ML20148G492
| ML20148G492 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/23/1988 |
| From: | Dave Hills, Jordan M, Nejfelt G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20148G484 | List: |
| References | |
| 50-263-OL-88-01, 50-263-OL-88-1, NUDOCS 8803290162 | |
| Download: ML20148G492 (60) | |
Text
iu U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-263/0 L-88-01 Docket No. 50-263 License No. OPR-22 Licensee: Northern States Power Company Monticello Nuclear Generating Plant Monticello, MN 55113 Facility Name: Monticello Nuclear Generating Plant Examination Administered At: Monticello Nuclear Generating Plant, Monticello, Minnesota Examination Conducted: Marc 2-3, 1988 Examiners: W'E.
D. tii 3/23/7o
. Date h~ 4 W M G. M. Nejf Date Ge w lda
& Approved By: . . r J!7 3 86 Date' /
Examination Summary
. Examination administered on March 2-3, 1988 (Report No. 50-263/0L-88-01) l A written examination was administered to one Reactor Operator retake candidate. In addition, operating examinations were administered to two Senior Reactor Operator retake candidates.
Results: All candidates passed these examinations.
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l i goo 3290162 880324 I PDR ADOCK 05000263 V DCD d L ..
, DETAILS
- 1. Examiners D. E. Hills, Chief Examiner G. M. Nejfelt
- 2. Exit Meeting At the conclusion of the examinations, an exit meeting was held. The following personnel attended this exit meeting:
Facility Representatives L. Eliason, General Manager Nuclear Power Plants W. Shamla, Plant Manager .
L. Waldinger, Manager, Production Training D. Antony, General Superintendent, Operations M. Brant, Site Superintendent B. Schmitt, Nuclear Training Administration B. McGillic, Operations Training Supervisor NRC Representatives M. Jordan, Operator Licensing Section Chief D. Hills, Chief Examiner G. Nejfelt, Examiner P. Hartman, Senior Resident Inspector At the exit meeting, the NRC representatives indicated that no generic deficiencies or concerns were identified during the examinations. In addition, the status of the NRC requalification examination pilot program was discussed.
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ATTACHMENT 1 Resolution to Facility Comments on Monticello R0 License Examination of March 2, 1988 QUESTION 1.12 Facility Comment: Double-jeopardy question - cannot answer second half of question if first half was not understood.
NRC Response: Disagree with comment. This question was not asked in two distinct parts. It was worded as part of the same integrated question to determine whether the candidate had complete and overall understanding of the term "reactivity anomaly." This includes why it may occur.
QUESTION 1.14(c) and (d)
Facility Comment: Rod Worth will tend to follow the keff vs. core age curve and will tend to decrease over core life for part (c). Part (d) should be "remains the same." The change in fuel temperature results in an insignificant change in Rod Worth.
NRC Response: For part (c) the answer key already shows "decrease" as a correct answer. Will however also accept "will follow keff vs. core age curve." For part (d) the reference supplied by the facility just states that this effect is small in comparison with the other effects. It does not support the statement that it "remains the same."
QUESTION 2.01 Facility Comment: Due to a recent design change on ',ne HPCI System, the answer is incorrect for current onditions. Should also accept "no adverse action." The rarrp generator is initiated from both the stop salve and the steam admission I supply valve (M0-2036) opening. HPCI should start normally, NRC Response: This recent design change was not depicted in references a originally supplied to the examiner. During the exam, the candidate mentioned this design change to the examiner.
i Tha examiner directed the candidate to then just describe the recent design change instituted to prevent any adverse
- auto 7atic action in this case. Thus the question and answer key have been changed to reflect this additional guidance to the candidate.
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QUESTION 2.04 Facility Comment: Answer is incorrect as stated. A high drywell pressure will cause a full, not partial, Group II Isolation.
NRC Response: Group III isolation was a typographical error; thus, the answer key will be changed to a Group II isolation.
Operations Manual B.4.2. states that this is a partial Group II isolation while the reference supplied by the facility indicates a full Group II isolation. Either is acceptable for credit.
QUESTION 2.08(d)
Facility Comment: RHRSW side of heat exchangers are not normally flushed.
Also, question did not specify which side of heat exchanger. Condensate service water provides flushing for the RHR side of the heat exchanger.
NRC Response: Operations Manual B.8.1.3, Page 3, states that "the demineralized water supply line provides water for flushing the heat exchanger after the system has been operationally tested." However, as indicated in the facility comment, the question did not pecify which side of the heat exchanger. Therefore, either Demineralized Water System or Condensate Service Water is acceptable for full cred R.
QUESTION 2.09 Facility Comment: Upon sensing decreasing pressure, the standby pump will start. However, a runback, if initiated, due to low pressure will not be seen by the operator. The time that the runback is in effect will not cause the speed / load changer to affect the Turbine Control System.
NRC Response: Will accept either turbine runback or automatic start of the standby pump on low inlet pressure for full credit.
QUESTION 2.16 Facility Comment: Question does not look for a technical specification based answer. Other reasonable answers should be accepted.
NRC Response: The cited reference specifically states that the items given in the answer are the major concerns. However, will accept other reasonable concerns.
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QUESTION 3.02 Facility Comment: Answer (d) should be considered because the operator knew enough to notify the control room, and at Monticello that is the way of notifying the fire brigade.
NRC Response: The purpose of the question was to determine whether the candidate knew that the flashing red warning light and the alarm klaxon at the entrance to the Compressed Gas Storage .
Building signified an area radiation alarm. The question specifically stated that there was.only one possible answer. Answer (b) is the only correct answer. While it may be sufficient for particular site personnel to only know enough to call the Control Room under these conditions, a licensed individual should know specifically what the conditions mean in case he receives the call.
-QUESTION 3.05 Facility Comment: Due to a design change for RWM, answer (a) or (b) should be considered because the question only looked for one response.
NRC Response: This recent design change was not depicted in references originally supplied to the examiner. Due to this design change, either answer (a) or (b) is acceptable for full credit.
QUESTION 3.12(a)
Facility Comment: High-high alarm will still result in a stack isolation regardless of whether or not the Storage System is being used.
NRC Response: Agree with facility comment. Although reference cited in answer key does not indicate this, the drawing provided by the facility does; therefore, answer key will be changed to "Stack Isolation."
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QUESTION 3.15(c)
Facility Comment: Other answers should be accepted. Answers such as:
(1) RBM IN0P trip light on C37 (2) RBM IN0P alarm on C05 (3) Process computer alarm (4) C05 benchboard indicating light for Hi-Hi/INOP (5) C05 Recorder fails downscale (6) Rod withdrawal block based on RBM INOP NRC Response: Disagree with facility comment. The question specifically asks for indications other than a RBM IN0P trip and its associated indications.
QUESTION 3.17 Facility Comment: What is covered ab;ut this system is purpose and general function. System specific details, as was asked in the question, are not covered. By our surveillance procedure, the only operator interface with the system is to do the valve lineups.
NRC Response: The calibration procedure supplied by the facility only indicates who operates the Hydrogen-0xygen Analyzing System sw!tch during a calibration check. It in no way indicates who operates this system under real post-accident conditions. The actual operating procedures in Operations Manual B.4.3 do not indicate who is responsible for each step in either the startup, shutdown, or post-accident conditions. (Steps that an operator would clearly do are not delineated from any other steps.) Since this system is required by Monticello technical specifications and the position of this particular control room switch has a direct bearing on operability of the system, it is not unreasonable to require candidate to know the significance of each of these switch positions. In addition, this knowledge is supported by NUREG-1123 K/A ratings of 3.5 and above. In light of no other references supplied by the facility to properly support this comment, the answer remains unchanged.
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QUESTION 4.01 i
Facility Comment: Question is misleading. The procedure in question does
, not restrict us to withdraw only. The E0Ps allow us to use elevated drive pressure to insert control rods.
!C D+ - NRC Response: Operations Manual B.1.3, "Control Rod Hydraulic System,"
Section VI.F.1.1, "Operation of a CRD Under High Drive Pressure," g!ves instructions to "follow the procedure below for CRD withdrawal, but not for insertion." The only insert signals applied in the procedure are an attempt to unlatch a drive with a series of insert and withdraw signals followed by actual withdrawal of the rod. In addition, Operations Manual C.4-B.1.3.B, "Control Rod Drive Failure," gives instructions to "not increase drive pressure beyond 265 psig to insert a control rod. Elevated drive pressure beyond 265 psig above reactor pressure should only be used to withdraw a control rod." The. basis of this procedure goes on to say that "elevated drive pressure may be used to withdraw a drive but not to insert one.
Increased drive pressure to insert a stuck rod may lead to sticking of the directional control valves." However, Operations Manual C.5-1103, "RPV Power Control," Steps 0.8 and 15 B do prescribe using elevated drive pressure to insert control rods. Due to this apparent contradiction between the various Monticello procedures, this question is being deleted from the exam.
QUESTION 4.07 Facility Comment: The intent of the 5 minutes is to ensure that the controller has rundown to minimum. In doing this review, the answer in the answer key was not found in the reference document, therefore, should accept any reasonable answer.
NRC Response: The answer given in the facility comment is just another way of saying what is in the answer key. Alternate methods of saying the same thing is acceptable as long as it is clear that the concept is understood.
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QUESTION 4.08(c)
Facility Comment: This question is unreasonable to expect an operator to know from memory the decay chains of various isotopes.
NRC Response: Agree with comment. The question is deleted from the answer key. The odds are that two different isotopes of two different elements will not decay in exactly the same way with the same energies. The argument still exists, however, failing to make this assumption, the candidate was not given all information required to answer the question.
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\ U. S. NUCLEAR RECULATORY COMMISSION
. REACTOR OPERATOR LICEMBE EXAMINATION
, FACILITV3 .)d Anf.i.sa.OR. ........ ..
REACTOR 1YPEt SWB-GE2........ _________
DATE ADMINSTEREDI 90/93/92____..__..._____.
EXAMINER! UILLSt_QA.____...__......
CANDIDATE .....______.. _____..____
INSIBWCIIDMS_ID GeHDIDeIE1 Use---sepente--pepet-fer-the-ensweMr--MHK.t-enswen cn ene-si1de-erriy .
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing Stade requires at least 70% in each category and a final grade of at least 80%. Exasiination pape s~ ill be iered six (6) hours after the e::aniination starts. g 7 [ gp[ IN i LLi *
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% OF CATECORY % OF CANDIDATE'S CATEGORY l __VeLVE. .IDI6L ___EG9BE___ _ Y 6 6 V E _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ G eIE G O BI_ _ _ _ _ _ _ _ _ _ _ _.
