_Sy@IEU3 PAGE B .. QUESTION 2.07 (2.80) A heat detector in the East switchgear room has activated an automatic initiation of the fire suppression system.
a.
What precautions should have occurred to warn personnel of imminent or past CO2 discharge? (2.00) b.
HOW and WHERE-can the discharge be defeated? (0.80) OUCETION 2.OS (2.50) a.
If tt e Instrumeat Air IIA) pressure inside cont 6inment starts to decrease, what 1WO outomatar actions should oc cur to mainLair the pressu m? (1.00) b.
On o c ompl ete losw of IA, how will the fo;Icwang velves fails (1.50) 1.
Excess }etduwn hand control vC ve (HCV-123) 2.
Di'e se l Generator s-er v: ce water outlet velves (AOV-39 A E. B) ~. Feedwater Regu1 M : r g val ven (FCV-510, 520. 530. L 540? 4.
Atmospherie: dump valves (PV-20A. Lt.
C, &D) ". L1quiC west o di scher ce val w ( H'/- 7 7 ) CUESTION 2.07 (2.25) The plant is ir Hot Stendby'. a.
Uculd a Con t a i n r.en t Isoletion Phase A signal CaJsF the Reactor Coolant Pumps (RCPs/ to stop? Justify your answer essaming no operator actions.
(1.00) b.
Would a Containme..t Isolation Phase B signal cause the RCPs to stop? Justif y your answer assuming no operator action.
(1.25) , It***A END C r CATEGORY O2
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(o. ) w.n z J.. .- -. Wsuld the below events cause the INCICATED nuclear power to increase, dccreesc, cr r e r.e i n the usmc a s sar.i n ;; thet the actual pcwer remains unchanged st 75*. befcre and after tha event.
Assume all ccntrol systems arc 2n a u t ar.c t i c. J a = t i -f, ycar an s.< r;. Considc -.sach event separataly.
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stic.
'r _. - L ; I c: f i s i r.cr es red to ir.'*. and r steser pressurc e de:. : '. ; - Jc: '* 1 2 '2 si;c; c al t h e-25*: value.
Eaplaim, in detail, HOW and h" thin n : : m rf_ct f: c *. c a r. gt: atar Ic.c. Assu.Te n Opereta-action.
_ QUEETION 3.04 (1.5C U tc t THREE. p cress radiation moni.or alarms that would be indicative of c stea:
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. a. l.c T,. J, s, v.C-(v v0s n a _ For each cf the following situations determine whether the rod control interlocks should prevent cutward rod motion.
For those conditions whers outward rod motion is prevented, explain what actions must be taksn to reerstablish cutward red moticn.
Assume rod control is in manual.
Consider each casa separately.
l' a.
An intermediate range nuclear instrument fails high during a plant startup with reactor power at 10%. b.
A power range nuclear instrument fails high during a plant startup with reactc: pcwcr at 60%. c.
The turbin; first sta;s imp;1ss pressu-c trarsmitter fails low dering p: ' ut;~ tw o i t h r t e c. '. c - p c-e-et CC'.. e s d.
E i C J 1 t a h E J _'. d l, Ltcp 1 Th fc.ils high and Lccp i Tc fails low while < c,e vting e; 'CO. powc. u,2 :,-n, . a,- - s: s. u s,., . ,- n m , e.
A s e. : thr pisnt is shutda - at 503 ~ and 20C0 psig and the Ccid C.crpresssre oratectic- (2:r' Eystem is inad.crtently aracd.
Will the CCC syster act tc re d.::e pl ant oresnu^2 if a locp Tc instrument fails ice-Lctify y t, s : 6.r.c u c;. ancludt hco ca:h Train is Mf ected, and what e c h 0 7.
taLII ple e.
(2.00? t.
Ll:. c '. a e r rJ r., r.
~2 A d tt.c F a c.c t c - Operator have of the condition . de u c e. h :< ir ;a-t 'ai at t n (1.00) ~ O.JreT4,,h, (.., ,,, - . ..m, _., Sketch tha ele:trical distribution to bus VIAC-3.
Show all alternate scartes cf electr: cal power (5 AC sources.5 1 DC source).
Indicate bus voltages, show all breakses, and label all buses and components.
Breaker numbers are not required.
Assume that all of the service station , transfernsre (NEETs ?, R3ETs) are deenergi ad.
> I l
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' .. Ei= INEIEMOgNIE_6GE_q2GIEO'_E PAGE
.. QUESTION 3.05 (4.00) _Give the' basis for-the following. protective actions.
Indicate which of the ectier.s can bc blocked, identify HOL' and WHEN'the Llock it-sccomplished: a.
Sourca Range High Flux b.
Power Range Hi gh Flux - Low c.
Power. Range High Flux Rate - Negative d.
High Pressurizer Leve: Centainment Pressure High-1 e.
. f '.. Eitar. Gener ator Lcw-Lew Level -, _... ne, . d,- [.' .-) v. a t* i t i *-
> Th? ca: tar _ s at 100% pows with all central s y'E t e.m s in auto.T.atic; and t_,c. : ten #rc,ing F---~.,e.---. 4t x .
,,Re_
c,ianne,. - t.
< r ac , c u,a n n e l ,ee, . u sm s fails a. gh.
necam . i-,
- n y o'.w answcr ir. 1ure. s!)the plant response, (2)what F:FS/EEF signals will b.
- Ire ettd 6 r. d (3'the fi m
- 1 appro i mate RCS pressurc.
Es tp oi r.t s a e n st r e qu i r e:. A sums no operator a:ticn.
- J (M t t t END C~ C,S.~ECCRY O! 11i11)
_..._.
. i.___ ESC C E DU BE l_;_UQBU66 t_8 EUQ BU GL t_E UEBG E NC Y _6M D PAGE
-E60106001GOL_COUIBOL QUESTION-4.01 (2.50) You are working inside the. containment while the reactor is at power. The rad i a ti ca fiele is.850 mrem /hr gamma and 30 mrad /hr thermal and fast neutron.
You ara 28 years old 6nd have a lif etime exposure through last quarter of 48 Rem on your NRC Form 4; additionally,.you have accumulated 1.5 Rem so far this querter, a.
How long can you wort: in the area before exceeding your 10CFP2O limits? Show all work and state all assumptions.
(1.50) b.
During a declared emergency, you volunteer to enter a high radiation area and pc. -farm work necessary to prevent further effluent release.
Irr acc or danc.-< u) t i. the t'P3 Procedures, what i s year maximum allcwed whole I; o6, expo wre? (0.50) c.
In ac cordanc e-with the He,d th Physics Procedures, whose authorization - is require < in pnl t.' (0.5G) OUESTICt! 4.02 (3.00) For each of t h e f o f I om r u,; si tuat i or L, deterraine whether the plent should be TRIPPED, SHU1DOWN, FEDUCED IN POUEP, c-ALLOWED TO CONTINUE AT POWER.
Assune the-plant is opesatir.g ct 10 t M'. pownr er:d com-i der ecch case - separately, a.
Instr ument air preswre falls to 90 psig.
D.
High RCS activit.y is indicated by chemistry sample and a failed fuel monitor alarm.
c.
Number I sesl cn a reactor coolant p u rr.p has failed.
d.
Pressurizer level is slowly decreasing due to a leak, one charging pump is in operation.
e.
Condenser vacuum decreases to 8.0 inches HgA.
f.
The contaioment pressure has been above 13.O psia for over one hour.
(t*11X C AT EGOR'. 04 CONTINUED ON NEXT PAGF 41114) r.... ...
. 4.
PROCEDURES - NORMA! t_AgyORMAlt_EMEEGENCY_ANQ PAGE
" EODIRLggICeL_CogIRg6-QUESTION 4.00 (1.50) What is the. reason for the follcwing note from EOP 35 ECA-0.0 " Loss of All AC Power": "CSF Status Trees should be monitored for intormation only."
OUESTION 4.04 (2.40) a.
What would happen if RHR were initiated when RCS ten.perature was 390 F? Why? b.
What woult happer. if PHP werc initisted wh er. PCS pressurre we 500 psi a' Wl; y , DUESTION 4.05 (2.25) In accordar.ce oi th ihe O P--3 2O 2 "feeattor St e r i e p and the (1P2.
Technical Speci f i c a ti on s : a.
What is the M : n i n,,ur.
Terr per a tur e for Cr i t i c e l i t(' ( 0. 5 '> > b.
Whet action must be tat.en in Mode 1 if the temaerature ic less than the fli r.i mum Ter:iper e t ur e for Criticality? (1.50) QUCSTION 4.06 (.'. 0 0 ) In accordance with the Emergency Operating Procedures: a.
What is the difference between a FAULTED steam generator and a RUPTURED steam generator? b.
What constitutes Adverse Containment? (*t*** CATEGDRY 04 CONTINJED ON NEXT PAGE *t848) /
_ _ _ _ _ _ _ _ _ _ _. - - ,
5:__[695EDUEEE_ 6980Sbs 6EU98066A_EdEBgEUgy_6BD PAGE
66 RIP! Os1C66 GOBIBOL QUESTION 4.07 (3.50) Answer the following in accordance with ADF 3566 "Imrrediate Boration": a.
List FOUR conditions that require immediate baration.
C. OO i b.
If boration flow is at least 33 gpm, how many gallons of 4% boric acid will have to be added to raise t '. -. -._ by ' ' 100 pctn? Nac'd*% (0.50) c.
Describe the TWO flowpaths available for immediate baration, indicate the preferred and alternate.
(1.00) Q:.JES i I GN 4. OEt ( -' m . Whet are the operefor's ac_ t i on s ie, t.FTER refrw rin to e " Response f40t Obtainec. (RNO)" action etep. the s pected RiiO r espoase cannot be sat.siied? l ' ,
CUE 31IGN 4.ui ('2.50i 4.
I r, acc or de.ic e n.th t he fo'r-o Linat e of the-Emergency Response c Pt ocedeser, WHEN at s ttc operatore tequired to s '. a r t ir.cn i t or i ng of LI e Cr-i L '. c e l Sa+e: P o:ic t i o. i Stetus Trees (CBFSTL).
( 1. Oro
L.
Frio-itize the b.= : ow lidt A C r i t. i c s ' Safety Function Status Trees f r on highest t c.
l ouest. (1.50) 1.
Conte.nmcnt -- F.
". '. Cor e Cooliog - Red 3.
Heat Sini - Yellow 4.
Integrity - Orange 5.
Inventory - Green 6.
Subcr i t i t.a1 i t y - Orange QUESTIOtJ 4.10 (1.60) It is discovered toca>. 15 December day shift (0800 - 1600). that a monthiy surveillance due last Fridar, 5 Dr:cember mid shift (0000 0900), was - not per f or rned. Thiu survei1:i r r e has been per f or tned on tirw for the pact si: months.
In thin i t e_ s c orisi dered operable" E:: p l a i n your answer.
(*#t4i C /.TT.G U;:i' n4 CON t T NUED CN IJEXT PAE.F 14411)
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QUESTION 4.11 (2.00) Answer the following in accordance with ACP-QA-2.06A " Station Tagging"s a.
Who are the only personnel that are authorized to issue Safety Tags to listed quali fied indi viduals? b.
What is the purpose of a " blue tag" found on a valve? c.
What col or tag, if any, may be attached to a component which alreatly ' has a DLUE tag on it? (Wittt END Oc CATEGORY G4
- )
($ts444*4A4**6 END OF EXAMINATION *****4**44****: )
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a o- _ _ - _ _ _ _ _ _ _ _ - -. -. . - . N .. SAMARIUM REACTIVITY vs TIME AFTER SMUTDOWN ' SUILDUP FROM EQUILI8RIUM wHICM IS Stepcm N-975 . ~ -955 ~< ~ )F 100 / / - -945
7 ' ' ~ ' / EQUILWtluM POWER LEVEL " 925 e pnen 70 snuinown -
- 915-9o5 l f x-895 ,
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TASIZ 1 r - I l . P!ane Condi!!o-s Correction Tactor l - (K) ' Fressura T (AVG) (psis) (* 7) ?ressurisar Level (See, Nota) 2235 347-570 Nor=al Operating 1.00 - 1600 500 No-Load 1.05 1200 450 No-Load 1.10 '800 400 No-Load 1.16 ADO 350 No-Load 1.18 400 300 No-Lead 1.20 A00 300 Solid Vater 1.35 , 400 200 No-tcad 1.28 {"s 400 200 Solid Water 1.40 400 100 Solid Water 1.47 - , . NCrs: COREC51CN FACORS ARE APPLIED AS TOLLCUS: (a) Bor'an Additica and Dilution Total Voluma Nocc5:aphs
(Corrected) (Nomograph) (b) Soron Addition and Dilution Rata Nacographs fx (NonoETEP ) h(Corrected) h = . . . . . . f/tG H #9- /. 9 - - - -- - -- ---- _ _-- ____
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- 200 -
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100 10.000 ~ I a 1000-g - _ . - -
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== E _ - = a-50.000 1000 =' - - 2000 - - - 100.000 - 3000 . . . . Bore,n Dilution - Refer to Toble 1 for Correcticn Facten Figure 4.
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~* '. 1; _P81NQ1ELEg_QE_NQCLE68_EQWEB_E(@NT QEERATIQN PAGE
z ISEEbQQYN6dlCS _ME61_IB6NSEEB_6NQ_E(QIQ_E(QW t ANSWERS _---MILLSTONE 3' -86/12/15-NORRIS, B..S.
ANSWER- -1.01 (3.00) a.
