ML20058P297
| ML20058P297 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/27/1990 |
| From: | Miller L, Royack M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20058P294 | List: |
| References | |
| 50-361-OL-90-02, 50-361-OL-90-2, NUDOCS 9008160190 | |
| Download: ML20058P297 (242) | |
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Examination Report No.: 50-361/0L-90-02 v u ; Facility Licensee: San Onofre Nuclear Generating. Station Units 2 and 3 facility Docket No.: 50-361'and 50-362 Facility License No.: ' NPF-10 and NPF-15 : Examination administered at San Onofre Nuclear Generating Station Units'2 and 3, . San Clemente California.
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L Chief Examiner: _e ,
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' Michael J. Moyack '
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- Lice sing Examiner s s ,
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. Approved: .
.. Date 7/27/9D Signed Nwis F. Miller . Jr. . L/
Chief, Operations Sec tion-i
. Summary: .
Examinations administered from June 18 through June 29,'1990. ; Written and operating examinations were a'dministered to one-Senior Reactor Operator Instant (SR01) and nineteen Reactor Operator (RO) applicants. The i SROI and fifteen of the nineteen R0 candidates passed the examinations.- All. of the other candidates failed the examination. . i
'The.failureratefortheReactorOperatorlicensecandidateswas21' percent. l This failure rate is higher than that whichLis normally experienced at San Onofre Units 2 and 3. _1his high failure. rate should be evaluated by - t the facility to assess its root cause and develop appropr.iate corrective action'- -
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i Enclosure 1 f REPORT DETAILS
- 1. NRC and Examiners
- M. Royack, NRC ** i G. Johnston, NRC T. Sundsmo. NRC ** ,
C. Caldwell, NRC SRI ** R. Pugh, PNL
- Chief Examiner
** Persons attending exit meeting.
- 2. Licensee Persons Contacted R. Krieger, Operations Manager i V. Fischer, Operations Units 2/3 !
R. 6ette, Operations Training Manager D. Brevig, Onsite Nuclear Licensing M. Speer, Onsite Nuclear Licensing D. R. Miller, Nuclear Training Instructor D. L. Miller, Compliance K. Rauch, Nuclear Training Instructor R. Hall, Nuclear Training Instructor r All of the above licensee personnel attended the Exit Meeting.
- 3. Examination Findings During the preparation and administration of the written and operating examinations conducted at SONGS Units 2 and 3 the examiner noted specific t concerns in the following areas:
- a. simulator initial conditions (ICs);
- b. procedures;
- c. candidate areas of wea ness; and
- d. licensee provided reference material,
- a. Simulator 1
The site specific simulator appeared to operate satisfactorily during the administration of the simulator portion of the examination. It was noted, during the conduct of the simulator portion of the operating examination, that specific initial conditions l (ICs) that were selected and prevalidated, by both the licensee: l representatives and NRC examiners, had malfunctions added between the time of simulator scenario validation and the administration l
t i I I of the examination. The malfunctions appeared to have been i added to the ICs for other scenario development, training, or l requalification program enhancements. The additional malfunctions, j added to the ICs, enhanced the scenarios provided by the NRC : examiners, however, they were unexpected, and therefore, more t difficult to evaluate. Examples of these were: an anticipated transient without a scram (ATWS), and a main steam, isolation signal ; (MSIS)onalossofainverter. The Chief Examiner reminded licensee : representatives that NRC scenarios should not be modified without ! NRC approval. [ Simulator modeling and hardware problems are detailed in Enclosure 4 ,
- b. Procedures ,
During the preparation and administration of operating and written i examinations the examiners noted some operators had difficulty in l the use and execution of specific procedures. Procedures of j particular concern were the Functional Recovery Procedure, 5023-12-9, and the Power Operations Instruction, 5023-5-1.7, Attachment 15 (Power Maneuvering Boration/ Dilution Guidelines). Also, procedures l requiring a rapid power reduction did not provide the operators with directions for a rapid power reduction. Functional Recovery Procedure The operators appeared to have difficulty and concerns in executing the functional Recovery Procedure (FRP) during , i multiple event casualties. This condition became apparent