.2E1E9._ 2EASS ___________ ________
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW i .20499._ .26225 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY
- AND EMERGENCY SYSTEMS
.25299._ .29223 ___. ______ ________ 3. INSTRUMENTS AND CONTROLS
.2522E__ .25462 ___________ ..______
- 4. PROCEDURES - NORMALe A E:N O R M AL ,
ENERGENCY AND RADIOLOGICAL CONTROL
.19922f_ ___________ ___-_____% Totals Final Grade All work done on this exasiination is siy own. I have neither given nor received aid.
Candidate's Signature l l l l l
l i . . . a . NRC RULES AND CUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules applyt
- 1. Cheating on the examination beans 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 i re the date on the cover sheet of the exaniination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name ir, the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Cons e cutively nurober each answer sheet, write 'End of Category __' as appropriate, start each category on a new page, write only on one side ot the paper, and write 'Last Page' on the last answer sheet.
- 9. Number each areswer as to category and nus,ber, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separ ate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use zbbreviations only if they are commonly used in facility literature.
l 13. The poirit value for each question is indicated in parentheset after the ! question and can be used as a guide for the depth of answer required. l l 14. Show all calculations, methods, or assumptions used to obtain an answer to mathesiatical problems whether indicated in the question or not.
- 15. Partial credit siay be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LE AVE ANY ANSWER E:L ANK.
- 16. If parts of the exas ination are not clear as to intent, ask questions of the examiner only.
- 17. You sust sign the stateen nt on the cover sheet that indicates that the work is your own and you have not received or been given assistance in cosipleting the exas.ination. This sivst be done af ter the exas-ination has j been cospleted.
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- 10. When you complete your examination, you shall! -
- e. Assemble your examination as followst (1) Exam questions on top.
(2) Exas sids - figures, tablese 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 f or answerins i.he,grestiens .
- d. Leave the examination area, as defined by the exasiiner. If after leaving, you are f ound in this area while the examination is still in progress, your license may be denied or revoked.
2"--flililiklis!Ei!!iki!!rillitil!!II:lE5!8!!!I6 "c
- GUESYION 1.01 (1.50)
Indicate whether each of the following statements are TRUE or FALSEt
- a. Core flow reduction through the cer. tral fuel bundles is minimized by placing smaller orifices at the inlet of the outer (peripheral) fuel bundles.
- b. Core flow will decrease if power is reduced from 100% by control rod insert 1on.
- c. Following a scran, froni power operations arid the establishnient of a cooldown rate, actual reactor water 1cvel will tend to increase as a result of the cooldown if the reactor level control system is in n.anual s. ode and no operator action is taken tc prevent it. ( Assume a stable level exists at the start of the cooldown.)
GUESTION 1.02 (1.50) List three operational siethods that will lengthen the cycle of plant operation. QUESTION 1.03 (1.00) Monticello Operations Manual C.5-2007 ' Level / Power Control' prescribes the siethod of lowering reactor water level in order to reduce the he st generation rate. Explain how lowering reactor water level will assist in reducing the heat generation rate. t (usuma CATEGORY 1 CONTINUED ON NEXT PAGE xxxxx) b
11..tBIWGIRLE5_DE.UWCLE6B t0 WEB.tL6HI 9tE66IIQUt Pase s IHE85991HeBIGEtWE61.fB6HEECB_680_ELUIQ.CL9W QUESTION 1.04 (2.50) Choose the correct responses in parentheses for each of the following stat.ements regarding Net Positive Suction Head!
- a. The Net Positive Suction Head REQUIRED by a recirculation pump (increases, decreases) with increasing pump speed.
- b. Increasing reactor power fros 70% to 90% by increasing v erTrev1Murrpunipm eed yi e l d s, i b i stTen-tw e r ) downruner temperatures and (increased, decreased) recirculation pump AVAILABLE Net Positive Svetion Head.
- c. A loss of feedwater heating (increases, decreases) core inlet subcooling resulting in ( st o r e , less) r e c i r cul a t ier:
pus p AVAILAOLE Net Positive Svetion Head. 00ESTION 1.05 (2,50) Choose the ccrrect responses in parentheses for each of the f ollowing sta tements regarding Reactor Pressure Vessel (RPV) water level indication. ! a. Elevated reference les temperature will (increase, decrease) the derts ity of water i re the reference lege resultirig in a ( s a,a l l e r , 1st ger ) serised differential pressure arid a (lower, higher) iridicated level.
- b. As siost RPV uater level indications sieasure the level in the l downcoser annulus, actual RPV water level is slightly (below, l above) indicated level at 100 % power.
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- c. When the reactor is in cold shutdown, level instrusientation I
calibrated at operating coriditions will indicate (higher, lower) than actual RPV level. QUESTION 1.06 (1.00) EXPLAIN WHY during high power operations (>60%) it is siore desirable to change power with recirculation flow than with control rods. (usurr CATECORY 1 CONTINUED ON HEXT PACE maxxx)
2f--Illlillillsfiiffill!!rilllicittli:lE5ft!!216 QUESTION 1 07 (1.50) The reactor is operating at high power under steady state conditions. A high worth control rod in WITHDRAWN one notch. The time required for power to stabilize is noted. Now the control rod is INSERTED one notch. The time for power to again stabilize is noted. Which transient should take longer for reactor power to stabilize and WHY? OUESTION 1.08 (1.50) For each of the following types of coretrol rods 'x' and 'y', choose those characteristics 1 through 3 which describe that type. (More than oree characteristic MAY apply to each type.) Cor.t r ol Rod Type
- x. Deep Cor. trol Rod
- y. Shallow Control Rod Characteristics
- 1. One-or-t wo riotch withdr awa l c a re result iri reverse power effect.
- 2. Sub s t erit i a l l y affects radial flux distributiore.
- 3. Also Lt.c un as power rods.
QUESTION 1.09 (2.00) For each of the f ollowing transients, indicate which reactivity coefficient (Moder ator Teniperature, Void or Doppler) tends to change powei first AND in what direction.
- a. Fast closure of one Main Stean. Isolation Valve (0 5)
- b. Isolatiere of a feedwater heater string (0.5)
- c. A contrcl rod drop (0.5)
- d. Relief valve lifting (0.5) t l
l t l (musma CATECORY 1 CONTINUED ON HEXT PAGE xxxxx) l_
. 1 11..tBIUCIELES 9E.UUCLE68.t0 WEB tL6HI.0EEB6130Hz Pas? 7 IHEB50DIW651 Git.UE61 IB6HIEEB.6HQ.ELUID.ELOW QUESTION 1 10 (1.50)
For each of the following initial conditions 'a' through *c' choose which DNE of the two conditions *1' or *2' would aske a generator load rejection without bypass a more severe transient.
- a. 1. 80% rated core flow 2. 100% rated core flow
- b. 1. beginning of cycle 2. end of cycle
- c. 1. all rods full out 2. 95% control rod density OUESTION 1.11 (1.00)
The reactor scranis froni full power, equilibrium. Xenon conditions. Four hours later, the reactor is brought critical and power level is maintained on range 5 of the IRMs for several hours. Choose which ONE of the following statenients is correct concerning control rod niotion during this period.
- a. Rods will have to be withdrawn due to Xenon build-in.
- b. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of Xenon burn-out.
- c. Rods will have to be inserted since Xenon will closely follow its nornial decay tin.e.
- d. Rods will approxisietely reniain as is as the Xenon establishes its equilibriuni value for this power level.
l GUESTION 1 12 (2.00) Define the terni ' reactivity anonialy' and list two possible reasons why a reactivity anonialy siay occur. l l t l l (manan CATEGORY 1 CONTINUED ON NEXT PAGE xxxxx) l L
.y 'Millillilidid"iti!!rillihittli:tE5!B216 QUESTION 1 13 (1.00)
The figure below shows K-effective versus core age for a reactor core. INDICATE whether the AVAILABLE SHUTDONN NARGIN will increase or decrease between points "a* and 'b' and EXPLAIN why. a. 1.t s g
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1.14
, 1.13
- i. , b.
i.s s o 1 a a e e r e e 50 CORE A0t (Westi st000) K,77 n. Core 4e DUESTION 1.14 (2.00) Indicate whether each of the fo110 win 3 effects will increase or decrease integral rod worth. (Consider each effect individually and neglect all other influences.)
- a. Increasing reactor water temperature
- b. Increasing voids
- c. Core Age (Rod worth at point of end of life as opposed to beginning of life)
- d. Increasing fuel tesiperatures (sanza CATEGORY 1 CONTINUED ON NEXT PAGE znxux) m . . _ - - -
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ii..tBIWCIELES 9E.UUCLE6B.t0 WEB.tL6HI 9tEB61IOUt Pas) 9 , IMEB500!Heb!GEt.NE61 IBebSEEB.6HQ.ELV19.tL9W ) QUESTION 1.15 (1.50) Choose for which condition 'a' or 'b' the reactivity coefficient contribution would be HORE NEGATIVE and EXPLAIN why.
- a. Void coefficier$t for a 1% increase in void fraction starting at 10% void f raction in the core.
- b. Void coef ficient for u 1% increase in void f raction starting at 70% void f raction in the core.
QUESTION 1.16 (1 50) Answer each of the following in regards to the attached e>:as ple of a P-1 Cor e Perfornience Los:
- a. For each of the following staten,ents, indicate whether it applies to MFLCPR, MFLPD, or MAPRAT froe. the attached e >: a h pl e of a P-1 Cor e Per f or n.ance Log. (1.0)
(1) Ens ures that 1% plastic strain on the clad is not e >:c e e d e d . (2) Ensures that peal. cladding tes.perature of 2200 degrees F is not e::ceeded followins a LOCA.