Reactivity due.to power increase [ Fig 1.23 200 - 1350 = -1150 pcm (0. 40 )- Reactivity due to-rod withdrawal EFig 1.1] 425 - O = +425 pcm (0.40) Reactivity due to Equilibrium Xenon Change [ Figs 1.3 t< 1.43 1320 - 2760 = -1440 pcm (0.40) Baron worth CFig 1.8] =* -9.2 pcm/ ppm (0.40) No reactivity changes due to Samarium (0.40) Total: -1150pcm + 425pcm -1440 pcm = -9.2 pcm/ ppm (12OOppm - BC) -2165 pcm = -9.2 pcm/ ppm (12OOppm - BC) BC = (11040 pcm - 2165 pcm)/(9.2 pcm/ ppm) BC = 965 ppm (0.40) b.
Dilute '12,000 gallons CFig 1.113 (0.60) REFERENCE Reactor Theory, Ch 7 " Controlled Reactivity Effects" Ch'8 " Fission Product Poisoning Effects", pg 27 MP3 Curve Book ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ K&A OO1010K5.21/IF 3.4 OO1000K5.09/IF 3.5 .
.. 1.
PRINQlELE@_OF NQGLEAB_EQWE8_ELANI_QEggATIQN PAGE
z
- ~
IME80QQyN6dlGSz_HE61_IE8NSEEB_6NQ_E(Q1Q_E(QW ANSWERS - ' MILLSTONE 3-86/12/15-NORRIS, B.
S.
ANSWER 1.O2 (3.00) a.
CRdf) = CR (i ) E (1-Ki ) / (1-Kf ) 3 (0.25) Reectivity in core (i) = -10,000 pcm + (-1000 pcm) -11,000 pcm (0.25) = (f) = -11,000 pcm + 5600 pcm = -5400 pcm (0.25) SDM = (1-Keff)/Keff OR p= (Keff - 1)/Keff ; Keff 1/(1 p) (0.50) = Ki 1/(1 + 0.11)
==> Ki = 0.9009 (0.25) = Kf 1/(1 + 0.054)==> Kf = 0.9488 (0.25) = CR (f ) = 50 cps C (1 - 0.9009)/(1 - 0.9488)3 96.8 cps (0.25) = b.
Assume Baron worth = 10 pcm/ ppm (0.25) K (i ) = 0.9009 CR (i ) = 50 cps Reactivity in core (f) = -11,000 pcm + (100 ppm * 10 pcm/ ppm) = -10,000 pcm (0.25) K(f) = 1/(1 + 0.10) = 0.9091 (0.25) CR(f) = 50 cpsE(1 - 0.9009)/(1 - 0.9091)] 54.5 cps (0.25) = REFERENCE CAF ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO1000K5.09/IF, '1z__EBINGlELES OF NUGLEAR_EQWER_ELANT_QEERATIQN PAGE
u , .ISEEM9QYNQMIGS _ME@l_IBQNEEEB_@NQ_ELUID FLQW t ANSWERS'-- MILLSTONE 3-86/12/15-NORRIS, B.
S.
-ANSWER . 1. 40 3 (2.75) a.
Lambda = 0.1 sec-1 (0.25) Beta eff = 0.006 (0.25) Rho = (10 sec) (1 min /60 sec) (10 pcm/ step) (48 steps / min) = 80 pcm (0.25) After rod motion: - SUR = 26(lambda (rho)) / (beta ef f - rho) (0.25) = 26(0.1(BE-4))/(5.2E-3) = 0.40 dpm (0.25) 6.
At EOL, beta effective is less (0.50) which gives a higher SUR (0.25).
c.
Due to the rho dot factor (0.50) - the SUR would be higher (0.25) REFERENCE Reactor Theory Ch 4 " Reactor Ki net i c s", pgs 212-216, 235-244 & 263-268 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO1000K5.03/IF 2.9 OO1000A1.06/IF 4.1 015000A1.02/IF 3.5 ANSWER 1.04 (2.50) a.
Tsat for 2250 psia (2235 psig) (0.2O) = 652.67 F (0.40) SCM = Tsat - That = 652.67 - 587 = 65.67 F (0.40) b.
SCM will decrease (0.25) That will increase as power increases (0.50) c.
1 - Natural Circulation Cooldown (0.25) Core delta-T during NC cooldown will approach full load delta-t (0.50) REFERENCE Steam Tables MP3 NSSS Vol 1, RCS, pg 33 NSSS Vol 1, P:r & PRT, pg 16 HTFF Fund, Ch 3 " Thermodynamics", pgs 43-51 Ch 13 " Natural Circulation", pg 4 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ KLA OO1000K5.56/IF 4.2 , - _
'
- .
'Iz__EBINQlELES_QE_NQQLE88_EQWEB_E68MI_QEE88IlgN PAGE 19 I s
ISE85QQYN8dlGQu_ME81_IE8NSEER ANQ_ELUlQ_ELQW ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
OO2OOOK5.09/IF 3.7
ANSWER 1.05 (2.25) a.
Decrease (0.25 for direction, 0.50 for justification) Lower temperature increases the density thus reducing the number of l neutrons available for capture by the rods.
b.
Decrease There is less flux available for interaction as the rods move towards the top of the core c.
Increase Decrease baron concentration increases the number of neutrons available for interaction with the rods.
REFERENCE Reactor Theory Ch 7 " Controlled Reactivity Effects", pgs 30-34 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO1010K5.04/IF 2.2 OO1000K5.02/IF 2.9 ANSWER 1.06 (1.50) WMv (5-7% acceptable)4.EK F* (0.50 each) a.
b.
4 47- (1-2% acceptable) /. J fe c.
er'M' (9. T-C "*' acceptabl e)#.Yf fo . _ 4. t *fe REFERENCE CA dans./or 7Itae ~ ~ ~ ~~ ~ ~ ~ ~~~~~~~ fE L M~~f f ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO5000A1.06/IF 2.7 <
- >5 1, __ PRINGlPLES: OE_NUGLEAR_PQWER_ELANT OPERAllQNs PAGE ~:20 IMEEM9DYNOMIGSz_ME61_IB6NSEER_AND_FLUIQ_E6QW ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
ANSWER 1.07 (2.00) 1/M = CRO ___ CR1 Starting graph at 1/M = 1.00 (see Figure 1.13) (0.50) Correctly plotting given points for 1400 -> 1000 ppm (5 @ O.20 each) Correct answer 966 +or-33 ppm (0.50) REFERENCE Reactor Theory Ch 5 " Sources & Subcritical Multiplication", pg 17 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ KLA OO4000A4.02/IF 3.2 OO4000A4.03/IF 2.7 OO4010A3.04/IF-3.9 ANSWER 1.08 (2.00) a.
Pres = 50 psig (0.50 each) Flow = 77 gpm b.
Pres = 65 psig Flow = 96 gpm REFERENCE HTFF Fund Ch 12 " Fluid Flow Applications", pgs 33-36 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~,.~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ K&A Components: Pumps-Centrifugal /IF.. _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ __ _ _ _ _ _ _ _ _ _ _ .. 1.. PRINGIPLEg_gE_NgGLEAR POWER __ PLANT OPEBATigN PAGE' 21
-IHEBDggyN6d1GS2_ME91_IB9NSEEB_8Ng_ELylg_E6gW ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
l ANSWER 1.09 (3.00).
' 1.
Thermal power (0.50) Increasing power results in increased heat flux and thus closer to DNB.
(0.50) 2.- Pressurizer pressure (0.50) Inc.sasing pressure causes an increase in the subcooling and thus further from DNB.
(0.50) 3.
Highest Tavg (0.50) Increasing temperature decreases the subcooling and thus closer to DNB.
(0.50) REFERENCE ' MP3 TS, pgs 2-1 & B2-1 HTFF Fund Ch 10 " Reactor Protection", pgs 4-12 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ KLA OO2OOOA1.01/IF 3.8 OO2OOOA1.06/IF 4.0 OO2OOOA1.09/IF 3.7 ANSWER 1.10 (3.00) a.
Tcold will remain constant (0.25) Since it follows S/G saturation pressure.
(0.50) b.
That will decrease (0.25) Since less fission product heat is being produced (0.50) c.
Core deltaT will decrease (0.25) Since the amount of decay heat is decreasing (0.50) d.
Loop transit time will increase (0.25) Since the driving head for flow is decreasing (0.50) REFERENCE HTFF Fund Ch 13 " Natural Circulation", pgs 2-12 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ 'KLA OO3OOOK3.01/IF 3.7 _____ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
- .. - _ _ _ _ _ - _ _ _ - _ _ - _ - _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _. _ - - _. ... A__ELONI_DEGIGN_lNGGyDING_S8EETY AND_EMERGENGy_ SYSTEMS PAGE
, ANSWERS -- MILLSTONE 3-C6/12/15-NORRIS,-B.
S.
ANSWER 2.01 (3.00) (bI N each) Saf ety injection signal, deed a[ /PM/* b
a.
MDAFWP:- Low Low level 1/4 SG C,p /P j l TDAFWP: Low Low level 2/4 SG ~ b.
Demineralized water storage tank (0.40 each for sources, Condensate storage tank O.30 for correct order) Service water system REFERENCE MP3 NSSS Vol 3, Auxiliary Feedwater System, pg 17 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ KLA 061000K4.02/IF 4.5 061000K4.01/IF 3.9 ANSWER 2.02 (2.40) a.
MDAFW pump auto start signal (0.40 each) TDAFW pump steam supply valve open High radiation on SG BD High radiation on steam jet AE exhaust b.
Blowdown tank high pressure Main condenser'1ow vacuum REFERENCE l MP3 BOP Vol 2 Steam Generator Bl owdown, pgs 3-4 j.
.~~~~~~~~~~~~~~~~~~~~~~~~~~~ss~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.~~~~ l K&A 035010K4.03/IF 2.6 ,. - -. _ _ _.. _ _ _ _ _ - _ _ _.. - _. -.. _ _
- , . Sz__ELON1_QESIGN_INC6QQ1NG_S@EEIY_6NQ_EMEEGENCY_SYSIEMS PAGE
, ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
ANSWER 2.03 (3.00) a.
(The normal steam pressure setpoint of 1092 psig maintains Tavg at 557 F.)
A decrease in the setpoint to 1007 psig would cause the dumps to open and cool Tavg to '553 F (0.50) where the P-12 interlock would close all steam dumps (u.50).
b.
Steam dump operation would be blocked (0.30), secondary pressure woulo rise to the setpoint of the secondary pressure relief valve would maintain pressure at 1125 psig (0.40) causing primary temperature to steady out at 561F (0.30).
c.
In pressure control mode, the output of the turbine trip and load rejection controllers is blocked (0.30).
Temperature would be controlled at 561 (0.30) corresponding to the pressure relief valve setpoint (O.40).
REFERENCE Steam Tables MP3 NSSS Vol 3 Ste6m Dump System, pgs 19-23 ~~~~~~~~~~~~~~~~~~~~~,ss~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 041020K3.02/IF 3.8 ANSWER 2.04 (3.00) Ok, o h CMN h MJ/****V"! ' 4.
A or C RPCCW pump (0.50 each) ~~~~ /Y A Safety Injection pump
"* A ' Auxiliary Feedwater pump Jg gg -Ag, g/g.d sc/ A&C Service. Water pumps A&C Containment Recirc pumps A or C CVCS pump [ NOTE: RHR and Quench spray pumps will NOT be running] REFERENCE MP3 BOP Vol 1 " Electrical Systems Introduction" pgs 10 & 20-21 , ~~~~~~~~~~~~~~~~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO5000K4.OB/IF 3.1 OO5000K4.12/IF 3.1 OO6000K2.01/IF, .. ~3t__EL6NI_DEgl@U_INGLUDIN@_S6EETY AND_EMERGENGY_ SYSTEMS PAGE 24 i , ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
ANSWER 2.05 (2.80) , t i.
RHR heat exchanger: both trains, SR header (Or20-for correct header, RCPs: both trains, Containment header Oc45-for correct train) ' Excess LDHX: B train, Containment header ,.gc LDHX: A train, Containment header Baron Recovery system: both trains, NSR header Charging pump cooling surge tank: B train, Containment header SI pump cooling surge tank: A train, SR header Fuel pool cooler: Both trains, SR header REFERENCE MP3 NSSS Vol 3 "RPCCW" pgs 2-3 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO8000K1.02/IF 3.3 ANSWER 2.06 (3.25) a.
RWST==> Safety Injection pump==> SI Accumulator Master Fill Isolation Valve (CV8888)==> SI Test Header Cntmt Inner Isolation Valve (MV8871)
===> Accumulator A Fill Valve (MV8878A) (0.25 each) b.
Pressure high: forces more tank water into the RCS before the end of blowdown and therefore the water is assumed lost.
Pressure low: the tank water is not injected fast enough Gas volume low: injection flow rate is low and sufficient injection may not occur Gas volume high: less available water volume.
(0.50 each) REFERENCE-OP-3310B, pg 4-5 MP3 NSSS Vol 2 "ECCS" pgs 47-48 &.52-53 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ .KLA OO6000K6.02/IF 3.4 OO6020A1.07/IF. 3t__PL@NI_ DESIGN _ LNG 69Q1NQ_ SAFETY ANQ_EMERGENgy_Sy@TEMS _PAGE
a ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B..S.
ANSWER 2.07 (2.80) a.
30-second time delay (0.50 each) Predischarge alarm Strobe lights at entrances and within zones Wintergreen scent b.
Locally (0.40) Lockout the system to prevenrl accidental discharge (0.40) REFERENCE MP3 BOP.Vol 6 " Fire Protection System" pgs 7-9 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~,,3,~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 086000K4.06/IF 3.0 ANSWER 2.08 (2.50) a.