- during a simulator scenario which included the following ,
I conditions; ' Er.d of core life, High(approachingTechnicalSpecificationlimits) ; reactor coolant system activity. ; Reactor Trip upon which two Control Element Assemblies , (CEAs) failed to insert. Steam generator (S/G) tube leak which increased to a rupture after the reactor trip. Main Steam isolation valve failure to close, stuck open. ; The scenario conditions required that operators enter the FRP based on the failure of two CEAs to insert and the indications , of a S/G tube leak (high radiation levels in the secondary system). , Failure of two CEAs to insert on a reactor trip alone requires i the operator to enter the FRP. Entry into the FRP required the 1 operators to readdress the steps for Reactivity Control, Vital i Auxiliaries, Reactor Coolant System Inventory, Pressurizer Pressure, and Heat Removal prior to entering functional recovery status check Containment Isolation CI-2. Cl-2 is the first i required. step that addresses the isolation of the affected S/G. l 1 i
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i The elapsed time it took the operators to address the isolation ' of the affected S/G was greater than twenty minutes. During the . twenty minute interval there was a potential for a radioactive ! leak to the environment from either the condenser off gas or the ! possibility of a S/G safety valve lifting. The elapsed time that was required to address the isolation of the affected S/G appeared to be excessive since the potential for a radioactive release to the environment existed. The operators interviewed, with respect to the use of the FRP, ! indicate a hesitancy to enter the FRP due to the inability to address two conditions simultaneously, as in the event with two CCAs sttck out of the core on a reactor trip and a simultaneous , steam generator tube rupture. The operators considered that the FRP was overly restrictive in this area. The examiner concluded that the licensee should evaluate allowing i operators the option of entering a Emergency Operating Instruction in parallel with the required entry to the FRP during multiple event situations. Power Operations i o During the preparation and the administration of the operating examination, bot? simulator and walk through, the 1 candidates appeared to struggle with the preparation of Attachment 15, Power Maneuvering Boration/ Dilution Guidelines, to 5023-5-1.7, Power Operations. The candidates seldom referenced Attachment 15 during the conduct of the simulator . examination. The candidates calculated boration and dilutions using a batch method that was in place prior'to the issuance of Attachment 15. When the examiners questioned the candidates and licensed operator: regarding the use of Attachment 15, the response was that the attachment was complicated and cumbersome, i and that it was intended that an Engineering group would fill ' out the attachment for the operators whenever a power change l was to be executed.
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Presently arocedure 5023-5-1.7, Power Operations, does not i indicate tlat an Engineering group is responsible for the j preparation of the attachment. i The responsibility for completion of and training for the use of Attachment 15 to S023-5-1.7 should be addressed by the licensee. Failure to use this Attachment in the plant, for power manipulations, would be considered a violation of regulatory l requirements. o SONGS 2/3 operating instructions appeared in complete in the area of rapid power reductions. Specifically, the operator actions to be performed to conduct a rapid i power decrease when required by-Technical Specification l i LCOs, or loss of secondary plant equipment were not clearly
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specified. While this level of detail is not required, it I would enhance and standardize operator performance in this area, and reduce the likelihood of unsafe operator actions. ! The licensee should include instructions for rapid power I reductions. There should address the approved methods for i turbine load reduction, calculations for boration and control ' element assembly (CEA) insertion, i
- c. Candidate Areas of Weakness
- During the administration of the walk through portion of the operating examinations, the examiners noted that the candidates {