- b. State whether an APRM GAF of 0.98 fromi the attached l e >: a n.p l e of a P-1 Core Perf orniance Los indicates a higher or lower APRM reading than actual power. (0.5) l 1
l i l (xxxxx END OF CATECORY 1 unzux) l
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DATE XX-XX-XX TIME XXXX *** PERIODIC NSS OCRE bERFORMANCE LOC *** . S20. 880. WW E LOCATION 1 2 3 e 5 6 7 8 9 to 11 12 CMWT 73et. AXf AL REL PWR t' . 69 1.17 1.13 1.37 1.2e 1.18 1.08 1.06 1.00 0.90 0.67 0.3s PCT PWR 98.9 mectoN mEL rwm 9.se 1.Os 0.se 1.09 1.05 1.09 0.se 1.06 0.8e Casert 397.6 RINC REL PWR 1.00 1.00 1.11 1.7e 1.77 f.c9 0.60 Ce8FCP 9.999 APRM CAf* 6.98 1.00 0.99 1.00 1.00 CM*LPD 9.199 ~
- CMPF 2.9H CAEO 9.147 CACA 9.137 p, REClON 1 7 3 e 5 6 7 8 9 CAVF' 9.473 x MFLCPR 9.874 0.900 0.874 0.92e 0.760 0.973 0.874 0.900 0.87a CAPD 97.37
- l 88 LOC 13-te 31-te 39-1e 9-7s 19-ye e3-7s 13-ee 21-e0 39-ee CRO 9.999 l j FLOW 0.1057 0.1085 0.1057 0.1032 0.113e 0.1032 0.1057 0.10e5 0.1057 CRSYN 2.
-. PKF 1.42 1.38 1.e2 1.e9 f.76 1.49 1.e7 1.38 1.e2 PR 1993.
- MFLPD 0.987 0.927 0.987 0.989 0.79e 0.999 0.9R7 0.927 0.987 DPC-M 96.13 F_8,
, LOC 13-to-3 19-17-5 39-Jo-5 9-28-6 33-3 t'-e e3-76-6 13-=0-5 33-e7-5 39-ec-5 DPC-C 19.99 .
PKFL 2.se 2.30 2.te 2.45 1.97 2.e5 7.4e 2.30 2.ee RW1. 38.99 e 0.977 0.977 DHS 39.48 7 = MAPRAT LOC 0.977 0.978 13-1e-5 19-12-6 39-14-5 0.977 0.980 9-28-6 0.737 0.980 19-7e-5 e3-7P-6 13-=0-5 19-e2-6 0.97= 39-to-5 WFW 99.92 1.36
- FD 39.93 PKFS 1.36 1.e1 1.36 1.e3 1.76 1.43 1.36 1.et g WTSUB 73.99 o e FTHE +5.99 i e 9 FAfLED SENSORS 2 WT 99.28 T UASE CRIT CODE PCT WTR 89.9 a FAILED re13.n.i L,PRM LIST'e879. A.2 irr .A.2 FTFLAC 1.
ITER 1 p tREC 9 s THE 12 MOST LIMITING DUNDLES 3 #Eot 1
" IXYFLO 2717.
3 e FOR MFLCPR rOR PNLPD FOR MAPRAT h ID mttvM tut utm 09-28 1.321 tmtm mayu tut. 1.270 uw 0.985 09-78-0G 13.20 wu t m 13.ee nams : 0.eRO tut 9-78-06 uavt%n 19.17 tv-t%s 19.37 < 0.97e 0.9?3 e3-28 1.322 1.220 0.5s5 43-28-06 13.70 13.e0 0.9n0 e3-78-06 19.17 10.37 0.923 09-26 f.327 1.220 0.985 09-26-06 13.70 13.e6 0.980 43-76-05 19.17 19.37 0.922 e3-26 1.374 1.220 0.985 43-26-06 13.70 13 e0 0.9so 9-26-06 19.17 19.37
- 0.900 31-18 1.355 1.270 0.9ee 13-te-05 13.19 13.to 0.977 13-14-05 19.28 19.52 6.900 21-ee 1.355 f.220 0.9se 39-14-05 13.19 13.e0 0.977 39-14-05 19.28 19.52 0.900 31-ee 1.355 1.270 0.9se 13-40-05 13.19 13.ee 0.977 33-e0-05 19.28 19.52 0.900 21.1e 1.356 1.270 0.988 39-e0-e5 13.19 13.e0 0.977 39-=0-05 19.28 19.52 i .
0.896 13-22 1.361 1.228 0.977 31-14-08 13.17 13.40 0.952 31-14-08 11.96 11.59 0.99s 39-32 1.361 1.270 0.977 71-e0-Os 13.17 13.40 0.957 21-ec-OS 11.96 11.59 0.895 13-37 1.363 1.220 0.977 31-40-09 13.12 13.ee 0.957 31-e0-08 11.96 11.59 0.895 39-27 1.363 1.220 0.977 21-14-09 13.12 13.e0 0.967 21-18-08 11.96 11.59 THE NUMBER OF RUNDLES WITH MFLCPR CREATER THAN 1.0 e 0 THE NUMDER OF BUNDLES WITH $8FLPD CREATER THE NUMBER OF DUNDLES WITH MAPRAT GREATER THAN 1.0 = 0 THAN1.0w 0 l Figbre 2 9 y I ..-
3 . ESIGH ItfGWDINE 56EEII.eWD.EBERGENCY Pcse to l QUESTION 2.01 (1 50) A ground in the 125 VDC System is traced to the electro-hydraulic governor on the High Pressure Coolant Injection (HPCI) System turbine. In an attempt to clear the ground, the HPCI Auxiliary 011 Pump is started to flush oil through the governor. The ground does not clear so the HPCI System operability test is initiated to demonstrate that the ground is not af fecting HPCI systesi operability. This test requires simultaneous start of the Auxiliary 011 Pump and opening of the Turbine S'.eas Supply Valve (h0-2036). Since the Auxiliary Dil Pun.p is already running, the operator opens the Turbine Stean Supply Valve. OCCCn!PE - r. y - s.m. -- - . a - . , . 4 ,. .-.<-- ; ; , , c ; -g , g t % i ,g + 4 ,4 4 r, g e < g r. , 7 t
. e e "1 + 4 r. ; h e- thir e g ; i c,7 a L vn oud EXPL A!!' Aj 14 unnld occur m Des u ,'6t a re. s d A e 5:3 a c ha 9 s sw s& httJ 4.p* ew s 4 awy a be m e A M A ai_b b s apan 4-h t kes t .p t .-* A w c. d *% .
QUESTION 2.02 (1.50) Describe HOW and WHY the Control Rod Drive Hydraulic Systen Flow Control Valve (3-19) will react to a reactor s c r a p. during power oper ations. QUESTION 2.03 (2 00) List f our means by which Recirculation Motor-Gener ator Set No. 11 will ba affected by a loss of 125 V DC E:vs 'A.' i 1 QUESTION 2.04 (1 00) Describe the dif f erence in the autoniatic actions that will occur upon an auton etic initiation signal to the Standby Gas Treatment l Systen. (SGTS) due to high drywell pressure versus depressing the SGTS TEST pushbutton. l l l I (zumma CATEGORY 2 CONTINUED ON NEXT PAGE mamax)
Ea*__tL6BI_DESIGH_INGLUDINE.96EEII eWD.EUEBEEWCI Pasa 11 i 815IEBE l l QUESTION 2 05 (1.00) l l Choose which of the following rooms or equipment are within the secondary containment. (More than one answer MAY apply)1
- a. Recirculation Pump Motor-Generator Set Room
- b. Standby Cas Treatsient System Rooms
- c. Fuei 5turgeTotri - -
- d. Residual Heat Removal punip s DUESTION 2.06 (2.00)
Answer each cf the following questions regerding the Drywell Ventilation Systes.:
- a. List the autosiatic initiation signal for Mode 5 (Scran. Mode) of the Drywell Ventilation Systen . (Include the setpcint)
- b. A loss of offsite power occurs sis.ultaneously with a spurious high drywell pressure signal greater than 2 psig. The Drywell Ventilation Fans that are in the ON position trip and all other loads associated with other systems sequence on the proper busses as designed. Subsequently, the high drywell pr essur e signal clears. Other than restoring offsite power and physically restar ting the f ans e what action needs to occur to allow the Drywell Ventilation Fans to restart af ter the high drywell pressure signal clears?
DUESTION 2.07 (7 00) A pressure switch located on the instrusient air line before the dryer automatical)y closes four solenoid valves on low pressure (80 psis) in order to supply siaxisius. air to specific equips.ent. List f our examples of this specific equipsient. (Listing redundant or sisillar types of equipsient in the sanie systes, counts as only one answer.) (ummum CATEGORY 2 CONTINUED ON NEXT PAGE arxxx)
Ia'.. EllGU.IUG'WQldG.I6EEII 6HD.EBEBGENGI Pcsa 12 g{ e QUESTIDH 2.08 (2.00) List the systesi that is the source of water for each of the following regarding the Residual Heat Removal (RHR) Service Water Systeal
- s. Coolant for the RHR Service Water Pump Motor Thrust Bearing Coolers.
- b. Seal water f or the NHN S e r vTe e Niit.e t ' PiTap7hlf t7M 1 glands.
- c. Maintains the RHR Service Water header slightly pressurized and full of water to prevent drairiing and danisse to the loops f rosi water henin.ers when the systen is placed in service.
- d. Water f or flushing the RHR Heat E>: changers af ter the systen, has beers operationally tested.