Standby Instrument Air Compressor starts (90 psig) _ (0.50) Service Air Cross-connect valve (3IAS-ADV14) opens (85 psig) (0.50) b.
1.
Fail closed (0.30 each) ,2.
Fail open-3.
Fail closed 4.
Fail closed 5.
Fail closed REFERENCE MP3 BOP Vol 6 " Instrument Air" pgs 7-9 OP-3332B " Containment Instrument Air" pg 7 P& ids _S&W Dwgs 12179-OO3A-4, 130A-4, 123B-4, & 106A-5 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OOOO65EK3.08/IF 3.7 078000K3.02/IF 3.4 078000K1.03/IF 3.3 . - -.. . y - ~ - - - - -.. -, -. . .. _. - - - - - - - - - -, .
. . . . l3i__ELONI_QE@l@N INGLUDINQ_EAFETY_AND_EMER@ENGY_ SYSTEM @ PAGE
. . ANSWERS -- MILLSTONE 13- -86/12/15-NORRIS, B.
S.
ANSWER 2.09 (2.25) a.
No - (0.25) Seal.injectionflowwouldcontinuetotheRCPs(butthereturnflow would be to the PRT vice the CVCjk (0.75) b.
Yes (0.25) The RPCCW to the pump motors would be isolated (0.50) causing the bearings to heatup and seize.
(0.50) -REFERENCE MP3 Required Drawings Nos. 304 & 311 MP3 NSSS Vol 1, " Reactor Coolant Pumps" Fig RCP-15 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO3OOOK1.08/IF 2.7 OO3OOOK6.04/IF 2.8 103OOOA2.'03/IF 3.5 i ? e
. 3c__1NSIBydENIS_6ND_CQUIBQLS PAGE
. ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
[a,7f) ANSWER 3.01 '2.25? a.
Om m.
(0.25 for direction, 0.50 for justification) _ Ortreter cra !ccc I r r':2;r ecultin; ' r-decrc rrd * c; b.
Increase f Due to the shifting of the neutron flux toward the perimeter of the core c.
??.;ht i n c r :2 c t-( Nc t; t!.c d cre __. Sc e-2 b : r p t i r -- 4- +"' d~" r-r- ~' f ^ ^ - REFERENCE C*A*F ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K.tA 015000A1.06/IF 2.5 015000A1.07/IF 3.3 015000A1.09/IF 3.3 W YYY & W W e*d25 } W. 0W W " N^f0) W n, im.a w --y.a z& & c n-q A.
A_ l I l l [ l l i
_ _ _ _.. _. _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ __ , _ _ _ , _ _ _ _ _, - _ _ _ _ _,
_.
.. g.
INSTRUMENTE_AND_GQNTROLS PAGE
. ANSWERS -- MILLSTONE 3-86/12/15-NORRIS,-B.
S.
ANSWER-3.02 (3.00) a.
Compared +d th Tref to develop rod speed and direction (0.50) R 72-l +++ (0.25) O l + D ! + + S 8-
<+++++++ P l v^ E _O--!
- ---l--- -+<+++++++++>+-:---l--- ---:--
E v^ D -8- +++++++> ! + l + l + l-72+++ l l ~
! ! ! ! !
-5- -4-3 -2-1
1
3
5 Tref - Tavg Auctioneered high Tavg (0.25) b.
Tavg is used as the reference for pressurizer level (0.50) ' Tavg 557F - 587F --> Level 25% - 61.5% (0.25) Auctioneered high Tavg (0.25) c.
&am Low Tavg coincident with Reactor trip causes feedwater isolation (0.50) Tavg = 564F (0.25) 2/4-logic (0.25) REFERENCE MP3 NSSS Vol 3 Temperature Indicating, pgs 13-16 Vol 4 Rod Control,pg 23 Pzr Pres & Level, pgs 23-24 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO1000A1.01/IF 3.8 011000A1.04/IF 3.1 059000A3.06/IF L.
- 3 __IN@IBUMENI@_BND_CONIBOL@
PAGE
t ANSWERS -- MILLSTONE 3' -86/12/15-NORRIS, B.
S.
ANSWER 3.03 (3.00) As power increases, steam flow detector delta-P increases (0.30) As power increases, steam pressure decreases (0.30) m-stm = K(P-stm) (sqroot (del ta-P) )
==>since the steam pressure component stays constant (it should go down) while the delta-P increases, indicated steam flow will be higher than actual steam flow (0.60) The summing network for flow will send a signal to the total controller to open the feed regulating valve.
(0.60) As level starts to increase, the level error will signal for the FRV to close 'O.60) Eventually, the flow error will be cancelled out by the level error, and the FRV will be positioned such that steam flow equals feed flow at some higher level.
(0.60) REFERENCE MP3 NSSS Vol 4 "Steae Generator Water Level Cont rol " 11-14 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~pgs ~~~~~~~~~~~~~~~~ K&A 035010K4.01/IF 3.6 '035010A1.02/IF 3.5 035010A2.03/IF 3.4 ANSWER 3.04 (1.50) Air ejector discharge (0.50 each) Steam generator blowdown Main steam line REFERENCE AOP-3556 " Steam Generator Tube Leak" pg 2 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 073OOOK1.01/IF 3.6 OOOO37EA2.01/IF 3.0 .
- . st__INEIRUMENI@_@ND_CQNIRQLS PAGE
, . ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
g,,- f*p [ ANSWER 3.05 (3.00) a.
Rod motion is prevented 4, p [[evf, C-/ (0.504 7;pa.c th; P ch:rnc! cr block C1 signal 25 each) m.
b.
Rod motion is prevented (0.50& ": ~ r m. ;- th; ^^ chcrnc! cr trip the PR bistables (0.20 each) t,/ fe l.
... s,
c.
No effect (0.50) d.
No effect (0.50) REFERENCE MP3 NSSS Vol 4 " Rod Control" pgs 62-64 " Rod Position Indicating" pg 21 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ KLA 012OOOK6.10/IF 3.3 ANSWCR 3.06 (3.00) a.
Train A inputs (PCV 455A) are auctioneered low Th (WR! (0.30) and wide range pressure (PT 405) (0.30) Therefore, Train A PORV will not open (0.40) Train B inputs (PCV 456) are auctioneered low Tc (WR) (0.30) and wide range pressure (PT 403) (0.30) Since the CDP system is armed and the Train B pressure setpoint is low due to the failed Tc instrument, PORV 456 will open.
(0.40) b.
ALARM warning when temperature instrument fails (0.50) ALARM when within 30 psi of setpoint (0.50) REFERENCE MP3 NSSS Vol 4 "Pzr Pres & Level Control" pgs 13-15 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 010000K4.03/IF 3.8 010000A2.03/IF 4.1 . L. _. . _. ..
3.
_INSIBQMENIQ_AND_CQNIBQL@ ~PAGE
ANSWERS.-- MILLSTONE 3-86/12/15-NORRIS, B..S.
- -ANSWER 3.07 (3.2O)- see attached drawing.
REFERENCE Millstone 3LRequired-Drawings #322 and 323 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 062OOOK4.~10/IF 3.1 062OOOK1.03/IF 3.5 '063OOOK1.02/IF 2.7 ' ANSWER 3.08 (4.00) . .a.- -BASIS: uncontrolled RCCA' bank withdrawal from (0.40 each) a subtritical condition-BLOCK: manual when both IR > 10E-10 amps b.
' BASIS: reactivity excursions from low, power which are too rapid for temperature and pressure protective actions BLOCK: manual when 2/4 PR > 10% J cc. lBASISt mul tiple rad drop d.
BASIS: back up to High Pressure tripgf [sesser/ W9 b "# - -BLOCK: automatic when 3/4 PR < 10% & 2/2 Pimp < 10% e.
. BASIS:.. LOCA. or: MSLB inside containment , f.
BASIS: loss of heat sink BLOCK: automatic if both' hot leg and cold leg isol ation valves are shut lin that loop REFERENCE.
MP3 NSSS Vol 5 "RPSAS" pgs 39-56 MP3 TS pgs 2-5, 2-6, & 3/4 3-26 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~- K&A 012OOOK4.02/IF 3.9 012OOOK4.06/IF 3.2 012OOOK6.03/IF 3.1 . 013OOOK1.01/IF 4.2 n ,. - -.- ,--., , -, - . -.. - ,-,-m,- - - -, -, - - ,, - -.., - - - -, -,,,, -. - - ~ - - -, -, -, -,., - - - - - -,.- - - - - - - -
. - . 3 __INSIRQMENIS;AND_GQNIRQLS .PAGE
t ... - . ANSWERS.-- MILLSTONE 3-86/12/15-NDRRIS, B.
S.
': ANSWER 3.09 (2.05) One PORV (PCV 455A) will open due to the actuation signal of all four pzr_ pres channels (> 2200 psia) (0.30) When. actual plant pres decreases towards setpoint (2200 psia), the functioning pzr press channels will cause the PORV to close (0.30); Both spray valves will modulate full open (0.30) Reactor trip on low pressurizer pressure (0.30) Pressure will continue to decrease until safety injection initiates (0.30) As SI continues, the pzr will eventually go solid and cause pressure to raise to the PORV setpoint (0.30) Pressure steady stcte pressure would be slightly > 2200 psia (0.25) REFERENCE MP3 NSSS Vol 5 " Instrument Failure Analysis" pgs 11-13 ~~~~~~~~~~~~~~~~~~~~~~~~~..~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OOOO27EA2.03/IF 3.3 OOOO27EK2.02/IF 2.4 010000K3.03/IF 4.0 s
4 y , ,_ ,. - -- - - .-m- .- --, -,, - - _ _--._ - - ., 4 __PBQCEQUEES_ _NQ8d@6t_6ENQ8d66t_EMEBGENCY ANQ-PAGE
- B6 Dig 6QGIC66_CQNIBQL ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
, ANSWER 4.01 (2.50) a.
5(N-18) = 50 REM gg s- - (O. 30). Total lifetime to date = + 1.5 = 45WF Rem Total lifetime available = 50 - 49 5 = p.5 Rem (0.30) Total this quarter available = 3 - 1.5 = 1.5 Rem (0.30) O_- _:-1-lifetime limit J.<5 H ' is more restrictive than O.85. Rem /Hr gamma + (.03 Rad /Hr)(10 QF) neutron =1.15 Rem /Hr dose rate - #. J J6 (0.30) 0.5 Rem /1.15 Rem /Hr = ._ : Hrs = 79 Min (0.30) EO.2 f or using the conservat' quality factor] a b.
25 Rem whole body one time exposure (0.50) c.
Director of Site Emergency Operations (0.50) REFERENCE' 10 CFR 20 SHP-4902, pgs 13-14 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A Plant Wide Generic #15/IF 3.4 ANSWER 4.02 (3.00) a.
Ch _M d ( m d.a f p p,- (0.50 each) b.
Continue at power c.- Reduce power (below 37%) d.
Continue at power if within capacity of both charging pumps e.
Trip f.
Shutdown REFERENCE ADP-3562 " Loss of Instrument Air" pg 3 AOP-3553 "High RCS Activity" pg 3 .ADP-3554 "RCP Trip or Seal Failure" pg 3 AOP-3555 " Reactor Coolant Leak" pg 3 ADP-3559 " Loss of Condenser Vacuum" pg 3 AOF-3565 " Loss of Containment Vacuum /Ir tegrity" p ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~g 3 ~~~~~~~~~~~~~~~~~~~ K&A System Wide Generic #10/IF 4.1 ,
. -. ,
4 __P8QQEQQ6ES _NQRMAbt_ABNQRMA6t_EMERQENQY_ANQ PAGE
t BeQ1060GIGe6_GQNIBQL-ANSWERS'-- MILLSTONE 3-86/12/15-NORRIS, B.
S.
. ANSWER 4.03-(1.50) -All FRGs are written on the premise that at least one AC emergency bus is energized.
REFERENCE WOG ERGS, Background Document.for ECA O.0, pg 75 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.,~~~~~~~~~~~~~~~~ K&A System Wide Generic #7/IF 3.5 ANSWER 4.04 (2.40) The RCS would heat up (0.60) a.- due to the RHR system being unable to remove all decay heat (0.60) b.
Relief. valve would' lift (0.60) to protect the RHR piping (0.60) REFERENCE OP-3201 Plant Heatup, pg 13 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO5000K4.07/IF 3.2 ANSWER 4.05 (2.25) a.- 551F (NO tolerance allowed) (0.75) b.
With Tavg < 551F restore to within limits withir. 15 minutes (0.75) or be in Hot Standby within the next 15 minutes.
(0.75) REFERENCE .TS, pg-B2/4 1-2 OP-3202, pg 6 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A-OO1000K5.15/IF 3.4 OO1000K5.16/IF 3.4 OO2OOOK5.18/IF 3.3 OO2020SG#5/IF 2.9 OO2020SG#8/IF 3.5 l-l I
-- . , e; [PRQQEDURES - NORMAlt_ABNQRMGLt_ EMERGENCY ANQ PAGE
'4.' !*- .BeQ106061G66_GQN18QL ANSWERS -. MILLSTONE 3-86/12/15-NORRIS, B.
S.
. ANSWER 4.06 (3.00) > Ja. FAULTED - a. break in -the secondary pressure boundary (0.75) RUPTURED - a break in the primary pressure boundary (specifically the steam generator tubes) (0.75) 'b.