i exhibited weakness in three specific areas. l' The first area of weakness was the candidates' general knowledge of San Onofre administrative procedures, specifically in responsibilities and execution of the SONGS Emergency Response , Procedures (Emergency Plan), and general tagging and clearance procedures. The weakness in the area of administrative topics was also documented in the results of the written examination. The second area of weakness noted was in the knowledge and operation of process and area radiation monitoring equipment. . Examples of weak areas were performance of operational checks, and locations of detectors. The third area of weakness was in locating equipment controls and indications for off normal, or infrequently used, equipment. i Examples of controls and instrumentation location problems were, l ! fill and drain valves for the safety injection tank (SIT) fill i operations, and locations of fuse holder locations for the SIT vent valves. *
- d. Licensee Reference Material The licensee.is required to provide the NRC examination team with site specific reference material, that is current and '
accurate, for the preparation of the written and operating examinations. The requirements for the reference material was provided in Enclosure 1 to the letter formalizing the examination from Mr. Dennis F. Kirsch of the NRC to Mr. Robert H. Bridenbecker, SCE, dated February 14, 1990. l In the future, the reference material provided by the licensee i I should be further reviewed by the licensee for. accuracy. l Information contained in system descriptions and design basis ' documents was not consistently current or complete. Complete and l ' accurate reference material is necessary for the satisfactory preparation of examinations. Inaccurate information contributed to the deletion or revision of prepared examination material during both the pre and post examination reviews. I I l
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- 3. Exit Meeting At the~ conclusion of the site visit the. examiners met with representatives of the licensee. staff to discuss the findings as '
detailed in Section 2. of.this report. Licensee representatives did not offer responses to the examiners findings at the conclusion of the meeting, when requested by the Chief Exaciner. t F b
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c iONGS2/3 RDEIAM 6/29/90 SONS $ 2/3 SR0EIAM 6/29/90 TOPIC try 10Plc keyR04 110CFR20 sai dose a 1 10CFR20 sax dose a 1 e2Shiftreliefstatussht alb 2 Shift relief status sht alb 62 53 CR toenand f untinen a/d 3CRconsandfunction a/d #3
+4 REP sat dest rate t 4 REP six dose rate t 64
*) RD 109 symbols als 3 RG log synbols alb 45 66Throttledvivposition c 6 Throttled viv positten t 6 e7Securitykeys t 7 Security keys c 87 IB Manual valve opt t B Manual valve ops t 'B
#9 Ctsson CD fire duties d 9 DELAYED S/U ADMIN a SRO 610 T.S. sanning - 2 hrs b 10 Ceeson CD lire duties d 6 '
oli Safe S/D designation b 11 1.5. sanning - 2 hrs e 410 - 812 WAR verbal authorir, t 12 Saf e S/D cesignation b 11 el3 DELETED. x 13LONSTERMFERMISSION a SR0 614 DELETED. x 14WARverbalauthorir. t 12
#15 DELETED. x 15 Deleted. I 13 16ESFASinputcarameter b 16RCSCHEMISTRY d SR0 17CSASloputparaseter d 17 EVENT CLASS!FICAil0N b SRO elB Max startup rate alb 1BESFASinputparasater b 16 ett l'ypass Hi Log Pwr stpt. b 19 CSAS input paraseter d 17 20 RCP B/0 flow path d 20 CVCS effett of UPS loss t 121 521 CYCS effect of UPS loss t 21 Cause of CWP t/d 22 822CauseofCWP c/d 22Dilutioncalculatten d 23
*23 Dilution calculation d 23 Inadvertent CSAS t 24 624 Inadvertent CSAS c 24 Mintaus SDM b 25 25 Minisua SDM b 25 SDM SASIS d SRO 26RPStripcircuit d 26 RPS trir circuit d 26 527 $/D outsice IR Nis b/d 27 Sip vetside CR Nil b/d 427 !
2BCPCtespshacomingccep. a 2B CPC teep shadowing tono. a 28 29 Auto Letdown isolation d 29Letdo.n/chargingprecaut d 30 e30 Letdo' /tharging precaut ,d 30TSLCDforCRradtonito b 31 e31 IS LCD for CR rad monito b 31 AFW valve poi.er supply b 34 532SSWLCposttripdesand a 32 TD AFW llow lielter b 35 33 MSl$ input paras & stpt a 33 AFW S/G legit a 36 34 AFW valve pcuer supply b 34 Manual vs aute Sips a 37 35TDAFWlicalimiter b 35 ECP errors b *30 36 AFW S/S legit a 36 ARM auto llolation a 40 37 Manual vs auto BlAS a 37 SRNis IN N00E 6 t SRO 63BECPerrors b 3B flin leap trit LCD time a 541 4 9 RCP seal failure d 39RWSTIt0RONCON.IASIS a SED e40 ARM auto isolation a 40 AS! talculation c 42 641 Min Trap Crit LCD time a 41 ASl LinliS w/ COLSS 003 b SR0 42 ASI calculation c 42 RCP ops w/o CCW a 843 543 RCP ops w/o CCW a 43 CCW w/ LDP 6 SIAS b 44 i44CCWw/LDP&SIAS b 44 SDCS aan teap/ press. d 45 45SDCSaanteap/ press, d 45 IS LCD - Log twr (S/U) d 46 646 TS LCD - Log Pwr (S/U) d 46 E0P basis for 1 ave b 47 47 E0F basis for T-ave b 47 Core boiling talt. t AB 4BCoreboilingtalt, c 49 PZR level cospensation a 649 e49 FIR level ctapensation a 49 SBTR EOF basis for P/T t 50 ! 50 SBTR E0P tasis for P/T c 50 CW posp trip par limit t 51 ' l