QUESTION 2.09 (1.00) During high power operations Stator Cooling Water Pus.p P-72A is to be takeri out-of-service f or n, a i n t e n a n c e . As the operator is swapping to Stator Cooling Water Pun >p P-720, he s,istakenly trips Pun.p ' A' pr i or to starting Pusip ' E: ' . DESCRIE:E any adver se consequences of this action and EXPLAIN why it occurs. QUESTION 2 10 (2.00) For each of the following valves indicate how that valve will fail (open, closede or as is) for the stated occurrences:
- a. Low Flow Feedwater Regulating Valve 6-13 Loss of control air (0 5)
(1) Loss of electrical control signal (0.5) (2)
- b. Main Feedwater Regulating Valve 6-12A (0.5)
(1) Lost of control air (2) Loss of electrical contr ol signal (0.5) (sxx : CATEGORY 2 CONTINUED ON HEXT PAGE ::: ) _ _ - .n
Ia .theWI_DEEIGN.IUCLUDING.56EEII.eHR.E9ERGENGX Pcs? 13 IIIIEnf QUESTION 2.11 (1.00) Only one recirculation pump is running at the time of a Low Pressure Coolant Injection (LPCI) System automatic initiation. Describe the automatic actions that occur in this situation (i.e., with only one recirculation pump operating) to ensure that LPCI Recirculation Loop Selection Logic chooses the correct loop in which to inject. QUESTION 2.12 (1.00) A fault occurs on an output load of the Uninterruptible power Supply that supplies power to Class 1E Uninterruptible AC Distribution Panel Y70. This results in a low output voltage of less than 108 volts. Describe how the inverter will respor.d to this condition. QUESTION 2.13 (2.00) DESCRIBE two autos,atic actions which siay occur to the Reactor Water Clennup Systen. upon a loss of Reactor Building Closed Cooling Water AND INDICATE what signals will cause these actions. (2 0) GUESTION 2.14 (1.00) The plant is operating at 100% rated thersa1 power. The Steas Jet Air Ejector Off-Gas Radiation Munitors activate the dual timer circuit and 30 riinutes later a trip of the recontbiner trains results. Explain HOW AND WHY this occurrence will effect plant operation. QUESTION 2.15 (2.00) During an Anticipated Transient Without Scran. (ATWS) event torus wat er tesiperature reaches 110 degrees F and thus boron injection is required. It f.s found that the Standby Liquid Control pus.ps are inoperable for boron injection. State two alternate systes,s that can be used to inject boron and the tank (s) that serve as the source of pentaborate solution for each of these systesis. (:: CATEGORY 2 CONTINUED ON NEXT PAGE Emmax)
21.. g(( EllEU.IUGLVDING.56EEII.6HD.EBEBEENGI Pose 14 QUESTION 2.16 (1.00) The reactor is in cold shutdown (with the reactor head on) with both reactor recirculation loops shutdown. The shutdown cooling mode of the Residual Heat Removal System is operating in a throttled mode which results in thermal stratification of the reactor water. Without sufficient circulation the reactor water cleanup thermocouples register significantly lower than the temperature at the surface of the water. Insufficient heat removal results in an unanticipated increase of surface water ten.perature above 212 degrees F and a pressurization of the reactor-pressure vessel. State the two major concerns with this type of event. 1 OUESTION 2.17 (1.00) During an event a High Drywell Pressure greater than +2 psis and Reactor Vessel Low Low Water Level less than -48 inches result in autometic initiation of the High Pressure Coolant Injection System (HPCI). Reactor water level is restored and continues to increase until the HPCI turbine automatically trips on a high reactor water l 1evel of +48 inches. High Drywell Pressure greater than +2 psig ! still r e s.a i n s . CHOOSE which of the following will allow restart i of the HPCI Systen.. (More than one answer MAY apply): l
- a. Reactor water level drops below the high reactor water level setpoint of +48 inches.
l
- b. Reactor water level drops below a reset level of 5 inches lower than the high reactor water level setpoint.
l c. Reactor water level drops below a low-low reactor water level of -48 inches.
- d. The reactor operator depresses the Auto Isolation Reset button (23A-S20).
(xxxxx END OF CATEGORY 2 xxxxx)
3A..IHEIBUBENIf.eHD.CONIBOLE Posa 15 QUESTION 3.01 (1.00) Explain the purpose of the Speed Droop Control on an Emergency Diesel Generator Governor. QUESTION 3.02 (1.00) An operator is entering the Compressed Cas Storage Building when he sees the rec warning l i grit 71 a sh i n g a rid h e a r s th e a l a ry KTa'acTi sound at the building entrance. Choose which DNE of the following actions the operator should take:
- a. Continue entering the building and ensure that it has been evacuated.
- b. Notify the control roon, of an area radiation nionitor a l a r ni condition.
- c. Notify Security of a possible unauthorized intrusion into the Conipressed Gas Storage Building.
- d. Notify the control rooni or the fire brigade of a possible fir e in the Con, pressed Gas Storage Building.
1 OUESTION 3.03 (1.50)
- a. The Standby Liquid Control (SLC) Systeni is actuated by use of its 3 position keylocked actuation switch. Describe an automatic action that occurs to another systeni other than the SLC System as a result of this actuation. (0.5) l
- b. Describe how the Standby Liquid Control Tank level as indicated in the control rooni on Panel C05 is measured. (1.0) l l
1 l l i l (ammus CATECORY 3 CONTINUED ON NEXT PAGE manax) l
Is..IHEIBUBENII.6HD.C9HIB9kl Pago 16 QUESTION 3.04 (2.00) During power operation, the three position TIMER TEST switch for the Reactor Manual Control Systesi (RMCS) on Panel C05 is turned to the test position.
- a. Describe the interlock that this switch is designed to test. (1.0)
- b. List two panel indication changes that result when the switch 0tT01 is turned to the test position.
OUESTION 3 05 (1 00) Group 4 rods are being withdrawn during a norsial startup with the Rod Worth Minisiirer in operation. With one rod reniaining in Group 4, the oper ator inadvertently skips a step in the sequence l and withdraws a Group 5 rod. CHOOSE which ONE of the following describes how the Rod Worth Mininiirer will respond upon withdrewal l of this Group 5 rod:
- a. Rod Wor th Mininiirer will apply a rod withdraw block.
- b. Rod Wor th Minin.irer will apply a rod insert block.
- c. Rod Worth Mininizer will apply a rod withdraw block on all rods and a rod insert block on all rods encept the withdrawn Group 5 rod.
I d. Rod Worth Minin.izer will NOT apply any rod blocks. 1 OUESTION 3 06 (1 00) During power operations, the ' CORE SPRAY I N0ZZLE HI DIF PRESS' I annunciator e l a r s,s . Indicate what problem this alert is designed to identify. (Be specific.) l l l l I (nurum CATEGORY 3 CONTINUED ON NEXT PAGE uxxxx) 1
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- 34. 3HEIBU5EHIf_6HQ.G0HIBDLR Pose 17 l 1
QUESTIDH 3.07 (2.00) For each of the following statements regarding the Fire Protection System, indicate whether it is TRUE or FALSE!
- a. Oil admission to the diesel fire pump day tank is controlled by a level-controlled air-operated control valve.
- b. When the screen wash / fire pump transfers to the fire system it locks into this systeni and can be released from fire pump service when no longer required by pushing its reset button located on a control panel in the intake structure.
- c. Following an autoniatic start e a tinier will keep the electric fire pus.p running for 7-1/2 siinutes to prevent on and off cycling of this unit.
- d. For the Cable Spreading Rooni Halon Systeni, sianual actuation of the actuation valve on the halon cylinders or manual actuation of the pressure switch adjacent to the bottles will cause isiniediate dischar ge.
DUESTION 3.08 (1 00) Choose which ONE of the following that will cause a continuous rod drift alars'
- a. Odd nusbered reed switch stuck open and the rod IS NOT being driven.
- b. Rod Position Inforniation Systen Clock sialfunctions.
- c. Odd nusbered reed switch stuck open while the rod IS being driven.
- d. Even nus.bered reed switch stuck open with the rod latched
( at that position. QUESTION 3.09 (2 00) List four cosiponents or equipsient for which the Process Coriputer's Large Motor Monitor Prograsi can provide guidance on startup or restart. (Listing redundant or siniilar components in the sasie system counts as only one answer.) (marx CATEGORY 3 CONTINUED ON NEXT FAGE *****)
31..IHEIBWHEMIS.6HD.GQHIBDLE Pcsa le QUESTION 3.10 (2.00)
, s. Describe two locations frosi which the Reactor Pressure Control System's Mechanical Pressure Regulator setpoint may be set.
- b. State the setpoint range for both the Mechanical Pressure Regulator and the Electrical Pressure Regulator of the Reactor Pressure Control System.
GUESTION 3.11 (2.00) Indicate whether each of the following statenients regarding the Safety Relief Valve (SRV) Low Low Set Systens is TRUE or FALSE:
- a. The Division II low-low set logic will NOT f uriction when both the Alternate Shutdown Systani (ASDS) nia s t e r transfer and SRV Division II transfer switches, located on Panel C292, are in the TRANSFER position.
- b. Manual controls (handswitches) for the low-low set SRVs each have three positions while o.anual controls for all other SRVs have only two positions.
l c. For the SRV to be opened n.anually, both the Division I l handswitch on Panel C03 anu' Division II handswitch on i Panel C292 MUST be placed in the OPEN positions. 1
- d. The low-low set logic will NOT autoniatically open an SRV unless a scrani has occurred.
1 l l l l l l l 1 (xxxxx CATEGORY 3 CONTINUED ON NEXT PAGE xxxxx) 1'
ir.1HEIBUDEHII_6HD CQUIBDLS Pcs* 19 QUESTION 3.12 (2.00) For each of the following Process Radiation Monitoring Systen trips, describe ONE autos,atic action (other than alares) that will
'r e s u l t . (If no automatic actions will result other than alarms, then state 'NONE' . )
- a. HIGH-HICH alarm combined with a loss of sample flow on
' opposite channels of the Stack Gas Monitoring System when operating with the off-gas storage system.
- b. HIGH-HIGH alatni on both channels of the Stack Gas Monitoring Syster when the of f-gas storage system is not being operated.
- c. Downscale trip of Main Steeniline Monitor A conibined with an upscale trip of Main Steanline Monitor B.
- d. Upscale trip of just one of the Reactor Building Exhaust Vent Plenum Monitors.
QUESTION 3.13 (1.00) A spurious Group 2 Isolation occurs resulting in the closure of Drywell Sun,p Floor Drain and Equipment Drain Isolation Valves A0-2541 AED and AO-2561 ARD. (All other subgroups of Group 2 also isolat e . ) The initiating signal is verified to be clear and an attetpt is n ade to first reopen Drywell Floor Drain Sunp l Isolation Valve A0-2541 A. This is done by placing its handswitch I on Panel C04 to the CLOSE position, turning the Group 2 and 3 ISOL VALVE reset switch on Panel C05 to the INBD and then the OUTOD positions, and then placing the Isolation Valve handswitch to the OPEN position. Explain why the valve will fail to reopen when attempted by this siethod. (1.0) OUESTION 3.14 (1.00) During a reactor startup, the reactor operator is increasing recirculation flow in each loop approxis,ately sisiultaneously with the Manual / Auto transfer station s,anual control. Upon reaching 35% recirculation pump speed, the reactor operator puts the individual manual / auto transfer stations for each pusp in AUTO. DESCRIBE the occurrence that can be expected when this l 1s done AND EXPLAIN why it occurs. l l l l (xxxxx CATEGORY 3 CONTINUED ON NEXT PAGE xxxxx) l l
Pago 20
- 34. HfIBW5ENIf 6t!D GDHIBDLE QUESTION 3.15 (2.00)
Answer each of the following in regards to the Rod Block Monitor (RBM) System!
- s. The reference Average Power Range Monitor (APRh) channel for RBM Channel 7 fails downscale. Which APRM channel serves as the alternate source of an APRM reference channel (0.5) to RBM Channel 7?