Containment temperature >_180 F (0.50 for each parameter, Containment radiation levels >'10E5 R/Hr 0.50 for each value) REFERENCE WOG-Executive Volume, Writer's Guide section, pgs 48 & 61 MP3 EOP Background Document ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A-035010A2.01/IF'4.G System Wide Generic #10/IF 4.1 ANSWER 4.07 (3.50) a.
Rod height below the Low-Low limit (4 required at 0.50 each) Failure of one or more rod clusters to fully insert on a reactor trip or shutdown Uncont- '. led cooldown following a reactor trip or shutdown Unconteulled or unexplained reactivity increase . Failure of the Makeup system to borate ~ b.
10 pcm = 10 gal==> must add 100 gallons boric acid (0.50) c.
Preferred - Boric acid tanks to BAT pumps to Immediate boration valve (MV 8104) to suction of charging pumps (0.50) Alternate - Boric acid tanks to Gravity baration valves (MV B507 ALB) ' to suction of charging pumps' (0.50) REFERENCE ADP-3566 "Immediate Boration" pgs 2-3 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OOOO24EK1.02/IF-3.6 OOOO24EK3.01/IF 4.1 OOOO24EA2.05/IF 3.3 OOOO24EK3.02/IF 4.2 .
. ..: ds__EBOGEEMEES_ _NQBd66t__ABNORMAkt_EMERGENQY_ANQ PAGE
- -
E601069 GIG 86_QQNIBQL ANSWERS -- MILLSTONE 3-86/12/15-NORRIS, B.
S.
ANSWER 4.08 (.75) If the action contained in the RNO column cannot be performed, then continue on with the next step in the left hand column.
(0.75) Acj * 191f3 f*f T/am'eseer Vffjf ANSWER 4.09 (2.50) a.
When directed to transition from EOP 35 E-O W G each) , or when ins (ructed to implement the CSFSTs in EOP 35 E-0.
09-Aen % Kri g F CA~e.O b.
2.
Core Cooling - Red (0.25 each, proper 1.
Containment - Red sequence required) 6.
Subcriticality - Orange 4.
Integrity - Orange 3.
Heat Sink - Yellow 5.
Inventory - Green REFERENCE EOP Training Text, pgs.11-16 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OOOO74SG# 10/IF 4.7 ANSWER 4.10 (1.60) The item is NOT operable.
(0.40) Surveillance intervals for a monthly are not to exceed 7.75 days (25% of 31 days) (0.40) The surveillance is nine days overdue (0.40) Failure to meet surveillance requirements constitutes a failure to meet operability requirements.
(0.40) REFERENCE TS, pg 3/4 0-1 L 3/4 0-2 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~o~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A System Wide Generic #5/IF 2.9 System Wide Generic #7/IF. .
- 4-E8QGEQU8ESf NORMAL _ABNQBMAkt_EMER@ENQY_ANQ PAGE 3 71 t
" 869196991G86_ggNIBQL ANSWERS--- MILLSTONE 3-86/12/15-NORRIS, B.
S.
,
ANSWER 4.11 (2.00) a.
- The onshif t SROs (0.757 b..
The valve is to be operated only by order of the individual to whom the tag is issued.
(0.75) c.
No other tag may be attached.
.(0.50) REFERENCE MP3 ACP-DA-2.06A " Station Tagging" pgs 4 & 15 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A Plant Wide Generic #14/IF 3.6 . I
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hWMf"7 K .* , U. S. NUCLEAR REGULATORY COMMISSION - , SENIOR REACTOR OPERATOR LICENSE EXAMINATION
FACILITY: MILLSTONE 3 REACTOR TYPE: PWR-WEC4 DATE ADMINSTERED: 88/12/15 R C0 EXAMINER: JENSEN.
N.
CANDIDATE INSTRUCTIONS TO CANDIDATE: Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY &V. 75/f _.2&-00 25.00 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND THERMODYNAMICS 25.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 25.00 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25.00 25.00 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.0 Totals All work done on this examination is my own.
I have neither given nor received aid.
. Candidate's Signature , g -, - - -. -,-- , -.., -. --. w - - _ _, _. _ - . -
e , NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS , During the administration of this examination the following rules apply: 1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each cection of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Category __" as EpPropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been completed.
.
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18. When you complete your examination, you shall: i .o.
Assemble your examination as follows: (1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
c.
Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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THEORY OF NUCLEAR POWER PLANT OPERATION.
Paga
FLUIDS.AND THERMODYNAMICS - l L QUESTION 5.01 (3.00) The reactor is at 15% power, equilibrium Xenon and Samarium, boron concentration is 1200 ppm, BOL, Control Bank C rods are at 200 steps.
Using Figures 5.1 to 5.11: < a.
What will be the final boron concentration if power is raised to 100%, ARO, steady state conditions? State which figures you use.
State all assumptions and show all work.
(2.40) b.
How many gallons of pure water or boric acid will have to be added? (0.60) b l.ON eY g pekd. h%c ey
QUESTION 5.02 (2.25) lu ( jgor , Figure 5.12 is a sketch, power v t e; prior to time "0", the reactor had been shutdown for 0 days.
Using Figures 5.1 through 5.11 as a reference, ! sketch he aproximate xenon concentration associated with this power history.
U the same form provided.
. (2.95), A $ QUESTION 5.03 42.40t A reactor at BOL is critical at 10E(-9) amps.
Rods are withdrawn at 48 steps per minute for 10 seconds in bank sequence.
a.
Calcu hte the SUR immediately BEFORE rod motion stops and immediately AFTER rod motion stops.
Assume differential rod worth is 10 pcm/ step, effective delayed neutron fraction is 0.006 and the effective precursor decay constant is 0.1 sec-1.
.(.1.4&T (2.lo) b.
How and why would the SUR change, if at all, if the initial conditions were at EOL? No calculations are necessary.
.00,60') p.n) l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) . .. _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _.
f e.g C.2.35 . 0UESTION 6 - , Tha r6cetor in in Mod 2 2, all chutdown bani's are full out, end the control banks are withdrawn to 180 steps on Bank C.
Ucing the belcw i nf orma ti on - and Figure J~12, determine the critical boron concentration by a 1/t1 pl ot. " Show all work: Baron Concentration Count Rate 1500 ppm 100 cps 1400 125 1300 165 1200 240 1100 445 1000 3330 e
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. . -5.
THEORY OF NUCLEAR POWER PLANT OPERATION.
Paga
FLUIDS.AND THERMODYNAMICS - ' QUESTION 5.04 (2.25) If reactor power is increased from 50% to 100%, compare how DIFFERENTIAL rod worth changes from the 50% value (increase, decrease, or remain the
same) for the following conditions? Justify your answer.
Consider each case separately.
a.
Rod position and boron concentration are held constant, temperature is allowed to decrease.
b.
Boron is held constant, rods are withdrawn from 150 steps to maintain temperature constant.
- .
c.
Rod position is constant, boron concentration is diluted to maintain ' . temperature constant.
-QUESTION 5.05 (2.00) List TWO factors which affect the Doppler Fuel Temperature Coefficent (FTC) as the core ages AND briefly explain how each factor affects the value of FTC.
(2.0) , QUESTION 5.06 (3.00) The Unit 3 calculated Shutdown Margin is 10% delta K/K assuming the most reactive control rod is 1000 pcm.
The Source Range count rate is 50 cys.
Show all work and state any assumptions.
Consider each situation l separately.
! What would be the final count rate if the shutdown banks are fully i - a.
l withdrawn? Assume the shutdown bank rod worth is 5600 pcm.
b.
What would be the final count rate if a 100 ppm dilution of the RCS were performed? . i QUESTION 5.07 (1.50) A cooling water pump is operating at 1500 rpm.
Its capacity is 250 gal / min at a discharge pressure of 15 psig which requires 40 KW of power.
Determine the pump capacity, speed and power requirement if the pump discharge pressure drops to 10.0 psig due to reduced speed.
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THEORY OF NUCMAR POWER PLANT OPERATION.
Pag 2
FLUIDS.AND THERMODYNAMICS . QUESTION 5.08 (2.50) a.
What is the subcooling margin of the plant if the following conditions exist: (1.00) Thot = 587 F Tavs = 572 F Teold = 557 F Ppzr = 2235 psig Psg = 1033 psig b.
If power is raised from 50% to 100%, how will the subcooling margin change (increase, decrease, or remain the same)? Justify.
(0.75) c.
Which one of the following conditions would result in the smallest subcooling margin? Assume RCS pressure is the same for all cases.
Justify.
(0.75) 1.
Controlled natural circulation cooldown immediately following a reactor trip from loss of flow.
2.
Continued operation at 5% power with all RCP's running.
3.
All RCPs are operating at normal no-load temperature after an extended shutdown.
QUESTION 5.09 (3.00) There are THREE primary parameters that affect DNB and can be controlled by the Reactor Operator.
Assuming fcur-loop operation, briefly explain how an increase in each of the parameters affects DNBR if the other two parameters remain constant.
QUESTION 5.10 (3.00) The plant is being cooled by Natural Circulation.
As the fission product heat decays away, describe HOW and WHY the below parameters will change.
' Assume that S/G pressure is being maintained at 1000 psia.
, a.
Tcold b.
Thot c.
Core delta T d.
Loop transit time (***** END OF CATEGORY 5 *****) - - -. - - -
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PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paga
. QUESTION 6.01 (2.50) While operating at 80% power, one safety valve on #2 S/G fails open.
a.
How will the nuclear and non-nuclear instrumentation alert the operator that a problem has occurred.
State TWO ways.
(1.0) b.
What three signals cause automatic closure of the HSIV's? Indicate coincidence; additionally, if.any of the signals can be blocked, identify HOW and WHEN the block is accomplished.
(1.50) QUESTION 6.02 (2.25) Would the below events cause the INDICATED power to increase, decrease, or remain the same assuming that the actual power remains unchanged at 75% before and after the event.
Assume all control systems are in automatic.
Justify your answer.
Consider each event separately, a.
Cooldown of the reactor coolant b.
Core aging c.
25 ppm decrease in boron concentration QUESTION 6.03 (3.00) oha For each of the following control systems, explain MOW"the Tavg input signal is,deri d and WHAT control functions the Tava signal provides.
Include set points as appropriate.
a.
Rod control b.
Pressurizer level control c.
Feedwater control (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pasa
. QUESTION 6.04 (2.40) a.
What FOUR conditions will automatically close the steam generator blowdown containment isolation valves (CTV 22s)? (1.60) b.
What TWO conditions will automatically close the blowdown flow control valves (HV 20s)? (0.80) QUESTION 6.05 (2.00) a.
What TWO conditions must be satisfied to open the CVCS Orif ce Isolation Valves? (1.0) WHatTWOconditionswillcause..heCVCSOrifdceIsolationValves b.
to automatically shut other than the interlocks stated above? (1.0) QUESTION 6.06 (3.00) a.
Assume the plant is shutdown at 500 F and 2000 psig and the Cold Overpressure Protection (COP) System is inadvertently armed.
Will the COP system act to reduce plant pressure if a loop Tc instrument fails low? Justify your answer, include how each Train is affected, and what action takes place.
(2.00) b.
What warnings would the Reactor Operator have of the condition described in part (a) above? (1.00) QUESTION 6.07 (2.00) a.
State TWO Diesel Generator conditions which will always automatically stop the diesel engine? (1.0) b.
What relay actuation will cause a Start Failure (SFR)? (0.5) c.
What action is required (as a minimum) before another starting.
attempt can be made? (0.5) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pc2
QUESTION 6.08-(1.50) What precautions are taken to warn personnel prior to or after the release of CO2 discharge? (1.5) QUESTION 6.09 (1.35) a.
State TWO actions / signals that would occur as a result of both Reactor Bypass Breakers being closed at the same time.
(1.0) b.
. List the location (s) where the Reactor Trip Bypass Breakers (BYA and BYB) may be opened AND closed.
(0.35) QUESTION 6.10 (3.00) ' ' a.
State TWO of THREE interlocks associated with the RHR suction valves (8701 A/B or 8702 A/B).
Include any applicable setpoints. (1.5) , b.
How is RCS pressure controlled during cold solid conditions with cleanup via CVCS domineralizers in progress? (0,5) c.
State TWO conditions that would AUTOMATICALLY trip an RHR pump, excluding loss of power.
(1.0) QUESTION 6.11 (2.00) What are FOUR of the FIVE major sources of Hydrogen generation that the Hydrogen Recombiner is designed to control.
(2.0)
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PROCEDURRR - NORMAL. ABNORMAL. EMERGENCY Paga 10 ' - AND RADIOLOGICAL CONTROL QUESTION 7.01 (2.25) You are the on-duty SCO and a Radiation Work Permit (RWP) is brought to you for approval.
In accordance with SHP 4912 Radiation Work Permit Completion and Flow Control, your approval signature signifies what THREE conditions with respect to the plant? QUESTION 7.02 (2.50) The following concern " Refueling Operations" procedure OP 3210B.
a.
Briefly explain the allowable exception to the requirement for maintaining residual heat removal flow during core alterations.
Include any applicable time limitations.
(1.0) b.
What is the basis / reason for maintaining 10.5 feet of water above the top of each fuel assembly during all handling operations.
(0.5) c.
Who, by job position / title can approve / authorize loads greater than that of a fuel assembly and RCC to travel over irradiated fuel in the vessel with the head removed? (0.5) d.
What count rate indication requires suspension of fuel handling operations? (0.5) QUESTION 7.03 (1.00) Answer the following with respect to OP 3330A " Reactor Plant Component Cooling Water": , a.
Why should the shift from A or B train RPCCW to C Swing RPCCW pump and heat exchanger be done swiftly? b.