'51 CW posp trip par limit c $1 Loss of HDP effects b e54 l e52 Cle;;ed CW sympttes c 52ResettingMSIS/CIAS a 55 l 853 Admin pcner lisit mFW) a 53 22tY lineup al VAls a e56 l 654 Loss of RP effects b $4 RAS reg sanual actions d 57 555 Resetting M31S / CIAS a 55 SDC/CS lineuo preregs. d 61
i 656 2?rV lineup al Dais $6 C11 viv sipal on SlAS a 1 662 U I.AB req. aantal actions d 57 T.S. FOR EDS FUEL O!L a Sk0 0 Min SDC flourate t $B PIR heater per supplies a 64 59 SDC L10P flompath a 59 HFS!/Sli FREREG (160C4) b SRO
$601 hot RfD f ailure c 60 RYL1 TC inputs d 565 661 SDC/CS lineup prereqs. d 61Maxtocideanrate b t6 662 Deleted. : 62 CFC inputs S logic a 567 863 Mi overspeed logic d 631.S.14516 FOR M!N TEMP c SRO 64 F'!R heater por supplies a 64 f.S. BASIS MS!VOPER. d SRO 865RVLITCinputs d 65 Deleted. I 568 66 Mas tooldown rate b 66 MS breakers open after t 6 470 467 CPC inputs 6 logic a 67 SFTA - sat RCS inyehtory $ Il '6BDeleted. : 6B SFTA - RCS heat renovel t 72 69 RCP SIU delta i basis b 69 SFTA - elettrata) L/U t 73 670 MS breakers open af ter t d 70 Manual MSly iso. rests c 74 71 SFTA - sat RCS inventory d 71 Lo er T-ave dilution 4 475 672 SFTA - RCS heat renoval t 72 Energ boration flompath b 76 73SFTA-elettricalL/U t 73 SFCS MAI 11ME 10 VERF. t SRO 74 Manual MSiv iso, rests c 74 E0P BAS 15 FOR MAI StM a SRO 475 Lower T ave dilution a 75MSISFASIS a SF.0 76 Estrg Poration flowpath b 76 CR EVACUATION SASIS b SR0 '
417 Instr Air nitrogen BIU c 77CREVACUAT10NBAS15 d Sk0 878 Loss of IA - plant SD a 78SIASandTS3.0.3 d SRO
*19 $1AS eflett on SFP e 79 Instr Air nitrogen S/U c 77 ' '80 Air to R pool seal a 80 Loss of IA - plant SD a 78 481 Natural Circ. verificati b Bl SIAS ellett on SFP a 79 B2 Energ Beration entry rqa d B2 Air to Rx pool seal a iB0 *B3 RCP hi teep - Rx trip t 63 CEA MISAllSNMENT PASIS a SRO B4 HPSI thrcttle triteria 4 $4 Natural Cirt. verificati b B1 885 ESCE isolation criteria c B5 Energ Beration entry rqa d E2 'B6 Cond, Vetuus bypass t 96 RCP hi teap - Rx trip t B3 *B7 !!<0 objectives d B7HPSIthrottlecriteria a 64 j *BB SPD - MSIS tasis _t BB ESDE isolation criteria e B5 89 SPD - battery basis d 69 Cend. Vacuus typass t B6 590Hotleginj. basis b 90 SB0 objectives d 87 591 Inverter loading tasis a 91 SB0 - MSIS basis t 500 692 Fire - aan in charge a 92 SB0 - battery basis 6 99 693ContetintegrityModee6 a 93 CLASS 1E !?0 VAC PER 151 a Sk0 i 94 Linear For Nl failure b 94 inverter leading basis a 491 95 Mk press. instr f alls d 95 Fire - san in charge a 492 '96 RPS single /redundent f ai c/d 96 Contet integrity Mode 6 e i93 497 CVCS - SlAS reset (FDPS) d 97 Linear Fur NI failure b 94 59B SBCS - T ave input t 98 FZR press, instr fails d 95 99 SEN! trarsfer to 'B' b 99 RFS single /redundent fat t/d 596 100
- I'repped CEA recovery c 100SLIFfEDCEAINDICA110N t SRO '
total number of questions: 95 Total number of questions: 97 o4
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Senior _ / Reactor A Operator i WRITTEN EXAMINATION ANSWER SHEET / POWER REACTOR: SONGS 2/3 EXAMINATION DATE: / Reactor Type: CE Region: V Passing Grade: 80% Time LTiiit: _4 hrs. l CANDIDATE'S NAME (please print): ! MULTIPLE CHOICE. MARK THE CORRECT ANSWER WITH AN X: l Example: .01: AXCD l
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.01: ABCD .26: ABCD .51: ABCD .76: ABCD l
.02: ABCD .27: ABCD .52: ABCD .77: ABCD
.03: ABCD .28: A B C D .53: ABCD .78: ABCD
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.04: ABCD .29: ABCD .54: ABCD .79: ABCD
.05: ABCD .30: ABCD .55: ABCD .80: A B C D
.06: ABCD .31: ABCD .56: ABCD .81: ABCD
.07: ABCD .32: ABCD .57: ABCD .82: ABCD
.08: ABCD .33: A B C D .58: ABCD .83: ABCD
.09: ABCD .34: A B C D .59: ABCD .84: ABCD '
.10: ABCD .35: ABCD .60: ABCD .85: ABCD
.11: ABCD .36: ABCD .61: ABCD .86: ABCD
.12: ABCD .37: ABCD .62:'A B C D
( .13: ABCD
.14: ABCD
.38: A B C D
.39: ABCD
.63: ABCD
.64: ABCD
.87: ABCD
.88: ABCD
.89: ABCD
.15: ABCD
.40: ABCD .65: ABCD .90: ABCD
.16: ABCD .41: ABCD .66::A B C D .91: ABCD
.17: A B C- D .42: A B C D .67: ABCD
.18: ABCD .92: ABCD
.43: ABCD .68: ABCD .93: ABCD
' .19: ABCD .44: ABCD .69: ABCD .94: ABCD
.20: ABCD .45: ABCD .70: ABCD .95: ABCD
.21: ABCD .46: ABCD .71:-A B C D .96: ABCD
.22: ABCD .47: ABCD . 72 :' A B C D .97: ABCD
.23: A B C D .48: ABCD .73: ABCD
.24: ABCD .98: ABCD (
.49: ABCD .74: ABCD .99: ABCD
.25: ABCD .50: ABCD .75: ABCD .100: A B C D' \'
t, Total Points Available: Score: Overall Grade: % NOTE:
- 1. Check that you responded to ALL questions, and that all responses are correctly marked ON THIS ANSWER SHEET!