- b. Upon f ailure of the ref erence AFRh cnannel tb 't,WE7BMMn is the alternate source inipose d ( i . e . , what actions need to be taken, if any, to apply the alternate source?) (0.5)
- c. Balance i t, not achieved during the null sequence between the reference level signal and the RE:M channel output signzl s through autoniatic adjustnient of RE:M aniplifier gain. Other than an RE:M Inop trip and its associated indicatioris, (1.0) describe two indications of this occurrence.
DUESTION 3 16 (1.00) Indicate whether each of the following statenients regarding the Reactor Level Control Systeni is TRUE or FALSE:
- a. The systeni can be f orced to select a f ailed narrow range level signal for control by pressing the OVERRIDE pushbutton while the INVALID coniniand light is on.
- b. The oper stor siay choose which reactor pressure signal is used to cosipensate for reactor water density changes by means of the PRESSURE SELECT pushbutton.
QUESTION 3.17 (1.50) Explain the dif f erence in ternis of equipment operation between each of the three positions (OFF, STANDBY, and ANALYZE) of the Hydrogen-Oxygen Analyzing Systeni switch on control room Panels C259 and C260. N (***** END OF CATEGORY 3 z****) 1 __ . - - _ _ -
ii ..tB Q G E Q U B E R _:_U Q B B e k t _6 B H Q B B e k t _E B E B G E N C Y Pag 9 21 GHQ.86DIQLQGIGek.GQHIBQL GU 4.01 (1.00) g Operations Man .1 3 ' Control Rod Hydraulic Systeme* Section VI.F.1.1 ' Operation Under High Drive Pressurer* describes the method of trying to mov; tuck control rod by elevated drive pressure. This procedure indicates elevated drive pressure may be used to withdraw a drive but not to irt one. Explain why elevated drive pressure should not be used to inse rive. GUESTION 4.02 (2.00) Answer each of the following in regards to inerting the primary containsient during a reactor startup in accordance with Oper ations Manual E:.4.1 ' Primary Containment'!
- a. Explain why the heating boiler seust be in operation with steani available.
- b. Explain why the CONTAINMENT VENT RUN MODE INTERLOCK switch s,ust be in the E:Yp ASS pos i tion.
i l GUESTION 4.03 (2.00) ! Choose the cor rect word in parenthesis in accordance with 4 AWI 4.5.1 'Equipnient Alignnient Verification Methods' for each of the f ollowing statements regarding equipsient alignnient verification
- a. Check that the side rack, screw shutter is fully to the I (rights left) with 'Dise' indicated in the window for a racked out 4 kv breaker.
- b. Check that the switch for a closed 480 V aotor control center breaker is (vertical, horizontal).
- c. When checking sianual valves for closed position, the operator should verify closed by attempting to turn the valve handwheel in the (clockwise, counter clockwise) direction.
- d. When a motor operated valve is (open, closed) the visible portion of the stesi should be saiooth.
l (xxxxx CATEGORY 4 CONTINUED ON NEXT PAGE xxxxx) t
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5I QUESTION 4.04 (1.00) Fill in the blanks with the correct numbers for the following statement in accordance with Monticello Technical Specifications and Operations Manual C.1 "Startup Procedures.' The average rate of reactor coolant temperature change during norsial heatup or enoldown shall not exceed __________ degrees F when averaged over a _ _ _ _ _ _ _ _ _ _- hour per iod .
^
QUESTION 4.05 (1.00) Choose which ONE of the following systenis or equippient does not have its operability requirements DIRECTLY addressed in the Monticello Technical Specifications.
- a. Reactor Water Cleanup S y s t e n. .
- b. Of f gas T r e a tsient Sys teni .
- c. pressure Suppression Chamber-Reactor E:vilding Vacuum E:r e aker s .
- d. Fire protection Systeni Hose Stations.
QUESTION 4.06 (1.00) Choose which of the f ollowing statements explain why post-LOCA operation of the Con.bustible Gas Control Systen. (CGCS) requires the concurrence of the En.ergency Director. (Hore than one answer MAY apply):
- a. Extresely high radiation levels may be produced around systes, piping .
- b. The entire CGCS will be operating at containnient pressure.
- c. The Standby Gas Treatment System sivst be shutdown when the CGCS is operating.
- d. There is an increased risk of a fire starting in the containment when the CGCS is operating.
(szuur CATEGORY 4 CONTINUED ON NEXT PAGE **xxx)
l Pose 23 Sr._tBQGEDUBEE.:!C66 GWR 88010LQE GQHfBQL50BU6Lt'6BH9806Lt EBEBGEUGY QUESTION 4.07 (1.50) In accordance with Operations Manual B.I.4 ' Reactor Recirculation System,' with the recirculation pump at minimum, a scoop tube lock may be reset by reducing the Manual-Auto transfer station output to minimum, waiting five minutes, and then depressing the RESET . button on Panel C-04. EXPLAIN why this five minute waiting period is prescribed following reduction of the Manual-Auto transfer station output to minisium. (Include any possible consequence of not providing the five siinute waiting period and an explanation of why this consequence siay occur . ) oo GUESTION 4.08 (1.56) Choose the correct word in parenthesis for each of the following'
- o. A neutron exposure of one ren, would have a (larger, s e.a l l e r ,
the sase ) biological effect as that of a ganin.a-ray exposure of one ren.. (0.5)
- 6. Due to their low penetrating capability, (alpha, beta, g a n.e a - r a y s , neutrons) are considered to do biological dan. age to the hunian body only when they are ingested or inhaled.
(0.5)
- r. n n e- ogde er t i r e r t ; t. ,,, 70 stad heve ; Li,e 3 e n, e ,
a 02iietest.4. g g q ti:1cgic:1 ef#cct o> inat of v..c :"-ie e# rahd+-^o- ' O .- 5J QUESTION 4.09 (1.00) An individual in the plant calls the control roon, by telephone to report a fire. In his excited state he forgets and cannot adequately describe his location in the plant. Short of sending out search parties, describe how this individusi (and thus the fire) should be located in accordance with Procedure A.3-004
' Fire Fighting Procedures and Strategies.' ( Assusie no fire alarsis are received.)
(xxxxx CATEGORY 4 CONTINUED ON NEXT PAGE xxxxx)
3 Pese 24 Di!!![!lbilggg BueLt Ertsstuct , QUESTION 4.10 (2 00) Safety Relief Valve 'A' is stuck open resulting in leakage Srester than 100,000 lbm/hr. Attempted cycling of the valve has no effect. As a result recirculation pump speed is reduced to minimum and the reactor is manually scrammed. Operations Manual C.4-B.3.3.As "Relief Valve Failure' indicates that under these conditions if the cooldown rate exceeds 100 degrees F/ hour then the followins actions need to be taken!
- a. Reduce Control Rod Drive coolins flow to s inimuni.
- b. Shutdown the Reactor Water Cleanup Systen..
- c. Close the Main Steaniline Isolation valves.
- d. E:reak condenser vacuun..
EXPLAIN WHY each of these four actions (a through d) are taken. ( A specific answer is requir ed f or each iten.. ) (2.0) I l i (zumma CATEGORY 4 CONTINUED ON NEXT PAGE maxxx)
Pcs2 2e St..tB9GEDVBER 650.86910L6 E!G6L.G9HIB9L50Bb6Lt 69HOBbekt EBEBGENGY QUESTION 4.11 (2.50) For each of the following conditions (a through d) indicate which of the corresponding procedures ,(1 through 4) would need to be entered. (More than one procedure may apply for each condition. If none of the procedures are required to be entered then state
'NONE').
CONDITIONS
- a. Drywell pressure has increased to 3.0 psis
- b. Reactor E:vilding High Radiation Annunciator alarms
- c. Torus water level has decreased to -3 0 inches
- d. Drywell Equipn ent Sun.p High Level Annunciator a l a r nis PROCEDURES
- 1. C.5 - 1100 'RPV Control'
- 2. C.5 - 1200 'Pr imar y Containnient Control'
- 3. C.5 - 1300 ' Secondary Containment Control'
- 4. C . 5 - 14 00 ' Radiological Control' OUESTION 4.12 (1.50)
Fill in the blar.ks with the correct nuesbers in each of the following statenients in accordance with 4 AWI-11.1.14
' Radiological Protection Standards' regarding plant administrative hold points and liniitsi
- a. No individual without Forni NRC-4 on file shall be authorized to receive in excess of _______ siten. in one quarter.
- b. No individual with Form NRC-4 on file shall be authorized to receive in excess of _________ sireni in one quarter unless approval of the individual's supervisor and the Superintendent, Radiation Protection have been obtained.
- c. No individual shall be authorized to receive in excess of
___________ mrem in one quarter unless approval from the Plant Manager has been obtained. (arran CATEGORY 4 CONTINUED ON NEXT PAGE xxxxx)
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UEI I OUESTION 4.13 (1.00) A spurious trip of both Reactor Feedpurips with degraded emergency high pressure injection systens causes reactor water level to drop below the low-low level setpoint. Prior to reaching the top of the active fuel, feedwater flow to the vessel is restored and resetor w0ter level stabilizes and slowly begins increasing. The reactor operator notes that the Automatic Depressurization System (ADS) 107 second timer is still counting down and that there is not enough tinie reniaining to restore Icvel to above the low-low level setpoint prior to automatic initiation of ADS. Operations Hanval E:. 3.3 'Resetor Prescure Relief Systeni ' states that the ADS inhibit shocid not be used unless the HPCI, RCIC or reactor feedwater systes.s have returned the reactor level to above the low-low level trip setting. However, Operations Manual C.5-1101 I 'RPV Level Coritrol' prescribes that if RpV water level can be maintained above the top of the active fuel and the ADS tin.cr has initiated ther, prevent autoniatic RPV depressurization by placing the ADS Inhibit switches i n INHIE:IT. The tisie reniainin3 before autos.stic initiation of ADS is insufficient for consultation with other personnel. CHOOSE which ONE of the f ollowing actions should be taken under these circusistances:
- a. Inhibit autoniatic ADS initiation.
- b. Allow ADS to autoniatically initiate.
- c. Anticipate an autoniatic initiation of ADS and therefore sianual l y initiate ADS prior to that tisie as conservative l action.
i
- d. Avoid a f ailure to follow procedure by decreasing f eedwater flow such that reactor water level drops to below the top of the active fuel before ADS autoniatically initiates.
l l (ksman CATEGORY 4 CONTINUED ON NEXT PAGE azzax)
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QChib!hbIlkh 00ESTION 4.14 (1.00) A break in the Fuel Pool Cooling and Cleanup System results in a loss of level in the fuel pool. Operations Manual C.4-B.2 1.A
'Energency Makeup to the Fuel Pool' lists the alternate methods of replacing lost inventory in the order of preference (most desirable to least desirable) as follows!