Why must you limit RPCCW flow to 8100 gpm? . I !' (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) . -. -. - . . -. . -. -. - -. - . -. - - - - -..-. .- -.-_ -.
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PROM >URES - NORMAL. ABNORMAL. EMERGENCY Paca 11 AND RADIOLOGICAL CONTROL ~- ' QUESTION 7.04 (2.50) The following concern " Reactor Coolant Pump Operation" procedure OP 3301D precautions, and a.
Why should the Reactor Coolant Pump (RCP) seal leakoff isolation valve be closed when RCS pressure is below 115 psia? (0.5) b.
How long can RCP operation continue if the seal leakoff isolation valve was shut due to #1 seal leakoff exceeding the limit for safe operation? (0.5) c.
If either Number 1 or 2 Reactor Coolant Pump (RCP) is stopped, what action is required AND why? (Assume both Number
3 and 4 RCP's remain in operation.)
(1.0) ' d.
Why are_RCP starting duty requirements necessary? (0.5) QUESTION 7.05 (3.50) , The following concern " Reactor Startup", procedure OP 3202.
a.
What TWO operator actions are required if criticality is NOT achieved when control rods reach the MAXIMUM rod position on the ECP? (1.0) , b.
What operator action is required when diluting the RCS boron concentration by more than 50 ppm? (0.5)
c.
Briefly explain why you ARE NOT permitted to adjust boron concentration while withdrawing the shutdown banks.
(0.5) d.
How is proper group alignment and bank overlap determined I during rod withdrawal for criticality? Include in your answer i WHAT is done and HOW OFTEN (or at what points) it is done.
(1.0)
e.
While recording data at 10E-8 amps, you determine that loop D Tavs l has been 535F for the last 15 minutes.
What action is required? I Include any applicable time limitation (s).
Assume no failed instruments.
(0.5) , (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) i l
. . 7.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pra 12 AND RADIOLOGICAL CONTROL . QUESTION 7.06 (1.50) In accordance with OP 3319A " Condensate System": When initially starting up the system, why should you ensure proper a.
motor amperage on the pump before letting go of the switch? b.
Prior to placing the system in operation, with steam generator pressure less than 700 psig, why should you ensure that the feed-water isolation valves are shut? QUESTION 7.07 (2.50) A maintenance man is working inside the containment while the reactor is at power.
He is working in a radiation field of 850 mrem /hr gamma and 30 mrad /hr thermal and fast neutron.
The man is 33 years old and has a lifetime exposure through last quarter of 72.5 Rem on his NRC Form 4.
Additionally, he has accumulated 1.5 Rem so far this quarter.
a.
How long can the man work in the area before he exceeds his 10CFR20 limits? Show all work and state all assumptions.
(1.50 b.
During a declared emergency, this individual volunteers to enter a high radiation area and perform work necessary to prevent further effluent release.
In accordance with the MP3 Procedures, what is his maximum allowed whole body exposure? (0.50 c.
In accordance with the Health Physics Procedures, whose authorization (by job title / position) is required in part b? (0.50 (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) . _ _. -
' . . 7.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Peca 13 AND RADIOLOGICAL CONTROL . QUESTION 7.08 (2.50) .For each of the following situations, determine whether the plant should be TRIPPED, SHUTDOWN, REDUCED IN POWER, or ALLOWED TO CONTINUE AT POWER.
Assume the plant is operating at 100% power and consider each case separately.
a.
Instrument air pressure falls to 90 psig, b.
High RCS activity is indicated by chemistry sample and a failed fuel monitor alarm, c.
Number 1 seal on a reactor coolant pump has failed.
d.
Condenser vacuum decreases to 8.0 inches HgA.
e.
The containment pressure has been above 13.0 psia for over one hour.
QUESTION 7.09 (1.00) According to the foldout page (Reactor Trip /RCP Trip Criteria) of AOP 3561 " Loss of Reactor Plant Component Cooling Water", state TWO of the five listed conditions which require you to stop the RCP's.
' QUESTION 7.10 (2.00) a.
What constitutes Adverse Containment? (Include specific parameters and values.)
b.
Under Adverse Containment conditions, some parameter values speci-fled in the EOP's are different than values listed for the same parameters under normal containment conditions - (for example: SG Narrow Range Level, RCS Pressure).
EXPLAIN why this is necessary.
. l QUESTION 7.11 (0.75) What are the operator's actions if, AFTER referring to a " Response Not Obtained (RNO)" action step, the expected RNO response cannot be satisfied? l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) .. _ _ -._. _ - - - - -. - - - _ . - - - . - ,___ -__-_ --_ ______ _
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PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pag 3 14 AND RADIOLOGICAL CONTROL - QUESTION 7.12 (2.50) a.
In accordance with the Rules of Usage of the Emergency Response Procedures, WHEN are the operators required to start monitoring the Critical Safety Function Status Trees (CSFSTs)? State BOTH criteria.
(1.00) b.
Prioritize the below listed Critical Safety Function Status Trees from highest to lowest: (1.50) , 1.
Containment - Red 2.
Core Cooling - Red 3.
Heat Sink - Yellow 4.
Integrity - Orange 5.
Inventory - Green 6.
Subcriticality - Orange QUESTION 7.13 (0.50) Concerning EOP 35 ES-0.3, Natural Circulation Cooldown With Steam Void In Vessel (With RVLMS), why is the operator cautioned NOT to depressurize the RCS until the entire system is below 200F7
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' . . 8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Peca 15 AND LIMITATIONS - QUESTION 8.01 (1.50) In accordance with the Station Tagging procedure ACP-QA-2.06A, what THREE conditions must be met in order to lift safety tags from equip-ment, thus allowing temporary operation, such as a retest or a functional check? QUESTION 8.02 (1.50) According to ACP-QA-2.02C Work Orders, which of the following ARE primarily the responsibility of the Shift Supervisor / Supervising Control Operator (SS/SCO)? (Assume work is being completed by an outside contractor.)
a.
Approve or ammend retest recommendations.
b.
Identify any LCO requirements involved on the Work Order.
c.
Prepare Supplemental Work Orders, if required.
d.
Assure a fire watch is provided, if necessary.
e.
Authorize work to commence.
f.
Obtain RWP, if necessary.
g.
Specify and place appropriate safety tags.
h.
Coordinate the performance of the retest / functional verification.
.
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) . , _ - - _ _ _ _ ~ _.___ __ . . .
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ADMINISTRATIVE PROCEDURES. CONDITIONS.
Paga 16 AND LIMITATIONS - QUESTION 8.03 (3.00) The following concern " Removing Equipment from Service for Maintenance", procedure OP 3250, a.
When is a system considered high pressure, high temperature? (0.8) b.
What is the required order of isolation of a high pressure system pump suction and discharge valves? (0.7) c.
Can a discharge check valve act as a pressure barrier? (0.5) d.
Maintenance is to be performed on a hydrogen system.
The affected portion has been purged with six system volumes, isolated, vented and tagged.
What additional action is required before allowing the work to be performed? (1.0) l QUESTION 8.04 (2.00) The following concern " Station Bypass / Jumper Control" procedure, ACP-QA-2.06B.
a.
If a Technical Specification change is required, or an unreviewed safety question is found to exist, then approval is required from (1) AND (2) before bypass / jumper installation can be authorized.
b.
What condition allows the Shift Supervisor to grant exception to performing a second verification of a jumper installation? Which TWO individuals, by job position / title, must complete and c.
sign the Assessment Section of the jumper-lifted lead-bypass control sheet? (Do not include the PORC review / approval signature.)
d.
Under what conditions can a jumper be installed on operable equipment WITHOUT using procedure ACP-QA-2.06B7 . (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
i . . . 8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Paga 17 , AND LIMITATIONS - QUESTION 8.05 (1.50) In accordance with ACP-QA-2.12 System Valve Alignment Control, state TWO of the four criteria which require independent verification checks of valve alignments.
QUESTION 8.06 (1.50) The CODE OF FEDERAL REGULATIONS (10 CFR 55) defines the general provisions Operator's Licenses. In accordance with these regulations: a.
What are the " controls" of a nuclear facility ? b.
Who may operate the " controls" of a nuclear facility without an Operator's or Senior Operator's License ? QUESTION 8.07 (2.00) a.
What is the basis for the Technical Specification limit on Reactor Coolant System activity? Include any specific postulated accident (s).
b.
What is the basis for the Technical Specification limit on Steam Generator activity? Include any specific postulated accident (s).
QUESTION 8.08 (2.50) During performance of routine surveillance, it is determined that the reactor coolant system total leakage rate is 2.2 gallons per minute, distributed as follows: 0.43 gpm through S/G "A" tubes.
0.18 gpm through S/G "C" tubes.
0.31 gpm through S/G "D" tubes.
0.15 spm through Loop "A" RTD manifold hot les isolation valve packing The remainder of the leakage is unidentified.
Describe any technical specification limiting condition (s) for operation which is/are exceeded by these leak rates.
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) _
l . . . 8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
P G 18 AND LIMITATIONS - , QUESTION 8.09 (3.00) The plant is operating at 300% power, all control systems are in automatic, and all equipment is operable except as stated below.
Using Section 3/4.0 the Millstone 3 Technical Specifications provided, state what Limiting Condition for Operations (LCO) is violated.
(NOTE: Reference to the page number & paragraph number is sufficient; if there is more than one action statement for a system, be specific as to which paragraph).
If the situation is not a violation of an LCO, justify by reference to page number and paragraph number.
Consider each situation separately: a.
The TDAFW pump (3FWA-P2) failed its latest surveillance.
b.
The "B" Train SI pump (3SIH-P1B) circuit breaker is racked out for replacement.
c.
Charging Pump (3CHS-P3C) motor is being replaced.
d.
Quench Spray Pump (3QSS-P3A) has seized bearings.
e.
Containment average air temperature is 76F.
QUESTION 8.10 (2.00) The plant is in Mode 3 with a routine startup in progress.
During surveillance testing on one of the Diesel Generators it is determined that its day tank level switch is out of commission and will not control its fuel oil transfer pumps.
The Maintenance Supervisor assures you that repairs can be completed within 48 hours.
Explain why the startup CAN or CANNOT proceed.
QUESTION 8.11 (3.00) a.
In accordance with EPIP 4010A "Shif t Supervisor'.', when must the SS assume the responsibilities of the Acting Director of Station Emergency Operations? (1.0) b.
In accordance with EPIP 4010B " Acting Director of Station Emergency Operations" what four conditions must you consider when determining whether or not to evacuate? (2.0) (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
- . . 8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Pago 19 AND LIMITATIONS
QUESTION 8.12 (1.50) Which individuals shall be allowed access to the Control Room during emergencies, according to ACP 6.01 Control Room Procedure? (List THREE categories of individuals.)
. (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********)
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' . . 5.
THEORY OF NUCr.rAR POWER PLANT OPltRATION.
Page 20 FLUIDS.AND THERMODYNAMICS - ANSWER 5.01 (3.00) a.
Reactivity due to power increase [ Fig 5.2] 200 - 1350 = -1150 pcm (0.40) Reactivity due to rod withdrawal (Fig 5.1] 425 - 0 = +425 pcm (0.40) Reactivity due to Equilibrium Xenon Change (Fiss 5.3 & 5.4] 1320 - 2760 = -1440 pcm (0.40) Boron worth (Fig 5.8) =~ -9.2 pcm/ ppm (0.40) No reactivity changes due to Samarium (0.40) Total: -1150pcm + 425pcm -1440 pcm = -9.2 pcm/ ppm (1200 ppm - BC) -2165 pcm = -0.2 pcm/ ppm (1200 ppm - BC) BC = (11040 pcm - 2165 pcm)/(0.2 pcm/ ppm) BC = 964.67 ppm (0.40) b.
Dilute ~12,000 gallons (Fig 5.11] (0.60) REFERENCE Reactor Theory, Ch 7 " Controlled Reactiv3ty Effects" Ch 8 " Fission Product Peisoning Effects", pg 27 UES~$?E!*~E??b~~b*$~*$$$PE? Ebb *~bb5?E*~$b$b*?!?b**$~"?'N*E*$$$$5??b~~~ K&A 001010K5.21/IF 3.4 001000K5.09/IF 3.5 ANSWER 5.02 (2.25) Sorm unadenllon vs.llM See attached Figure 5.12 (*****+;r: vide f4:er: titi t^t:Cr') REFERENCE , Reactor Theory, Ch 8 " Fission Product Poisoning Effects", pas 7-25 MP3 Curve Book (****** provide figure title *******) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 001000K5.13/IF 3.7 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
__ __ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ __ . . 5.
THEORY OF NUCr.r4R POWER PLANT OPERATION.
Page 21 FLUIDS.AND THERMODYNAMICS
u.as-), STEf ANSWER 5.03 L? 6Gt 0.30 a.
Lambda = 0.1 sec-1 & Beta eff = 0.006 (0.25) Rho = (10 sec)(1 min /60 sec)(10 pcm/ step)(48 steps / min) o, yo = 80 pcm (AM) Rho dot = (10 pcm/ step)(48 steps / min)(1 min /60 sec) ,, s o = 8 pcm/sec (ove6) Before rod motion stops: ,, 3, SUR = 26(lambda (rho) + rho dot)/(beta eff - rho) (4re5) = 26(0.1(8E-4) + 8E-5)/(6E-3 - 8E-4) = 26(1.6E-4)/(5.2E-3) o. J o = 0.80 dpm (Gres) After rod motions o. so SUR = 26(lambda (rho))/(beta eff - rho) (Aces) = 26(0.1(8E-4))/(5.2E-3) o. so = 0.40 dpm (Ares) o. s o b.