- 2. Sign this answer sheet and turn in ALL exam materials to the NRC ;
EXAMINER, All work done on this examination is my own. I have neither given or received aid: candidate's Signature , _- - - _ .o - - _ . . _ . _ _.
t t ESo201 Rev 5 01/01/89 i 1
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ATTACHMENT 2 (continued) ' Enclosure 3 , PROCEDURES FOR THE ADMINISTRATION OF WRITTEN EXAMIii
- 1. i Check identification badges. !
2. Pass out examinations and all handouts. Remind applicants not to review ^*' examination until instructed to do so. ! t,
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READ THE FOLLOWING INSTRUCTIONS VERBATIM: l During the administration of this, examination the following rules apply:
- 1. ;
and could result in more severe penalties. Cheating on the eI 2. After the examination has beeh completed, you must sign the statement o ' the cover sheet indicating that the work is your own and you.have not received or given assistance in completin done after you complete the examination. g the examination. This must be READ THE FOLLOWING INSTRUCTIONS ( 1. i)) ! Restroom trips are to be limited and only one appl leave. .
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room to avoid even the appearance or possibility of cheating.
- 2. 3 Use black ink or dark pencil only to facilitate legible reproductions. i
- 3. (
Print your name insheet, the blank provided in the upper right-hand corner of P the examination cover I au, , a :.
- 4. ;
Fill in the date on the cover sheet of the examination (if necessary). 4 5. You separate maysheet writeofyour paper. answers on the examination question page or on a j THE BACK SIDE OF THE PAGE. USE ONLY THE PAPER PROVIDED A ' 6. If you write your answers on the examination question n ge and you nee; i more space to answer a specific question, use a sepc paper provided and insert it directly after the spec,iw sheet of the question. DO NOT WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION P:
- 7. ;
Print your name in the upper right hand corner of the first page of each section pages of your sheets or separate answer of sheets paper. whether you use the examination question Initial each page. . i 8. Before sheet, in'cluding you turn any inadditional your examination, pa consecutively number each answer ! on the examination question page. ges inserted when writing your answers ( ! Examiner Standards 14 of 18 t
ES-201 Rev 5 01/01/89 ( ATTACHMENT 2 (Continued) Enclosure 3 (Continued) 9. If you iare 1.04,6.10) number using separate sheets, number each answer as to cat allow sp(ac.e. e for grading. and skip at least 3 lines between answers to
- 10. Write "End of Category _
" at the end of your answers to a category.
11. Start each category on a new page. . 12. Write "Last Page" on the last answer sheet.
- 13. Use abbreviations only if the Avoid using symbols such as <y are commonly used in facility _ literature.
error resulting in an incorrect answer.or > signs to avoid a simple transposition Write it out. 14. The point question. value for each question is indicated in parentheses afte NOT an indication The amount of the of depthblank space required. of answer on an examination question page 15. Show all calculations, methods, or assumptions used to obtain a
- 16. .
Partial credit may be given. 3 AND multiple00 choice NOT LEAVE ANY ANSWER BLANK. questions.- Partial cr given on (. / 17. Proportional grading will be applied. that is provided may count against you. Any additional wrong information worth one point and asks for four re For example, if a question is points, 0.20 paints.and you give five responses,sponses, each of which is worth 0.25 If one of your five responses is incorrect, 0.20 will bee deducted 1.00 even though and yodt you got total credit the' four for that correct answers. question will be 0.80 inste 18. 19. If the intent of a question is unclear,,ast. questions of the When examination turning questions in your e examination, assemble the completed ex turn in all scrap pape,r.xamination aids and answer sheets. In addition,
- 20. To pass the examination, i greater and at least 70% you mustcategory.
in each achieve an overall grade of 80% or ! 21. There is a time limit of (6) hours for completion of the examination (or some other time if less than the full examination i
- 22. .
When area (DEFINE youTHE areAREA). done and have turned in your examination, nation leaveI
- 23. is still in progress, your If you are found license mayinbethis area while denied the examination or revoked
'Tf Ensure that all information you wish to have evaluated as part of your without answer is on your answer sheet. Scrap paper will be diposed of
~ review immediatly following the examination.
Examiner Standards , 15 of 18
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QUESTION: 001 (1.0) According lists the to 10 CPR 20, which ONE (1) of the following correctly ) whole body requirements doso. limit is which must be met if the 1-1/4 Rem /Qtr to be exceeded?