Filter /Demineralizer Backwash Connection Co~ridetir8W SeWitU StTtTen Fir e Hose Station Explain the r ationale behind this order of preference. (1.0) DUESTION 4.15 (1 50) Fill in the blank for each of the following statements regarding General Safety Tag Control;
- a. The ___________ card shall be used when it is essential that the specified position be maintained to safeguard equiprent or service or for operational reasons.
- b. _______________ cards are printed in blue on white weather-resisting cardboard.
- c. There is no more important safety device than a properly filled out and attached ____________ card.
1 i I e l l l l l l (xxuma CATEGORY 4 CONTINUED ON NEXT PAGE **xxx)
._a - _
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. @! . I OUESTION 4.16 (1 25)
During a reactor startup, each of the following individuals asks permission to be the one to actually withdraw control rods. In each case, the reactor operator believes the individual to be capable and to possess the requisite knowledge required for the j control manipulations. The reactor operator also intends to : meintain direction and his presence during the manipulations. I Under these conditions for each of the following cases, indicate l whether or not that person can be granted persiission to withdraw ! control rods.
- a. A nonlicensed plant engineer who is currently in an SRO training program to qualify for a license pursuant to 10 CFR 55.
- b. A nonlicensed assistant equipsent operator who is scheduled to attend the ne::t R0 license training class. (That class has not yet comn.enced.)
- c. A training instructor whose intent is to activate his inactive RO license status in accordance with 10 CFR 55.
- d. The NRC Senior Resident Inspector.
- e. The Vice President Nuclear Generation who in nonlicensed but very insistent.
QUESTION 4.17 (2.00)
- a. Explain why it is NOT advisable for an Eniergency Diesel Generator to be tied to an offsite power systen, in anticipation of a loss of offsite power (such as during severe weather ) . (1.0)
- b. Explain why prolonged operation of an Eniergency Diesel Generator at zero load (e.g., running but not tied to the bus) is also not advisable. (1.0) l (xxxxx END OF CATEGORY 4 marxx)
(xxxxxxxxxx END OF EXAMINATION xxxxxxxxxx) l l
.~ - .
lr. tBIWCIELEl_9E_UUGLE6B t0 WEB EL6HI 9tEB6IIQUt Pas) 29 IDEB500168BIGEt_UE61_lB6HEEIB_6ND ELUID ELOW l CNSWER 1.01 (1.50)
- a. True (0.5) l
- b. False (0.5)
- c. False (0.5)
REFERENCE Heat Transfer and Fluid Flow Fundamentals, page 8-47. Monticello LER 87-001. 292008K120 293008H131 290002K403 ..(KA's) ANSWER 1.02 (1.50) DeratinS (Continued operation at a lower but constant power level) Coastdown (Load is allowed to drop while oper ating with all rods out) Feedwater Tenperature Reduction (Results in decreased plant ef ficiency but allows siaintenance of full turbine load provided licensed reactor power is not exceeded) Excess Core Flow (Reaching 100% power on less than 100% flow control line by enceeding 100% cove flow) ( Any 3 at 0 5 pts each) l REFERENCE l Reactor Theory, page 7-20. 292002K114 292002K111 292002K109 ..(KA's) ANSWER 1.03 (1.00) l As reactor water level is lowered the natural cirevlation l driving head is reduced (0.25) resulting in a reduced core flow
- (0.25) and cor responding increase in void fraction (0.25). The l negative reactivity froni the increased void fraction causes a decrease in reactor power (0.25).
REFERENCE C.S.1-2007 Emergency Operating Procedure Basis, page 10. 295031K103 ..(KA's) (xxxxx CATEGORY 1 CONTINUED ON NEXT PAGE xxxxx)
'i'--Illlillills!Inffikit!rilllicitali:tEl!811286, " **
6 1 CNSWER 1.04 (2.50)
- a. increases (0.5)
- b. higher (0.5) decreased (0.5)
- c. increases (0.5) more (0.5)
REFERENCE Heat Tiansfer and Fluid Flow Fundamentals, pages 6-81, 9-36. 293008K119 293006K110 293006H108 ..(KA's) ANSWER 1 05 (2.50)
- a. decrease (0.5) smaller (0.5) higher (0.5)
- b. below (0.5)
- c. higher (0.5)
REFERENCE C.5.1-1000 Es.ergency Operating procedur e E: asis , page 11. Heat Transfer and Fluid Flow Fundacientals, page B-56. 291002K108 291002K107 ..(KA's) ANSWER 1.06 (1.00) Changing Power with recirculation flow changes total core power while keeping the flux profile relatively 'anchanged OR to avoid
- (1.0) localized flun peaking.
(Either answer acceptible for full credit.) REFERENCE Reactor Theory, pages 7-13, 7-17, 7-18. 292008K119 292008K120 292008K118 ..(MA's' (maxxx CATECORY 1 CONTINDED ON NEXT PAGE unamm) b_ _ _ _ _ _ _ _ _ _ _
- ~~
21--flifillills!!i!!st!!!rilllici!!sI:l!818!!?le
-?
ANSWER 1.07 (1.50) The downpower transient should take longer to stabilize. (0.5) On a downpower transient, the rate of power change is limited to the rate of decay of the longest lived neutron precursor. (1.0) REFERENCE Reactor Theory, yese 3-29. 292003K103 292003K106 ..(KA's) ANSWER 1.08 (1.50)
- x. 2 (0.5), 3 (0.5)
- y. 1 (0.5)
REFERENCE Reactor Theory, page 5-25. 292005K112 292005K111 ..(KA's) ANSWER 1.09 (2.00) Void Coef ficient (0.3) Increases Power (0.2) Moder ator Tee:per atur e Coef ficient (0.3) Increases Power (0.2) Doppler Coefficient (0.3) Decreases Power (0 2) Void Coefficient (0.3) Decreases Power (0 2) REFERENCE Reactor Theory e pages + 46 through 4-49. 292004K110 292004K105 292004K101 ..(KA's) l ANSWER 1.10 (1 50)
- a. 2 (0.5)
- b. 2 (0.5)
- c. 1 (0.5) l (smans CATEGORY 1 CONTINUED ON NEXT PAGE xxxxx) l
4?--fillilit!!stEs!!iki!!rilllicittli:2!Ill!!!1e REFERENCE Heat Transfer and Fluid Flow Fundamentals, page 9-36. 292005K105 292004K102 292004K109 292004K113 ..(KA's) ANSWER 1.11 (1.00) a (1 0) RtrERENCE Reactor Theory, pages 6-6 through 6-13. 292006H107 ..(KA's) ANSWER 1.12 (2.00) A r eactivity anon.aly is the reactivity equivalent of the dif ference between actual critical rod configuration and expected rod configuration during steady state reactor power operation. (1.0) (E::act wording is not required.) l - Loss of control blade boron
- Crud buildup on the cladding surface
- Channel bowing Core thers.a1 evaluation errors
- Fuel loading errors Other correct answers will be accepted
( Any 2 of 5 at 0.5 pts each) REFERENCE Reactor Theor y , page 7-24. Technical Specification 3.3.E. I 201003G005 292002K111 ..(KA's) l I ANSWER 1.13 (1.00) l Increases (0.5) due to fuel burnup (0.5) (usxxx CATEGORY 1 CONTINUED ON NEXT PAGE usamm)
'e It_.tBIWGIELER 0E.UUGLE6B EQWEB.tL6HI_9tEB6139W1 Pa89 33 INE85901665IGEz_UE6I_fB6HEEEB 6HD.ELul9 ELQW
~
REFERENCE Reactor Theory, page 1-35. 292002K110 292007K103 292002K114 ..(KA's) ANSWER 1.14 (2.00)
- a. Increase (0 5)
- b. Decrease (0 5)
~
- c. ve c r ea s e ( v f o susws K s ei v s . cert ag e car *T) (o.d;
- d. Decrease (0.5)
REFERENCE Reactor Theory, page 5-13. 292005K109 ..(KA's) ANSWER 1 15 (1.50) b (0,5) The^e is a larger % change in water volunie f or the sanie increase OR the voids produced at 70% VF have a larger effect on core reactivity since they are in an area of higher neutron flu::. (1.0) (Either acceptable for full credit.) REFERENCE Reactor Theory, page 4-19. 292004K111 ..(KA's) ANSWER 1.16 (1 50)
- a. (1) HFLPD (0.5)
(2) MAPRAT (0 5)
- b. higher (0.5)
REFERENCE Heat Transfer and Fluid ' low Fundanientals, pages 9-15, 9-56. 215005A107 293009K111 293009K109 293009K107 ..(KA's) (maxxx END OF CATEGORY 1 xxamm)
4 rese 34 Ei _gyERIGH_INGW9ING_B6EEII_6HD ENEBGENGI ANSWER 2.01 (1.50) 3: sy n I.vi . i. m r. ' ^ 3 ? ; r, ;;T C I n . . l i r.: "i? rl:t '^ 2: L te th: 7.us wnss nvaiiaery vil-Peep :p:rrtirr ;ege;; th; hyd..wiawelly riert:d t=5in: 5'ar and t' a ni c a ' / 1 . 1,o move i, - th;is ivily q ;,, rv.n a u .. ; 0 . Tr. T- A c. nmp g tw t r 4.v .y 6%.Ak 4 km As4 4.bt. 54*r v AWE a M b* Ntam a dm,*15th napply vatve. e, w(3, REFERENCE Monticello LER U7-007. Oper ations Ma nual E:.3.2, Noveniber 27, 1985, Page 5. Operations Manual E:.3.2, Noveniber 23, 1985, Page 35. 206000A203 206000A106 206000K402 ..(HA's) ANSWER 2.02 (1.50) The flow contr ol valve approaches the closed position (0.5) due to the flow control station flow elenient being located upstrear. of the charging water circuit causing it to provide a high flow signal. (1.0)
- REFERENCE l
Oper ations Manual E: .1. 3 , October 27, 1983, page 34. 201001A204 201001K401 ..(KA's) t I ANSWER 2.03 (2 00) Protective trip circuits will be inoperable. Scoop tube lock.s in position. Control power to the drive siotor breaker is lost. A and E: auxiliary lube oil pusips are inoperative. l The en,ergency D.C. Ivbe oil punip will lose its auto start capability on low lobe oil pressure. A loop flow drops to zero as the MG set fluid coupling is lost. Danisse to the drive siotor will result without either prosipt l restoration of oil flow or pronipt shutdown of the drive motor. l (4 of 7 at 0.5 pts. each). 1 I REFERENCE Oper ations Manual C.4-E:. 9.10. A , Revision 0, Page 4. 263000K303 263000K201 ..(KA's) l L (msman CATEGORY 2 CONTINUED ON NEXT PAGE m*xxx) 3 __ _
. I 21.. EIIH.lHGWQING.HEEII.6HD.EBEBGENCY Pcse 35 CNSWER 2 04 (1.00)
T _ e r-tirl primary containsient Group.344- isolation associated with an auto initiation will not occur when the TEST pushbutton is used (1.0) REFERENCE Operations Manual B.4.2, January 29, 1987, Page 12. z61 uovt;1oy . . t r;n s i ANSWER 2.05 (1.00) e, d (2 at 0.5 pt s . each - 0.5 pts. also subtracted for each incorrect answer up to total value of question.) REFERENCE Oper ations Mz nuel E: . 4 . 2 , January 29, 1987, Page 3. 290001G004 290001G001 290001K401 ..(MA's) ANSWER 2 06 (2.00) l a. High Tetperature in the Control Rod Drive Area (0.75) l of 150 degrees F (0.25) (Setpoint nivst be within 10% for I credit. ) 1
- b. Shutdown one of the RHR pusips that is interlocked with that f an. (1.0)
REFERENCE
- Operations Manual E:. 4 1 , March 27, 1986, Pages 26, 31.