At EOL, beta effective is less (Aves) which gives a higher SUR (Gre5) o. a o REFERENCE Reac. tor Th.eory Ch 4 "R.eactor Kinetic.s"..,. pa..s 212-216, 235-244 & 263-268 i ' .. . .. .. ... .. ... . --. . .. . K&A 001000K5.03/IF 2.9 i ' 001000A1.06/IF 4.1 015000A1.02/IF 3.5 ANSWER 5.04 (2.25) a.
Decrease (0.25 for direction, 0.50 for justification) Lower temperature increases the density thus reducing the number of neutrons available for capture by the rods, b.
Decrease There is less flux available for interaction as the rods move towards the top of the core l c.
Increase Decrease boron concentration increases the number of neutrons available for interaction with the rods, i
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) - _. - _ - _. _ - - _ _. _ . _ -. - - - - - _ _ _ _-_-_ --- __ _ -_ - . -.
. . 5.
THIORY OF NUCf,RAR POWER PLANT OPERATION.
Page 22 FLUIDS.AND THERMODYNAMICS - REFERENCE b**f13 fib'3fEbb!.293335333'$b'*31b!b1E{{{eg}s]f,,,pgsgg,,pj,,,,,,,,,,,,,, ,, K&A 001010K5.04/IF 2.2 001000K5.02/IF 2.9 ANSWER 5.05 (2.00) 1.
BUILDUP OF PU240 [0.5] -- Causes an increase in the resonant absorption resulting in a MORE negative react.
Per degree temperature change [0.5]. 2.
CLAD CREEP / PELLET SUELL [0.5] -- Causes an increase in gap conductivity and with increasing power the pellet edge remains cooler at EOL resulting in a LESS negative FTC [0.5]. 3.
FISSION GAS BUILDUP [0.5]--Decrease gap conductivity making FTC more negative [0.5]. , _, ,, , ,......, 4.v ..mo u.., If anner // is i i is wor /] 2.c paials.
U "" **" O * given,fren> / bey are woei), l.0 foid mef. (dbeslio., ur}f a,aai,,.a, '# i REFERENCE Rin/2 sullrocle/ [or incorreef bsSormallon.
o f 2.0 poi,,0s.)
Millstone Re @ or Theory, Chapter 6, pp 317-321 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
. . 5.
THEORY OF NUCMAR POWER PLANT OPERATION.
Paga 23 FLUIDS.AND THERMODYNAMICS - ANSWER 5.06 (3.00) a.
CR(f) = CR(i) [(1-Ki)/(1-Kf)] (0.25) Reactivity in core (1) = -10,000 pcm + (-1000 pcm) = -11,000 pcm (0.25) (f) = -11,000 pcm + 5600 pcm = -5400 pcm (0.25) SDH = (1-Keff)/Keff OR p= (Keff - 1)/Keff ; Keff = 1/(1 p) (0.50) Ki = 1/(1 + 0.11)
==> Ki = 0.9009 (0.25) Kf = 1/(1 + 0.054)==> Kf = 0.9488 (0.25) CR(f) = 50 cps [(1 - 0.9009)/(1 - 0.9488)] = 96.8 cps (0.25) b.
Assume Boron worth = 10 pcm/ ppm (0.25) K(1) = 0.9009 CR(i) = 50 cps Reactivity in core (f) = -11,000 pcm + (100 ppm * 10 pcm/ ppm) = -10,000 pcm (0.25) K(f) = 1/(1 + 0.10) = 0.9091 (0.25) CR(f) = 50 cps [(1 - 0.9009)/(1 - 0.9091)] = 54.5 cps (0.25) REFERENCE 5 $ E. ~ $.?..$ ~ $l f E ~.!! W ~ '. .!. .'. E!. ~ " ~ $ '$."..' ~ ~ K&A 001000K5.09/IF 3.5 ANSWER 5.07 (1.50) N1=1500 rpm V1:250 gym Hp1:15psis Pl= 40 KW l N2 = 1500 sqr root (10.0/15) = 1224.7 rpm [0.5] V2 = (1224.7/1500)(250 rpm) = 204.1 gym (0.5]
3 P2 = (1224.7/1500) (40 KW) = 21.7 KW [0.5) REFERENCE Millstone 3 Thermodynamics text, Chapter 12, pp. 21-40
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) -- _ _ __ . - _. _ _ _ __ _... _ _ . _ _ _.., _ _ _ , _. _. _ _ _ _ _--.
- _ . . . 5.
THEORY OF NUC W AR POWER PLANT OPERATION.
P c3 24 FLUIDS.AND THERMODYNAMICS - ANSWER 5.08 (2.50) a.
Tsat for 2250 psia (2235 psig) (0.20) = 652.67 F (0.40) SCM = Tsat - Thot = 652.67 - 587 = 65.67 F (0.40) b.
SCM will decrease (0.25) Thot will increase as power increases (0.50) c.
1 - Natural Circulation Cooldown (0.25) Core delta-T during NC cooldown will approach full load delta-t (0.50) REFERENCE Steam Tables MP3 NSSS Vol 1, RCS, pg 33 NSSS Vol 1, Pzr & PRT, pg 16 HTFF Fund, Ch 3 " Thermodynamics", pgs 43-51 Ch 13 " Natural Circulation", pg 4 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ ' K&A 001000K5.56/IF 4.2 002000K5.09/IF 3.7 ANSWER 5.09 (3.00) 1.
Thermal power (0.50) Increasing power results in increased heat flux and thus closer to DNB.
DNBR decreases.
(0.50) 2.
Pressurizer pressure (0.50) Increasing pressure causes an increase in the subcooling and thus further from DNB.
DNBR increases.
(0.50) 3.
Highest Tava (0.50) Increasing temperature decreases the subcooling and thus closer i to DNB.
DNBR decreases.
(0.50) l [ (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) l { , .,- n.
,-w---..e---e--n--+, -, - - - - -, - - - - - - - -. ,-,e---m.,,, -nr-., , -, , - w - ------ -- r - - - -, - - - - - - - - -, - - -
- . . 5.
THEORY OF NUCWAR POWER PLANT OPERATION.
Pr o 25 FLUIDS.AND THERMODYNAMICS
REFERENCE MP3 TS, pas 2-1 & B2-1 HTFF Fund Ch 10 " Reactor Protection",~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ pas 4-12 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 002000A1.01/IF 3.8 002000A1.06/IF 4.0 002000A1.09/IF 3.7 ANSWER 5.10 (3.00) . a.
Teold will remain constant (0.25) Since it follows S/G saturation pressure.
(0.50) b.
Thot will decrease (0.25) Since less fission product heat is being produced (0.50) c.
Core deltaT will decrease (0.25) Since the amount of decay heat is decreasing (0.50) d.
Loop transit time will increase (0.25) Since the driving head for flow is decreasing (0.50) REFERENCE HTFF Fund Ch 13 " Natural Circulation",~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ pas 2-12 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 003000K3.01/IF 3.7 i l l i (***** END OF CATEGORY 5 *****) _ ____ _ __ .. __ _ _ _ ~ . . . - -. - - .
. w ~ 6.
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pc.co 26 ANSWER 6.01 (2.50) a.
Nuclear power will increase above turbine power.
Tavs will decrease below Tref.
Tavg/ Tref deviation alarm.
g, p 0.5 #-} Steam flow / Feed flow change on #2 S/G.
OR other LOGICAL indications on a case by case basis.[Ad/ 11.^) I b.
1) 2/ igh tainment pressure *f y .ias] (0.00) sw4tgf.;/,I 2) 2/3 oaf low steam pressure blockedaby steamline SI block ty icep isoletion velves she' " uler leep(0.00) ' 3) 2/5B/G hi negative rate steam pressurk'Or e pek s blocke en steamline SI is NOT blocked (P-11)f./as] 'O.50) REFERENCE [)nwig 108Dd89 mp MP3 BOP Vol 2 " Main Steam System". pg 20 and Figure 1 . ~ ~. - ~ ~. . . -~~ . -~ . K&A 039000K1.01/IF 3.1 039000K4.05/IF 3.7 039000A2.05/IF 3.3 ANSWER 6.02 (2.25) a.
Decrease (0.25 for direction, 0.50 for justification) Detectors see less leakage resulting from decreased Tavg b.
Increase Due to the shifting of the neutron flux toward the perimeter of the core c.
Slight increase Due to the decrease in boron absorption in the downcomer region REFERENCE EIb25.Y. d.J b[[$
- '.../
f' fIC 5 f"Off I 1~ [ . . K&A 015000A1.06/IF 2.5 015000A1.07/IF 3.3 015000A1.08/IF 3.3 , (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) , , - , -,, - - -, -, - -, ,,-,,,,,,,_,_n_..,- .,.,. _,.,,,, ..m,.,-,m ,,_,,,,-,,,,,___e___ .__._____g- _ _ _ _ _ _ _ _ _. . __ -
. _ , ' . . 6.
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paca 27 . ANSWER 6.03 (3.00) a.
Compared with Tref to develop rod speed and direction (0.50) R 72-l +++ (0.25) O l + D l + l + s 8-l <+++++++ P l v^ E 0--l---l---l---l-+<+++++++++>+-l---l---l---l-- E v^ l D -8- +++++++> l + l + l + l-72+++ l l l l l l l l l l l l-5 -4-3 -2-1
1
3
5 Tref - Tavg Auctioneered high Tavg (0.25) b.
Tavg is used as the reference for pressurizer level (0.50) Tavg 557F - 587F --> Level 25% - 61.5% (0.25) Auctioneered high Tavg (0.25) 32( Low Tavg coincident with Reactor trip causes feedwater isolation 0.7f c.
-(0.50) Tavs = 564F (0.25) 0,' : -J = _ ( 0. ^ ^i7 ' REFERENCE MP3 NSSS Vol 3 Temperature Indicating, pgs 13-16 Vol 4 Rod Control,pg 23 Pzr Pres & Level, pgs 23-24 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 001000A1.01/IF 3.8 011000A1.04/IF 3.1 059000A3.06/IF 3.2 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) - -._ - . . -. - - - -. . -
.. ' . . 6.
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION raga 28 . ANSWER 6.04 (2.40) CEA a.
MDAFW pump auto start signal-( 0. ' O ::;i. ; TDAFW pump steam supply valve open High radiation on SG BD (SSR-REos) High raAiation on steem jet AE exhaust (ARC -REAh Loss of Yower loss of Air ( A,,y y, g o, yo ea,] b.
Blowdown tank high pressure Main gondenser low vacuum loss of Le# bss o f Air [Arry A> @ 0 YO ea..] REFERENCE hP3 Nsss %I. 2, C,or bi.,,,,e.,d Is./a/to, S sier.,, pl,le CIS -1.
y MP3 BOP Vol 2 Steam Generator Blowdown, pgs 3-4 . ~~ ~. ~. ~~ . ~ ~ -~.., ~ . ~ ~ ~~ K&A 035010K4.03/IF 2.6 ANSWER 6.05 (2.00) a.
1.
>17 percent Pzr Level.
2.
Valves 459 and 460 open.
b.
1.
Loss of Power.
2.
Loss of Instrument Air.
REFERENCE MP3 NSSS Vol 1 ".CVCS". ~page 4 ~ ~ ~ . . ~ ~ ~ ~ ~ ~ K&A 008000K1.02/IF 3.3 i (***** CATEGORY 6 CONTINFID ON NEXT PAGE *****) . . .. . _. _. - . . .. . . . . .. . -
. - . . . 6.
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paga 29 . ANSWER 6.06 (3.00) a.
Train A inputs (PCV 455A) are auctioneered low Th (WR) (0.30) , and wide range pressure (PT 405) (0.30) Therefore, Train A PORV will not open (0.40) Train B inputs (PCV 456) are auctioneered low Tc (WR) (0.30) and wide range pressure (PT 403) (0.30) Since the COP system is armed and the Train B pressure setpoint is low due to the failed To instrument, PORV 456 will open.
(0.40) b.
ALARM warning when temperature instrument fails (0. 50 P ALARM when within 30 psi of setpoint SI and (0. 50 P Olso accepf 4llawi au mM alar-s, y alarms:CoPPS aro,ef CoPPS ac/walef foRV *[t PRT n, a.,d i,ra&calia.,e : failrip de-p. t, PRT level, pn ssure t' REFERENCE PRT / emf. 1, RCS fressur-e L) [ A,,y.2, g o.Sn ea, } MP3 NSSS Vol 4 "Pzr Pres & Level Control" pgs 13-15 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A 010000K4.03/IF 3.8 010000A2.03/IF 4.1 ANSWER 6.07 (2.00) a.
1. Low lube oil pressure.
2. Engine overspeed.
3. Generator differental.
[2 required, 0.5 each] (1.0) b.
The cranking time limit relays time out (7 secs) before the low speed contacts energize the low speed relay.
(0.5) c.
Operation of the engine shutdown reset.
(0.5)
REFERENCE Millstone 3 Vol 1, Diesel Generator and Support Systems, p.
42,50,51.
S&W DWG. NO 12179-LSK-24-9.3J and -LSK-24-9.2D.
l l l l ANSWER 6.08 (1.50) Predischarge alarm (0.5 each) Strobe lights at entrances and within zones Wintergreen scent ! (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) i ._- --. -- .-. . -. - - -
- . . . , U LANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paga 30 REFERENCE MP3 BOP Vol 6 ". Fire Protection System" pgs 7-9 . _,. . . _ . . K&A 086000K4.06/IF 3.0 ANSWER 6.09 (1.35) a.