- a. Cannot exceed 3 Rem /Qtr whole body, 5(N-18) Rem whole body lifetime, and NRC Form 4 completed ;
- b. Cannot exceed 5 Rem /Qtr whole body, 5(N-18) Rem whole body lifetime, and NRC Form 4 completed c.
Cannot exceed 3 Rem /Qtr whole body, 3(N-18) Rem whole body lifetime, and NRC-Form 4 completed
- d. Cannot exceed 3 Rem /Qtr whole body, 3(N-18) Rem whole body lifetime ANSWER: 001 (1.0) pwg 3
- a. [1.0)
REFERENCE:
- 1. 10 CFR CH. 1 (1-1-89), Part 20.101 KSA 194001K103 2.8/3.4 l
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QUESTION: 002 (1.0) SONGS Procedure SO123-0-10, " Operations Shift Relief", requires you to complete Shift Relief Status sheets. Which ONE (1) of the following correctly describes when the Shift Relief Status sheets should be completed?
- a. One hour prior to shift turnover
- b. Continually during the shift
- c. During shift turnover
- d. 1-2 hours following shift relief ANSWER: 002 (1.0) pwg B
- b. [1.0) or 4,
REFERENCE:
1. SO123-0-10 " Operations Shift Relief", TCNO-15, p.4 hoA 194001A106 3.4/3.4 i I i 4
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g Nt i QUESTION: 003 (1.0) SO 123-0-15 " Control Room Access and Conduct", describes the control room nanning and responsibilities during a Loss of i Coolant Accident (LOCA). Which ONE (1) of the following personnel, by title, is responsible, per the procedure, for the control room command function, during a LOCA? !
- a. The SRO officially stationed in the control room !
- b. The senior operations manager on site
- c. The senior control operator i
- d. The shift superintendant ANSWER: 003 (1.0) pwg B
- d. (1.0) or 4.
REFERENCE:
1. p.7 SO123-0-15 " Control Room Access and Conduct", TCNO-3, KSA 194001A109 2.7/3.9 N EL
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QUESTION: 004 (1.0) I In accordance with SO 123-7-9.9, " Radiation Exposure Permit Program", an extended Radiation Exposure Permit (REP) is issued ' to cover work involving relatively low radiological hazards. Which ONE 1 selections (pr)ovided,of the following general area doso rates, from the p considered a relatively is the low MAXIMUM radiologicalradiation hazard?for an area to be i j
- a. 50 Mr/Hr
- b. 75 Mr/Hr !
- c. 95 Mr/Hr
- d. 150 Mr/Hr ANSWER: 004 (1.0) pwg B i
- c. (1.0)
REFERENCE:
1. SO123-7-9.9 " Radiation Exposure Permit Program", P.15 KSA 194001K104 3.3/3.5 I 1 i
QUESTION: 005 (1.0) Sol 23-O'-11," Narrative Logs", requires specific entries be made in the control Operators log when specific types of systems or components are-declared out of Service (00S). San Onofre's Unit 2 Shutdown Cooling System (SDCS) has been declared 00S. Which ONE (1) of the following symbols must be entere<' in the margin of the control Operators Log?
- a. Red E
- b. Black E
- c. Red 00S
- d. Black OOS ANSWER: 005 (1.0) pwg B
- b. (1.0) or a.
REFERENCE:
- 1. S0123-0-11 " Narrative Logs", TCN1-1, p.12 KSA 194001A106 3.4/3.4 l
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QUESTION:,006 (1.0) A system alignment is being performed that requires independent verification. An initial operator has placed a throttled rising Btem valve in its required throttled position. i Which ONE (1) of the following best describes the actions the j SECOND-operator should take?
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- a. Shut the valve completely and place in throttled position I l b. Open the valve completely and place in throttled
- position i i
- c. Observe that the valve is in an intermediate position
- d. Operate the valve only as necessary to verify it is in an intermediate position
! ANSWER: 006 (1.0) pwg B l c. [1.0] l
REFERENCE:
- 1. SO123-0-23 " Control of System Alignments",TCNO-7,p.10 KSA 194001K101 3.6/3.7
~
4 l 1
)
QUESTION: 007 (1.0) Which ONE (1) of the following_ personnel can authorize a security key being taken outside of the protected area?
- a. The shift superintendent
- b. Any control room operator
- c. The security shift commander
- d. The senior control operator- ,
ANSWER: 007 (1.0) pwg B
- c. (1.0)
REFERENCE:
- 1. SO123-0-27 " Key Control", p.4 KSA 194001K105 3.1/3.4 l
l
'1 l
l
l QUESTION - 008' (1. 0) Valve Operation procedure, SO123-0-23.1, allows-the SRO operations supervisor to approve the use of a leverage device on a manue.1 globe . val ve. Which DNE (1) of the following is the maximum' length of the leverage device that can be authorized? l
- a. As.long as'the handwheel radius
- b. 2 times as long as the handwheel radius
- c. 2-1/2 times as long as the handwheel radius
- d. 3 times as long as the handwheel radius ANSWER: 008 (1. 0) pwg B
- c. [1. 0] '
REFERENCE:
- 1. SO123-0-23.1 " Valve Operation",TCNO-7,p.3 KSA 194001A112 3.1/4.1 J
l l l l l l l l l 1
-l l
QUESTION: 009 (1.0) SO23-13-21 describes the duties and responsibilities of personnel during a fire. A fire is reported inside the protected area of San onofre Unit.2. Which ONE (1) of the following correctly describes the duties of the Common Control Operator-(41), after obtaining a copy of the ' zone evaluation?