295024C006 295024A114 295024K218 295028A102 ..(KA's) l i 6 ) (amman CATEGORY 2 CONTINUED ON NEXT PAGE *****) 1,
- 22. EllGH.IHEWPINE.96EEII.6HQ.EBEBEEHEY Pese 36 ANSWER 2.07 (2.00)
Feedwater Regulator and Bypass Valves Feedpump Seal Systes. Scram Valves CRD Flow Control Valves MSIVs Reactor Auto-Pressure Relief Valves Breathing Air Eondensate Derrine r alize r Byye>> Velve (4 at 0.5 pts. each) REFERENCE Oper ations Ma nual E:.8.4.1, June 19, 1986, Page 2. 295019A202 ..(KA's) ANSWER 2.08 (2.00)
- a. RHR Service Water Systeni (0.5)
- b. Seal Water Systena (0.5)
- c. Service Water Systeni (0.5)
- d. Den,iner a li ed Water Systeni 04 4, Ls % 4eSt..A w k (0.5)
REFERENCE 8.5.63 Operations Manuali April 7, 1983, Pages 2, 3. 205000G007 205000K608 205000K105 205000H115 295018G006
..(KA's)
ANSWER 2 09 (1.00) Drop in inlei pressure (0.5) will cause a turbine runback 40.5). O R Aw%k Ska e4 of 4he Mmibyfamp (o . SD REFERENCE (t". % e e o u.. M (* A + c ,t ai s ) Operations Manual E:.6.2.4, February 26, 1987, Page 18. 245000K605 245000K406 241000A210 241000K616 ..(KA's) , (ammum CATEGORY 2 CONTINUED ON NEXT PAGE mauxx)
--t _ _ y.. . _ . . yy__ _r.
- e. s i
l 2I..tL6HI.QERISM.INGLUDING86EEII6HQ.EMEBGEUGY Posa 37 I EXEIEUR l f . ANSWER 2.10 (2.00) a.1. Open (0.5)
- 2. Closed (0.5) b.1. As is (0.5)
- 2. As is (0.5)
REFERENCE Oper ations Ma nual E:.5.7, June 19, 1986, Page 5. 259001K603 259001K601 259001K106 ..(KA's) ANSWER 2.11 (1.00) The operating punip is tripped (0.5) and any further action is prevented until reactor pressure decreases to 900 psis. (0.5) REFERENCE Oper ations Manual E: . 3 . 4 , December 11, 1986, Page 17. 203000K411 203000K105 ..(KA's) ANSWER 2.12 (1.00) ' The static switch will transfer the load to the alternate source long enough to clear the fault (0.5) (blow the fuse in the fuse-di sconnecte4- s wi t ch ) . When the fault has cleared the static switch will then transfer the load back to the inverter. (0.5) REFEPENCE Oper ations Manual E:.9.13, April 30, 1987, Page 38. 262002K401 262002K601 262002A201 ..(KA's) ANSWER 2.13 12.00) RWCU System Isolation (0 5) on high non-regenerative heat exchanger outlet temperature. (0.5) RWCU pumps trip (0.5) on high pump bearing cooling water t e sipe r a tur e (0.5) I (unxxx CATEGORY 2 CONTINUED ON HEX) PAGE xxxxx)
A.. ESIGN.INGLUDING.86EEII 68Q.EBEBGENGI Pcs2 as REFERENCE Operations Manual C.4-B.2.5.Ae Revision De Page 2 204000A201 204000K601 204000K404 204000K403 204000K104
..(KA's)
ANSWER 2.14 (1.00) Reactor Scram (0.5) due to condenser low vacousi (0.5). REFERENCE Operations Ma nual E:.7.2, January 6, 1987, Page 196.2. 295002K201 271000K301 271000K110 ..(KA's) ANSWER 2.15 (2.00) Contr el Rod Dr ive Systeni- (0.5) Stendby Liquid Control Tank (0.5) Reector Water Cleanup Sys teni- (0.5) Cleanup Precoat Tank (in the Radwaste E:vilding) (0.5) REFERENCE Operations Henval C.5-3002, Revision 1, Page 5. Oper ations Hz nual C.5-3003, Revision 1, Page 6. 211000C007 211000A201 211000A109 211000K105 ..(KA's) I l i ANSWER 2 16 (1400) Various systen.s siay not be available as required, causing a Technical Specification violation. (0.5) l Pressuri:stion could raise vessel pressure /teciperature l above the Residual Heat Resioval Systeni isolation i interlocks ( (0.5) l ( oW co s an t k awwe-s act o c c v f a A /t. J. REFERENCE l I l Heat Transfer and Fluid Flow Funda4entals, Page 8-55. 295021K203 295021G003 295021K101 ..(KA's) l l l t (xxxxx CATECORY 2 CONTINUED ON NEXT PAGE xxxxx) L .
+fm.:qv-h__tL991_QEEIGM.IWGLUDING 56EEII eHD EBERGENGI rosa 39
. RIEIE55
. - ;j ,
ANSWER 2.17 (1.00) c (1.0) (Will subtract 0 5 pts for each incorrect 2nswer up to total value of question. ) REFERENCE Operations Manual B.3 2, November 27, 1985, page 4. 206t00M407 206000K40S 206000H401 . . M(A ' - ) l l l l l i i l l l l a d l (mmmmm END OF CATECORY 2 summa) h
32..XHEIBU5EMIf eHR.CONIBQLH Pog2 40 CNSHER 3.01 (1.00) Controls the magnitude of the speed decrease as load is increased and thus diesel generator responsiveness anti stability. (1 0) (Alternate wording is acceptable) REFERENCE Operations Manual B.9.8, August 23, 1985, Pages 4, 62. 269000nque 264000n403 ..tKA;s) ANSWER 3 02 (1.00) b (1.0.) REFERENCE Operations Hanval E: . 5 .12 , January 16, 1984, Page 26. 272000G008 ..(KA's) ANSWER 3.03 (1.50)
- a. Reettor Water Cleanup Systen Isolation. (0.5)
- b. Measured by injection of a continuous stresci of i n s t r u n.e n t air from a purge controller through a dip tube and the transe,itted backpressure signal fron the dip tube. (1.0)
REFERENCE Operations Manual, March 28, 1985, Pages 4, 5. 211000K506 211000K105 204000K108 ..(MA's) l l l l (axxxx CATEGORY 3 CONTINUED ON NEXT PAGE xxxxx) I
l
. . . \
. l 32__IHRIBWBENIE_6HQ.CONIBDLI Pcso 41 l 1
.~
ANSWER 3.04 (2.00)
- a. Disconnect of the directional control solenoid valve of the selected rod from the drive control buses of the RMCS in the event of a rod drive timer switch failure. (1.0)
- b. 'TWO SEC DELAY' indicating light (on Panel C05) illvainates SELECT BLOCK indicating light (on Panel C05) illuminates (when the interlock operates)
Associated rod select light on the full core display is extin3013hed. (Any 2 of 3 at 0.5 each) REFERENCE Oper ations Manual E:.5.5, January 6, 1987, Page 4. 201002K401 201002K301 201002G007 294001A113 ..(KA's) ANSWER 3.05 (1 00) ga 04. b (1 0) E.'%ee o t t < r is. L l" /sv L (I o~ d s '4 REFERENCE Operations Hznual E: . 5 2 , March 27, 1986, Page 12. 201006K407 201006K402 201006K401 201006K101 ..(KA's) ANSWER 3.06 (1.00) Break of the core spray piping internal to 'he reactor vessel and outside the shroud. (1 0) REFERENCE Operations Manual E:.3.1, January 8, 1987, Pages 8, 16. 200001G008 209001A205 209001K404 ..(KA's) ANSWER 3.07 (2 00)
- a. False (0 5)
- b. True (0.5)
- c. True (0.5)
- d. False (0.5)
(sumur CATEGORY 3 CONTINUED ON HEXT PAGE xxxxx)
ha__INSIBUBENIIeNDCONIB9kB Pc82 42 REFERENCE Operations Manual B.B.11e June 19, 1986, PaBes 2, 8, 21. 286000K402 286000G007 286000A406 286000A212 286000K505
..(KA's)
ANSWER 3.08 (1.00)
- d. (1.0)
REFERENCE Operations Ma nual B.5 4, April 30, 1987, Page 9. 214000A201 201002A303 201002K403 ..(MA's) ANSWER 3.09 (2.00) Reactor Fe e d p u nip . Circulating Water P u e: P . l Condensate Pus,p. ! Recirculatior. Punip M-G Set. Cooling Tower P u nip . (4 of 5 at 0.5 each) REFERENCE l Oper atioris Ma rival B.5.10, Dececiber 30, 1986, Page 11. 256000A305 259001G008 259001A402 259001A309 294001A115
..(KA's) l ANSWER 3.10 (2 00) l a. C orit r ol Rooni Panel (Panel C-07) (0.5)
Turbine Front Standard (0.5)
- b. Mechanical Pressure Regulator -
150 psis to 1050 psig (0 5) l Electrical Pressure Regulator - 900 psis to 1000 psig (0.5) REFERENCE Operations Manual B.5.9, May 2, 1975, Pages 2, 3. 241000G007 241000K401 ..(KA's) (usaxx CATEGORY 3 CONTINUED ON NEXT PAGE xxxxx) i '
12__INSIBUDENII.6HD.GQUIBOLI P:s2 43 ANSWER 3.11 (2.00)
- a. False (0 5)
- b. True (0.5)
- c. False (0.5)
- d. True (0.5)
REFERENCE Operations Manual B.3.3, June 12, 1986, Pages 12, 13. 239002G007 239002K409 239002K40s zav002H401 ..(na si ANSWER 3.12 (2.00)
- a. Stack isolation (closure of SV-1928, SV-2353, SV-7677) (0.5)
- b. .NettE- $4e t k. .T.To l d * h (0.5)
- c. NONE (0.5)
- d. Reactor Building Ventilation Systes, shutdown Standby Gas Treatnient Systes, initiation.