1.
Energizes both bypass shunt coils (Trips both Bypass Breakers) 2.
Sends Trip signal to both trains of Reactor Protection Q )' g0.5 each] (1.0) to trip RTA, RTB, BYA and BYB.
Y 3.
Gennf Uarni,,y dl arm (on NB sf At the bypass Weaker enclosure.), b.
(0.35) REFERENCE f. M 37.
MP3 NSSS, Vol 5, "RPSAS", page 21) f ANSWER 6.10 (3.00) a.
-Will not open unless RCS (loop WR) pressure is <390 psia.
-Will auto close at 765 psia-Will not open unless valves 8804,8812,8837,8836 are closed.
[2 required, 0.75 each] (1.5) Pressure P /3/ b.
iiend control valve RHR to CVCS (RCV-138-). (0,5) -fou-RWST level lowf(<36 feet) WITH an SI signal present c.
-Motor protection lockout relay-Buss differential lockout relay (Bus 34C or 34D) [Any 2, 0.5 each) (1.0) REFERENCE NSSS MPQ) Vol 1, "RHR".
Mi'3 OP 3908' P I d a l 4 n,. (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
. . . 6.
PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION Paca 31 , ANSWER 6.11 (2.00) (4 of 5 required 9 0.5 each) 1.
Metal water reaction from Zir. MJ.
2.
Pressurizer gas space or " rys es 3.
Radiolytic generation,in water collected on containment floor.
4.
Radiolytic generation in reactor core.
5.
Corrosion of metals by solutions used for containment spray.
REFERENCE MP3, Vol 3, Hydrogen Recombiners, page 6 l l l (***** END OF CATEGORY 6 *****) - - _
. . . 7.
PROCEDURES - NORMAL. ABNORMAL, EMERGENCY Pcco 32
AND RADIOLOGICAL CONTROL ANSWER 7.01 (2.25) 1.
No evolutions are planned which could change the radiological conditions as listed on the RWP.
2.
Any evolution that would change the radiological conditions will be reported to HP.
3.
The work on the RWP does not jeopardize the plant.
REFERENCE MP3 SHP 4912, Para.
4.2, Pp.
3-4.
ANSWER 7.02 (2.50) t a.
Flow may be suspended for up to 1 hour per 8 hour period during performance of core alterations in the vicinity of the vessel hot legs.
(1.0) b.
To ensure the gamma dose rate at the water surface is within limits (2.5 mrem /hr).
(0.5) c.
Shift Supervisor.
(0.5) d.
Excessive (or unanticipated) count rate multiplication (i.e. doubling).
(0.5) REFERENCE
l MP3 OP 3210B, Pp. 5-7.
i , ANSWER 7.03 (1.00) a.
There is a momentary loss of RPCCW flow [0.5]. b.
To avoid tube vibration effects in the RPCCW heat exchangers [0.5]. (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) ! . . .-- =. .. _. .... -. - .. - - - . .
. . . ._ .- . -. .. . - - _ . _. ... . -.. . .
- 7.
PROCEDURES - NORMAL. ABNORMAL. EMIN ENCY Pasa 33 . AND.. RADIOLOGICAL. CONTROL - REFERENCE . MP3 OP3330A, Pp. 6, 21.
ANSWER 7.04 (2.50) ! a.
To prevent contaminants from the RCP seal leakoff line from being forced back into the RCP seal chamber.
(0.5) J b.
30 minutes (0.5) c.
Associated pressurizer spray valve should be taken to manual and closed [0.5] to prevent short cycling the remaining spray flow [0.5]. A
d.
To prevent damage to the RCP motor windings.
(0.5) REFERENCE MP3 OP 3301D Pp. 5-6.
ANSWER 7.05 (3.50) < a.
1.
Terminate the startup by fully inserting all control banks.
2.
Recalculate the ECP.
(1. 0.
b.
Energize PZR heaters (to induce spray flow).
(0.5) c.
Because with the reactor in the source range, positive reactivity must not be changed by more than one controlled method at a time. (0.5) d.
Rod motion is stopped every 114 steps, and bank demand position is compared to digital rod position indication.
(1.0) e.
Be in Hot Standby within the next 15 minutes.
(0.5) REFERENCE MP3 OP 3202, Pp. 6, 7-8, 10.
!
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_ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _. _ _ _ _ _ _ _ _ _ _ _ . . . 7.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Paga.34
AND RADIOLOGICAL CONTROL ANSWER 7.06 (1.50) a.
When starting the pump, if the breaker has not closed (proper amperage) the other pumps will Auto-Start if their switch is not in Pull-to-Lock [0.75]. b.
To prevent seat leakage past the SGWLC valves [0.75]. REFERENCE MP3 OP 3319A, Pp. 6, 12.
ANSWER 7.07 (2.50) a.
5(N-18) = 75 REM [0.30]. Total lifetime to date = 72.5 + 1.5 = 74 Rem.
Total lifetime available = 75 - 74 = 1.0 Rem [0.30]. Total this quarter available = 3 - 1.5 = 1.5 Rem [0.30]. Lifetime is more restrictive than quarterly limit.
0.85 Rem /Hr gamma + (.03 Rad /Hr)(10 QF) neutron = 1.15 Rem /Hr dose rate [0.30]. I 1.0 Rem /1.15 Rem /Hr = 0.87 Hrs = 52 Minutes [0.30',. b.
25 Rem whole body one time exposure [0.50]. c.
Director of Site Emergency Operations [0.50]. REFERENCE 10 CFR 20 MP3 SHP-4902, pas 13-14 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A Plant Wide Generic #15/IF 3.4 { (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. ,. . 7.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pcca 35 AND RADIOLOGICAL CONTROL - ' ANSWER 7.08 (2.50) W "'if
- "-
I# O lip q e a[fouer E050.], OR - I'f air [mwre rY h g is
" ' ' ' ' ~ ~ " ' ' bl ten $tif Oe reac by- [0 95*.), b.
Continue at power 9Y $..m.. 032._. _ _ s t._,.._. ,,w c.
.,... w ..., d.
Trip e.
Shutdown [0.5 ea.] REFERENCE AOP-3562 " Loss of Instrument Air" pg 3 AOP-3553 "High RCS Activity" pg 3 AOP-3554 "RCP Trip or Seal Failure" pg 3 AOP-3559 " Loss of Condenser Vacuum" pg 3 AOP-3565 ".. Loss of Containment Vacuum /.. Integrity". pg 3 ~ . .... ~ ~ .. ~ . .. ~ __ ~~ K&A System Wide Generic #10/IF 4.1 ANSWER 7.09 (1.00) l 1.
RCP bearing oil temperature exceeds 190F.
2.
RCP thermal barrier cooling flow is lost, AND seal injection flow CANNOT be maintained >6 GPM.
3.
RPCCW pressure CANNOT be restored to BOTH containment headers.
] 4.
RCS temperature is >400F, AND VCT temperature is >135F.
5.
VCT temperature is >150F.
[Any 2, 9 0.5 ea.] REFERENCE MP3 AOP 3561.
l l I (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) . _., _ __ . . _ -. - .- -. _ _ _ _ _. _ _ _ _ -. -
. . . 7.
PROCEDURRR - NORMAL. ABNORMAL. EMERGENCY-Pcca 36 - AND RADIOLOGICAL CONTROL ANSWER 7.10 (2.00) a.
Containment temperature [0.25] >180F [0.25], containment radiation [0.25] >10E5 R/Hr. [0.25]. b.
Instrument accuracies (or indications) are affected (by high temperature and/or radiation levels.) [1.0] REFERENCE MP3 EOP Background Document Westinghouse Emergency Response Guidelines Executive Volume.
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A System Wide Generic #10/IF 4.1 ANSWER 7.11 (0.75) Goes on with the procedure [.5] while still attempting the RNO action.[.25] REFERENCE CAT, CNS Exam Question Bank Number Q-EP-EP1-51.
MP3 EOP Training Text, Chapter on EOP Format and Use, P. 4.
ANSWER 7.12 (2.50) a. /.When directed to transition from EOP 35 E-0 hr 8 g(0.50 each] 3. Whe,yhen instructed to implement the CSFSTs in EOP 35 E-0.
A12 <de r,'*4 Eof 35' EC4- 0.0 , b.
2.
Core C6oling - Red [0.25 each, proper
1.
Containment - Red sequencerequired} 6.
Suboriticality - Orange 4.
Integrity - Orange 3.
Heat Sink - Yellow 5.
Inventory - Green REFERENCE EOP Training Text, Pp. 11-16 of EOP Development Cha ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~pter.
~~~~~~~~~~~~~~~~~~ K&A 000074SG# 10/IF 4.7 (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) _ __
_.
. . . . 7.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Paga 37 ' AND RADIOLOGICAL CONTROL ' ANSWER 7.13 (0.50) May result in additional void formation in the RCS.
REFERENCE MP3 EOP 35 ES-0.3, P. 7.
(***** END OF CATEGORY 7 *****) -. - - - - .. - . . - - .-
. . . 8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Pcca 38 AND LIMITATIONS
ANSWER 8.01 (1.50) 1.
The work is coordinated between the SS/SCO/SRO and the Job Supervisor (s).
2.
Any workers protected by additional tags / clearances, (that would be affected by operation of the equipment), shall be advised and remain in the clear before the tags are lifted.
3.
The lifted tags shall be maintained with the work documentation in the custody of the SS/SCO (until the operational evolution is completed).
REFERENCE MP3 ACP-QA-2.06A, P. 16-17.
ANSWER 8.02 (1.50) a,b,e,g, h REFERENCE MP3 ACP-QA-2.02C Pp 8-10.
-ANSWER 8.03 (3.00) a.
>500 psig, >200F (1.0) b.
Discharge first, then suction.
(1.0) c.
No (0.5) d.
Chemistry samples the immediate area and inside closed volumes for explosive concentration of hydrogen.
(1.0) REFERENCE OP 3250, Pages 4-8.
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
. .. . 8' ADMINISTRATIVE PROCEDURES. CONDITIONS.
Paco 39 . AND LIMITATIONS - ANSWER-8.04 (2.00) a.
1.
NRC 2.
NRB [0.25 ea.] b.
If the second verification would result in significant radiation exposure.
[0.50] c.
SS and Duty Officer [0.25 ea.] d.
If identified and controlled in another approved procedure. [0.50] REFERENCE ACP-QA-2.06B, Pp 7-10, 12.
ANSWER 8.05 (1.50) 1.
All safety related systems.
2.
. Components that if mispositioned could cause an unmonitored release of radioactive material to the environment.
3.
Systems considered important to the safe shutdown of the plant.
4.
Surveillances whereby the valves manipulated could directly affect the safety function of a system.
[Any 2 9 0.75 ea.] REFERENCE o2 MP3 ACP-QA-y.12, O f.
ANSWER 8.06 (1.50) a.
Apparatus and mechanisms, the manipulations of which directly effect the reactivity or power level of the reactor, b.
An individual may manipulate the controls as part of his training to qualify for an operator license under the direction and in the l presence of a licensed operator or senior operator.
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) .. . - - -. .. . - - -.. -. - - -
i
. . . 8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Paco 40
AND LIMITATIONS REFERENCE 10 CFR 55 ANSWER 8.07 (2.00) a.
Based on not exceeding 10 CFR 100 exposure limits at the site boundary [0.5] following a STGR. [0.5] b.
Ensures that the resultant offsite dose will be limited to a small fraction of 10 CFR 100 values [0.5] in the event of a steamlime rupture [0.5]. REFERENCE ' MP3 TS B3.4.8 MP3 TS B3.7.1.4 ANSWER 8.08 (2.50) 1.
Max. S/G tube leakage of 500 spd, (exceeded by leakage through #1 S/G - 619.2 gpd).
2.
Max, unidentified leakage of 1.0 gpm, (exceeded by remaining unidentified leakage of 1.13 gpm).
REFERENCE VEGP, TS 3.4.6.2.
MP3 TS 3.4.6.2.
MP3 TS B3.4.8.
ANSWER 8.09 (3.00) ' a.
Pg 3/4 7-4, para 3.7.1.2 Action Statement "a" (72 hours) [0403 f 0 403 b.
Pg 3/4 5-3, para 3.5.2 Action Statement "a" 1 -1, 3.1.3.3. b (NOT an LCO) [0 Go](72 hours) 4 1-10, 'parapara 3.1.2.4 (NOT an LCO) L d. 2 03 c.
5-3, 3.5" 3.a CNO Tan L CO) [0.,20) 3/4 6-12, fre para 3.6.2.1 Action Statement (72 hours) [0.40,7 . d.
c.
Pg 3/4 6-9, para 3.6.1.5 Action Statement (8 hours) [0.60.eeeh] , (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) , . _ - _ _ ... -. - -. - - - - -. - ... .. ..
. - - ~ -. . _- .. _, _ _ - - _ .- .. . . 8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Pcco 41
AND LIMITATIONS REFERENCE MP3 TS Pages as noted in answer.
ANSWER 8.10 (2.00) A diesel fuel oil transfer train is out of service due to the inability of components to perform their intended functions [1.0]. Thus, you are in an action statement and the startup CANNOT proceed [1.0] (until the LCO is met.
Entry into a operating mode is not allowed unless all LCOs are met for that operating mode without reliance on action statements.)
REFERENCE > MP3 TS 3.8.1.1 and 3.0.4.
, ,
ANSWER 8.11 (3.00) a.
If the emergency condition is classified as an Alert, Site Area Emergency, or General Emergency.
(1.0) b.