- a. Remain in the control room, and coordinate the fire fighting effort
- b. Remain in the' control room, and act as technical.
advisor !
- c. Proceed to the scene, and act as fire brigade leader
- d. Proceed to the scene, and act as' technical advisor 5 ANSWER: 009 (1.0) pWg B
- d. (1.0)
REFERENCE:
- 1. SO23-13-21 " Fire", TCN1-5,p.6 KSA 194001K116 3.5/4.2 1
1
QUESTION: 010-(1.0) It is 2:00 AM in.the morning on Sunday. San Onofre Unit 2 is in Mode 1.with minimum shift manning. One (1) Reactor Operator (RO) assigned to Unit 2 Control room becomes sick and leaves the site. i Which ONE (1) of the following is the maximum time that Unit 2 shift may be undermanned?
- a. One (1) hour
- b. Two (2) hours '
- c. Three (3) hours i
- d. Four (4) hours ANSWER: 010 (1.0) pwg B
- b. [1.0) i
REFERENCE:
- 1. SO123-0-30, TCN 0-7,P.3 KSA 194001A103 2.5/3.4 s
'?
)
i I
QUESTION: 011 (1.0) l l Which ONE (1) lof the following components should be marked with I an orange triangular shaped emblem adjacent to its controls?
- a. Steam Bypass Control System Valve 8423
- b. Train A Charging Pump P190' l
- c. Containment Spray Pump P190 l
- d. RCP Controlled Bleedoff Containment Isolation Valve HV !
9318 ANSWER: 011 (1.0) pwg B
- b. [1.0) l
REFERENCE:
1. KSA 194001K101 3.5/3.7 l l l l l
l QUESTION: 012 (1.0) I SO 123-0-21 " Equipment' Status Control", lists maintenance that J requires a Work Authorization Record-(WAR) and maintenance that can be authorized verbally without a WAR.- ) j Which ONE (1) of the following tasks can be authorized verbally without a WAR?- a.. Removal of-lagging on boric acid heat traced pipe
- b. Packing adjustment of valves requiring inservice tests
- c. Vibration analysis
- d. Lubrication of control room emergency air system ANSWER: 012 (1.0) pwg B (1.0]
c.
REFERENCE:
- 1. SO123-0-21, " Equipment Status Control"TCN 1-6, pp.34,35 i
KSA 194001A112 3.1/4.1 i l
l QUESTION: 013 (1.0) UC5 / - SO123-0-21 " Equipment Status Control" dictates which equipment may be tested with the circuit in a non test. position. ! Of the following equipment, which ONE (1) SHALL be tested under an "In Test" position? Work Authorization with the circuit in the test
- a. Solenoid Operated Valves
- b. 120 Volt Power Circuits
- c. Power Receptacles
- d. ACB Breakers ANSWER: 013 (1.0) pwg B !
- d. [1.0)
REFERENCE:
- 1. (
SO123-0-21 " Equipment Status Control",Rev.1.TCN1-6,p.36 ' KSA 194001K107 3.6/4.1 l l n'r e - w y-- y y ,
QUESTION: 014 (1.0) yC5Y/ M ' Eh . l San Onofre Unit 2 is critical with T-avg of 532 degrees F. At.-least once per which ONE (1) of the following time limits must T-avg be verified' to be above its Technical Specification li. nit?
- a. 15 minutes
- b. 30 minutes
- c. 45 minutes
- d. 1 hour ANSWER: 014 (1.0) pwg R
- b. (1.0) i
REFERENCE:
- 1. Tech Spec 3.1.1.4 i
KSA 194001A102 4.1/3.9 KSA 294001A102 4.1/3.9 i
=
l
. D
gesI/M Dsweb. QUESTION: 015 (1.0) Plant conditions indicate likely major failures of plant
- functions needed for protection of the public.
1 Which ONE (1) of the following is the correct MINIMUM emergency I action level for these conditions? l
- a. Unusual Event I
- b. Alert ,
- c. Site Area Emergency
- d. General Area Emergency-ANSWER: 015 (1,0) pwg B
- c. (1.0]
l
REFERENCE:
I l
- 1. SO23-8-1," Recognition and Classification of j Emergencies", TCN7-1, p. 48 KSA 194001A116 3.1/3.5
QUESTION: 016 - (1. 0) ~ A LOCA has occurred and the following plant conditions exist.