Priniary Containnient Purge and Ventilation Valves Close (1 of 3 required at 0.5 pts.) REFERENCE Oper ations Mer.ual B.5.11, August 19, 1982, Pages 7, 32, 39. Operations Manual B.5 11, July 1, 1986, Page 33. 272000K402 272000H109 272000K106 272000K103 ..(KA's) ANSWER 3.13 (1.00) l It is necessary to place all the valve handswitches for the valve subgroup associated with that valve to the CLOSE position to energize the reset interlock. (1.0) REFERENCE . Operations Manual B.7.1, January 9, 1986, Page 26. Operations Manual B.5 6, June 5, 1986, Page 34. 223002G009 223002A403 223002K406 223001K104 ..(KA's) l l (xxxxx CATEGORY 3 CONTINUED ON NEXT PAGE xxxxx) l
hi..IHEIBUNEHIE.6HR.GDHIBDLE Pcge 44 CNSWER 3 14 (1.00) Automatic increase of pump speeds to 45% (0.5) due to minimum speed output of the dual limiter (0.5). REFERENCE Operations Manual B.1.4, January 29, 1987, Page 61. 202002A101 202002K407 202002K402 ..(NA's) ANSWER 3.15 (2.00)
- a. Channel 2 (0.5)
- b. Bypass the failed APRM channel (0.5)
- c. ' Rod Out Inhibit' light on RBM cabinet remains i l l unii n a t e d (0.5)
'NO BALANCE' light on RE:M cabinet t e r.ie i n s illumi na ted (0.5)
REFERENCE Oper ations Ma nual D.5.1.2, Dec eniber 6, 1984, Pages 23, 45. 215002G008 215002A306 215002A305 215002K301 215002H101
..(MA's)
ANSWER 3.16 (1 00)
- a. True (0.5)
- b. False (0.5)
REFERENCE Operations Manual E. 5.7, June 19, 1986, Page 9. 259002A203 259002K414 ..(MA's) i (xxxxx CATEGORY 3 CONTINUED ON NEXT PAGE xxxxx) b -
Pogo 45 32..IHEIBUBEHIE_eHD.G9HIBDLE CNSWER 3.17 (1.50) 0FF Analyzer is de-energized. (0.5) STANDBY Sample compartment (hotbox) heater is energized but both the analyzer electronics and sample pump remain off (0.5) ANnLYZE Electronics are energized and the sample pump starts (0.5) REFERENCE Oper ations Mariual E:.4.3, October 31, 1984, Page 2. 223001G007 223001A405 223001A404 ..(KA's) ? (unxxx END OF CATEGORY 3 xxxxx) i -_ _ . - .____ _-
'"'"S' '
--Il8!!!HillairtlE!Ibiril2'"""
AS 4.01 (1.00) May lead to icking of the directional control valves. (1 0) REFERENCE Operations Manual B.1.3, Octobe e 1983, Page 72 Operations Manual C.4-B.1.3.B Revis 1, Page 5 201001G010 201001K408 201001K11 ..(KA's) ANSWER 4.02 (2.00) l
- a. The nitrogen purge vaporizer utilizes steae to convert liquid nitrogen to gaseous nitrogen. (1.0) i
- b. Various velves that ciust be opened for inerting are interlocked closed with the Reactor Mode Switch in RUN. The CONTAINMENT RUN MODE INTERLOCK switch bypasses that i interlock. (1.0)
REFERENCE Oper ations Ma nual E:.4.1, March 27, 1986, Page 35, 37, 95. 22301G007 223001K404 223001K103 ..(KA's) l ANSWER 4.03 (2.00)
- a. Right (0.5)
- b. Vertical (0.5)
- c. Clockwise (0.5)
- d. Open (0 5)
RE F ERD'CE 4 AWI-4.5.1, Page 2, 3. 262001A401 262001G001 294001K107 294001K101 ..(KA's) ANSWER 4.04 (1.00) 100 (0 5) One (0.5)
; (maxxx CATEGORY 4 CONTINUED ON HEXT PAGE xxxxx)
$2..tBQGEDWBES.: UQBU6Lt.68HQB56kt.EBEBGENGY Pose 47 699.86030LDEIG6L.GQHIBQL REFERENCE Monticello Technical Specifications, Page 121.
Operations Manual C.1, Septesioer 10, 1986, Page 29. 290002A204 290002K505 ..(KA's) ANSWER 4.05 (1.00) a (1.0) REFERENCE Monticello Technical Specifications, Pages 46a, 198be 226. 286000G011 223001G011 271000G011 204000G011 ..(KA's) ANSWER 4.06 (1.00) j a, b (2 at 0.5 pts. each - will subtract 0 5 pts. for each incor rect answer given up to total value of question) l REFERENCE 1 Operations Henval E:.4.3, Novenber 26, 1986, Page 49. 223001G010 223001K404 ..(KA's) ANSWER 4.07 (1.50) Integration of a positive error signal (caused by the speed deniand i being greater than actual pusip speed) say have caused the controller to saturate high. (0.5) Without waiting for a return to an unsaturated condition, depressing the RESET button could result in an increased speed deniand (0.5) and corresponding scoop tube lockup due to the control signal s,isniatch circuit. (0.5) REFERENCE l l Operations Manual E:.1.4, Septeniber 25, 1986, Page 53. l 202002G010 202002A205 202002A101 ..(KA's) 1 (uxxxx CATEGORY 4 CONTINUED ON NEXT PAGE mamma)
.o ,
" resa es OCIISO.8h8IgggDBuett_EntBGiacy ao ANSWER 4.08 (1c594 (0.5)
- a. The same (0.5)
- b. AIPhas t o CPE(6M e- ni++erent REFERENCE 4 AWI-11 1.14, Revision le Page 2.
zf40UTK1h 29400iM 03 .. ^KA's) ANSWER 4.09 (1.00) Tracing the phone extension. (1.0) REFERENCE Procedure A.3-004, Revision 0, Page 2. 294001A104 286000G001 294001K116 294001A116 ..(KA's) l l ANSWER 4.10 (2.00)
- a. To lisiit stratification in the bottom head region. (0.5)
I
- b. To stop injection of relatively cold water and reduce s t e a s, u s e . (0.5)
- c. To reduce heat loss. (0.5)
- d. To prevent cold air being drawn across the turbine shaft on the loss of sealing steam. (0.5) l l REFERENCE Ope r a tions Ma nual C . 4. -E: . 3. 3. A , Revision 0, Page 8.
256000K310 201001A211 239002A203 239001K604 239001K306
..(KA's) l l
l 1 (urzum CATEGORY 4 CONTINUED ON NEXT PAGE xxxxx) L
9 St..tBOGEQUBEl :.WQBuekt.6BHQBB6Lt.EBEBGEUGY Pcs2 49
, eHQ.8601066EIG6L.GQUIBQL CNSWER 4.11 (2.50)
- a. 1 (0.5) 2 (0 5)
- b. 3 (0.5)
- c. HONE (0.5)
J. NONE '0:5) REFERENCE Operations Manual C.5 - 1000, Revision in Page 7. 295036G011 295030G011 295033G011 295024G011 ..(P9's) ANSWER 4.12 (1.50)
- a. 1000 (0.5)
- b. 2000 (0.5)
- c. 2500 (0.5)
REFERENCE 4AWI - 11.1 14, Revision 1, Page 7. 294001K104 294001K103 ..(KA's) ANSWER 4.13 (1.00) a (1 0) REFERENCE Operations Manual B.3.3, Detober 9, 1986, Page 38. Operations Hanval C.5-1101, Revision 1, Page 7. Operations Manual C.4-B.5.7, Revision le Page 4. Operations Manual C.5-1000, Revision 1, Page 4. 295031K301 295031K208 295031G012 294001A102 , ..(KA's) (xxxxx CATEGORY 4 CONTINUED ON NEXT PAGE xmmru)
~~ ~
. 6 . 65 - 1 ANSWER 4.14 (1.00)
Cleanest sources of water are tried first. (1.0) REFERENCE Operations Manual C.4-B.2.1.A, Revision 0 Page 4. 295023G007 295023A202 ..(KA's) ANSWER 4.15 (1.50)
- a. Secure (0 5)
- b. Unsafe (0.5)
- c. Hold (0.5)
REFERENCE I 4 ACD-4.5, Revision 6, Pages 4 and 5. l 294001M102 ..(MA's) ANSWER 4.16 (1.25)
- a. Yes (0.25)
- b. No (0.25)
- c. Yes (0.25)
- d. No (0.25)
- e. No (0.25) l l
l REFERENCE l 4 ACD-4.7, Revision 11, Page 16. l 10 CFR 55. 294001A109 201003G001 ..(KA's) l l ANSWER 4.17 (2.00)
- a. If a loss of power occurs, the diesel senerator will alsiost certaanly trip out on overload. (1.0)
- b. Will result in carbon buildup. (1.0)
(xxxxx CATEGORY 4 CONTINUED ON HEXT PAGE xxxxx) L _ _
DBH t 6 HQBH6Lt.E9EBGENCY Pese 51
$a.t9Gk0 6 9 e 19LQ BEE.i!6L_hhMIB!k REFERENCE Procedure A.6, January 6e 1987 Page 2.
Operations Manual B.9.8, February 2, 1987, Page 51.1.
'264000K101 262001K302 262001K101 262001K602 ..(KA's) l l
l l (xxxxx END OF CATEGORY 4 xxxxx) (xxxxxxxxxx END OF EXAMINATION xxxxxxxxxx) _ _.-__. . _ - _- , _ , _ _ _ . - _ _ . _ _ _ _ _ _ - _}}