1. Will assembly help or hinder the situation? 2. Personnel dose rates exceeding limits (limits: >10 mrem /hr or I-131 levels > 10 x MPC) 3. Personnel exposure exceeding limits (limits: >500 mrem whole body or 500 I-131 MFC hours) l 4. Actual or potential personnel hazards exist.
[0.5 ea.] . REFERENCE MP3 EPIP 4010A, P. 5; EPIP 4010B, P. 3.
! ' ANSWER 8.12 (1.50) 1.
Authorized individuals responsible for direct operation of the unit.
2.
Authorized personnel who may be required to support or advise the operation.
' . 3.
Resident NRC inspectors.
. (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) i
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. . . 8.
ADMINISTRATIVE PROCEDURRS. CONDITIONS.
Pcco 42 AND LIMITATIONS
REFERENCE ACP 6.01 Pp. 9-10.
t (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********) -- . . . - -. - . - - - - -. _. ---._._--_. __,__ . - - - - - _ _ -. _ -, _ - - - _.. _ _ _
- -. - .. - .-. . . . . . - - . TEST CROSS REFERENCE Pega
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. . e MILLSTONE-3 REACTOR OPERATOR DECEMBER 15, 1986 1.02.a Facility Comment: The question combines independent concepts: MP3 definition of Shutdown Margin, and actual shutdown reactivity effects.
This may lead to misinterpretation, as MP3 SDM calculations add other conservati sm to actual reactivity beyond the stuck rod assumption; c.andidates may make assumptions-beyond face value of the question and should be evaluated on a case basis.
(Ref. OP 3209) Resolution: Candidates' answers will be evaluated closely on a case basis.
The question requires candidates to " state any assumptions," and candidates will be graded based on any assumptions stated.
If no assumptions are stated by a candidate, his answer will be graded in strict accordance with the answer key.
1.06.c Facility Comment: Agreed that a range of values greater than 0, but less than 1% would be acceptable as an answer.
Resolution: Answer key modified accordingly.
1.09 Facility Comment: Question asks for effects upon DNB (not DNBR, and not " proximity to DNB"), which is misleading.
Candidates may interpret literally and provide opposite direction from answer key.
Agreed that this may have to be graded on a case basis.
Resolution: Candidates will be graded on a case-by-case basis, , assumptions and consistency within the answers will be considered.
___. _- _ -.. - - - -
(- - . . 2.01.a Facility Comment: Key excluded LOP.and CDA signals which start MDAFW pumps.
(Reference: AFW Text p.
9) Resolution: Answer key modified and points redistributed.
2.03.b Facility Comment: , Agreed that statement that steam dump operation is blocked is not required for full credit as long as ' obvious understanding of plant response is evident.
Resolution: Candidates will be graded based on an understanding of the operation of the steam dump system.
2.03.c Facility Comment: ,, _.
As 2.03.b ' / , Resolution: . . . As 2.03.b - . ' 3.02.c Facility Comment:
, , Low Tave coincident with reactor trip not low-low Tave causes FWI.
(Reference: MP3 Text Temperature Indicating System p.
14) Resolution: Answer key changed to Low Tavg, vice low-low Tavg.
3.05.a Facility Comment: Agreed that answer key should account for I.R.
high flux trip.
If trip is blocked, C-1 is blocked.
If IR,'.. ' ' instrument fails high and no block is present, a reactor trip will occur.
' , Resolution: ' ai _.
Answer key modified to allow for candidates' assumptions, will be graded on a case-by-case bas.ss.
, +
.
3.08.d Facility Comment: Agreed that_" Prevent water passage through safety valves" could be acceptable.
Reference: CRx Trip Handout] Resolution: Modified answer key to allow for additional answer.
4.01.a Facility Comment: Question gives value of 48 Rem lifetime exposure, Key uses 44 Rem.
Resol'ution: ' ,. Answer key corrected.
t, i I 4.02.a Facility Comment:
- ![
/ Agreed that ADP 3562, Loss of Instrument Air does not require shGtdown os plant based on air pressure.
(Ref: AOP 3562) ' i' '),f .,,, b
r Resolution: , -fu ( Answer key changed to read, "Continuh at power CO.53, , OR - If a:r pressure is rapidl/f ' decreasing CO.253, then ' e / I' trip the reactor CO.25]./ ' , , , , l 4.02.d Facility Comment: J' Agreed that will accept tripping the reactor if proper assumptions s'peci+ied i.e.
AOP 3555 Reactor Coolant ' - Leak directs operator to trip the reactor if PZR LVL-
- ' ! continues to decrease with both CCP's running.
(Ref: , ADP 3555) f - i Resolution: , Candidates' answers will be graded on a case-by-case basis depending on assumptions.
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e 4.04.a Facility Comment: Agreed that question was vague and placing RHR in service at 390 F could have many effects.
OP 3201 says not to initiate RHR when RCS temp >350 F and pressure >440 psia.
(Ref: OP 3201, Plant Startup) Resolution: Question did not ask whether or not RHR could be initiated at the given conditions, the question was what would happen IF RHR were initiated.
Question will , be graded based on Candidates' assumptions.
4.04.b Facility Comment: Agreed that answer key should allow for RHR suction valve interlocks.
RHR suction valves from RCS are interlocked to prevent opening if RCS wide range pressure is >390 psia, preventing RHR from being initiated at 500 psia.
(Reference: RHR Chapter p.
3) Resolution: Same as 4.04.a above.
The following additional changes were made to this exam during the grading process: 1.06-Answer key corrected to show specific values for Millstone 3; also, reference added.
2.04 If Candidate assumes that a CDA has initiated, then will allow credit for a Quench Spray Pump (but not for an RPCCW Pump).
2.05 Corrected point distribution on answer key to be consistent with question.
2.09 Modified answer key to denote that part of the answer was supplemental information for the grader.
3.01 Parts a and c were deleted.
Several candidates assumed
that control rods would move to compensate for temperature chant,es.
The questions became ungradable due to too many variables.
4.08 Added reference.
l 4.09.a Added additional answer to key and redistributed i points.
!
l - - - .- , __ .. - - . - -. .
1Tnebn7ctyf Y c . MILLSTONE-3 SENIOR REACTOR OPERATOR DECEMBER 15, 1986 5.03a Facility Comment: Transient SUR equation is not included on the SR0 equation sheet.
Agreed that an accurate discussion of SUR during rod withdrawal will suffice for full credit.
Resolution: Only for the portion of the question which deals with SUR imme-diately BEFORE rod motion stops - an accurate discussion of transient SUR will be accepted for full credit. The remainder of the question will be graded as keyed.
5.05 Facility Comment: Question asks for " Fuel Temperature Coefficient", however the answer key appears to deal with the Doppler Power Coefficient. Agreed that the actual answer is given by #1 in the key and must be graded on a case basis.
(Ref. Millstone Rx Theory, Chapter 6, pp 317-321) Resolution: If keyed answer #1 is given, it will earn full credit.
If keyed answer #2 or #3 are given, they will car, half-credit each. More than full credit will not be given.
Points will be subtracted for any information given which is incorrect.
5.06 Facility Comment: (same as comment on 1.02 of R0 exam) 1.02a Facility Comment: The question combines independent concepts: MP3 definition of Shutdown Margin, and actual shutdown reactivity effects. This may lead to misinterpretation, as MP3 SDM calculations add other conservatism to actual reactivity beyond the stuck rod assumption; candidates may make assumptions beyond face value of the question and should be evaluated closely on a case basis.
(Ref. OP 3209) Resolution: i Candidates' answers will be evaluated closely on a case basis. The question requires candidates to " state any assumptions," and candidates will be graded based on any assumptions stated.
If no assumptions are stated by a candidate, his answer will be graded in strict accordance with the answer key.
- - - - _ -..
- I . . 6.Olb Facility Comment: Answer #2 no longer includes loop isolation valves shut.
(Ref: 108 D. 684 Sh 7 Rev. 6) Resolution: Removed reference to loop isolation valves position from the answer key.
Point distribution remains unchanged. Added MP3 Drawing 108D684, Sh 7 to the reference list.
6.03c Facility Comment: See 3.02 c - Low Tave, not low-low Tave coincident with reactor trip causes FWI.
3.02c Facility Comment: Low Tave coincident with reactor trip not low-low Tave causes FWI.
(Reference: MP3 Text Temperature Indicating System p. 14) Resolution: Answer key changed to Low Tavg, vice low-low Tavg.
6.09a Facility Comment: Agreed that answer may be reactor trip and general warning alarms on MB4.
(Reference RPSAS Text pp 36-37) Resolution: Answer key modified to accept " General Warning Alarm (on MB4)" as an additional correct answer.
Two answers required at 0.5 points each; total question value remains unchanged. Added facility supplied reference to the list of references.
6.10b Facility Comment: Agreed answer is PCV131 (Reference: OP 3208 Plant Cooldown) Resolution: Answer key changed to " Pressure control valve RHR to CVCS (PCV-131)."
Point distribution remains unchanged. Facility supplied reference added to the reference list.
6.10c Facility Comment: Agreed that Low low level RWST not Low level RWST trips RHR pumps.
(Reference: ECCS Chapter p. 82)
- , . Resolution: Answer key changed to require "RWST level Low-low..." vice "RWST level low...". Point distribution remains unchanged, and the facility supplied reference is added to the reference list.
7.08a Facility Comment: See 4.02a 4.02a Facility Comment: Agreed that AOP 3462, Loss of Instrument AIR does not require shutdown of plant based on air pressure.
(Ref: A0P 3562) Resolution: Answer key changed to read, " Continue at power [0.5], OR - If air pressure is rapidly decreasing [0.25], then trip the reactor [0.25]." 7.08c Facility Comment: Agreed that SHUTDOWN should also be acceptable since we are not licensed for 3 loop operations.
Resolution: The NRC did not agree to accept both " Shutdown" AND " Continue at power (<37%)." Since Millstone 3 is not licensed for 3-loop operation, the only acceptable answer can be " Shutdown." Answer key changed to " Shutdown", and point distribution remains un-changed.
It should be noted that reference to and/or procedures for 3-loop operation appear in several places throughout the reference material supplied to the examiners, INCLUDING the Technical Specifications.
(A copy of the unit's operating lictnse was not forwarded with the reference material.)
It is recommended that the facility change all procedures and training materials which imply or allow 3-loop power operation such that they are in conformance with actual license limitations.
7.12a Facility Comment: Agreed that when entering ECA 0.0 is acceptable answer.
(Reference ECA 0.0, Loss of All A.C. Power) Resolution: Answer key changed to read: "a.
1.
When directed to transition from E0P 35 E-0.
" .
,- . . 2.
When instructed to implement the CSFST's in E0P 35 E-0.
3.
When entering E0P 35 ECA 0.0.
[Any 2, 9 0.50 each]" E0P 35 0.0 added to the reference list.
The following additional changes were made to this exam during the grading process: 6.11 Keyed answer #2 has been expanded to accept either " Pressurizer gas space OR System H2", for full credit.
8.05 Reference corrected to "MP3 ACP-QA-2.12, P.S."
8.09 Following components added to answer part c: "Pg 3/4 1-8, para 3.1.2.2.b (NOT an LCO) Pg 3/4 5-3, para 3.5.2.a (NOT an LCO)" Three components now required for part c, at 0.20 points each. Point distribution for remaining parts of answer remain unchanged.
5.02 Figure title " Boron concentration vs.1/M" added to answer key.
5.03 Redistributed point values because original distribution did not total question value of 2.50.
Each component of part "a" now worth 0.30 (was 0.25), for total part "a" value of 2.10; each component of part "b" now worth 0.20 (was 0.25), for total part "b" value of 0.40.
5.06 Incorporated reference: MP3, Reactor Theory, Ch. 5, Subcritical Multiplication.
I 6.01 Clarified point breakdown on answer key.
6.02 Incorporated reference: MP3 NSSS Vol. 3. "Excore Nuclear Instru-mentation," Pg. 9-11.
l 6.03 Deleted "2/4 logic" from keyed answer for part "c", because this is not elicited by the question.
Redistributed part "c" point break-down such that the description of the FWI/Tavg relationship is worth l l 0.75 points (was,0.50).
, 6.04 Added in parenthesis "SSR-RE08" behind "high radiation on SG BD", and " ARC-RE21" behind "high radiation on steam jet AE exhaust."
l Added the following to list of acceptable answers for part "a": - CIA - Loss of Poder - Loss of Air Require any 4 (of 7), at 0.40 point each. Total value of part "a" remains unchanged.
- _ - _ - - - - -. - - - - - . _ _ - _ _ - -.._,
. .- .. . , . Added the following to list of acceptable answers for part "b": - Loss of Power - Loss of Air Require any 2 (of 4), at 0.40 point each. Total value of part "b" remains unchanged.
Added "MP3 NSSS, Vol. 2, Containment Isolation System, Table CIS-1" to list of references.
6.06 Part "b" of answer key modified to "also accept following alarms: COPPS armed COPPS actuated, PORY open, SI and associated alarams, and indications: tailpipe temperature increasing, PRT level in-creasing, PRT pressure increasing, PRT temperature increasing, and RCS pressure decreasing." Any 2 (of 11) required at 0.50 point each.
During administration of the exam, the following changes were made: 1.
Original question 5.02 was deleted and replaced with a new question to determine candidates' ability to predict criticality using a 1/M plot.
(Question and key answer attached as part of Master Copy of exam.)
2.
Question 6.03 clarified. Now reads, "... explain what the Tavg input signal is, and WHAT control functions the Tavg signal provides..." Announcement made to all candidates concerning , this change during the exam.
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