- Reactor Coolant Pressure 1850 psia Containment Pressure 3.5 psig Steam Generator Pressures 810 psig Pressurizer level 10%
Which one of the. plant conditions will cause an Engineered
- Safety Features Actuation Signal (ESFAS)?
i a.-Reactor Coolant Pressure
~
- b. Containment Pressure
- c. Steam Generator Pressures
- d. Pressurizer level ANSWER: 016 (1.0) epe2 B i
- b. (1.0] ;
Reference:
- 1. SD-SO23-720 p. 6
- 2. SD-SO23-740 p. 10 KSA 000011A104 4.4/4.4 i
i 1 4 l
QUESTION 017 (1.0) Which one (1) of the following is the MINIMUM plant conditions that will cause a Containment Spray Actuation Signal (CSAS)?
- a. Containment Pressure 3.25 psig and a Pressurizer level of 10%.
- b. Containment Pressure 8.3 psig and a Pressurizer Pressure of 1800 psia.
- c. Containment Pressure 12.25 psig and a pressurizer level of 10%.
- d. Containment Pressure 16.3 psig and a pressurizer pressure of 1900 psia.
ANSWER: 017 (1.0) epe2 B
- d. [1.0)
REFERENCE:
- 1. SD-SO23-720 p. 13 KSA 000009A113 4.4/4.4 l
l I l l l
-~
QUESTION 018 (l'. 0) A reactor startup is being conducted in accordance with SO23 1.3.1, Plant Startup From Hot Standby to Minimum Load, and SO23- ' 3-1.1, Reactor.Startup. I Which one (1) of the following is the maximum positive startup' l rate that shall not be exceeded while conducting a plant startup?_ j
- a. 0.5 DPM.
- b. 1.5 DPM
- c. 2.5 DPM '
- d. 5.0 DPM ANSWER: 018 (1.0) gips R a.- [1.0) dr b.
REFERENCE; l l 1. SO23-5-1.3.1 p7 ! 2. SO23-3-1.1 p 29 KSA 001000K518 4.2/4.3 I I l w
QUESTION 019 (1.0) A reactor startup is in progress in accordance with SO23-3-1.1, Reactor Startup. .. Which one (1) of the following is the MINIMUM reactor power level at which the operator can bypass.the HIGH LOG' POWER Trip?
- a. 1 E -3 %
- b. 5 E -4 %
- c. 1 E -5 %
- d. 5 E -5 %
ANSWER 019 (1.0) g1Ps R
- b. (1.0)
REFERENCE:
- 1. SO23-1.1 p 29
- 2. SO23-5-2.11 KSA 015000K406 3.9/4.2 l
L t I . l l
1 l QUESTION 020 (1.0) l Which one (1) of the following receives-reactor coolant pump seal i bleedoff flow during normal plant operations? a.. Miscellaneous Rad Waste tank.
- b. Reactor Coolant Drain Tank. '
- c. Reactor Coolant Quench Tank.
- d. CVCS Volume Control Tank.
ANSWER: 020 (1.0) gips R
- d. (1.0]
REFERENCE:
1..SDSO23-360
- 2. 015000K406 KSA 004000K104 3.4/3.8 003000K103 3.3/3.6 l
l l l i
4 QUESTION 021 (1.0) l The unit is operating at 100% power with charging pump P-191 in- l operation and with charging pump P-192 selected as first backup ' when the NON 1E Uninteruptable Power Source-(UPS) is lost. Which one (1) of the following statements correctly describes the response of the Chemical and Volume Control System? L a. Charging pump P-192 will start due to a simulated low i pressurizer level. ! b. Charging pump P-191 will stop and charging pump P-192 will start to maintain pressurizer level.
- c. Charging pump P-191 Will stop due to-a simulated high pressurizer level.
- d. Charging pump P-192 and P-190 will start due to a simulated low pressurizer level.
ANSWER 021 (1.0) gips B I l c. [1.0) i
REFERENCE:
- 1. SO23-3-2.1 p 5 (4.11)
KSA 004000K101 l
l 1 QUESTION 0221(1.0) I 1 The unit is operating at power with the CEAs in Manual Sequential. Which of the following PRE-TRIPS would prevent you-from withdrawing _the CEAs?-
- a. 1 of 4 Low pressurizer pressure,
- b. 2 of 4 Low pressurizer pressure.
- c. 1 of 4 Low DNBR. I
- d. 2 of 4 Low DNBR. i i
ANSWER 022 (1.0) gips B
- d. (1.0) or c,
REFERENCE:
- 1. SD-SO23-510, p 38 KSAs 001000K1043.2/3.4 001000K105 4.5/4.4 ,
I
,_3 A0;d
QUESTION 023 (1.0) It is desired to raise Tave 5 degrees by changing boron concentration. The following plant conditions: exist: Reactor Power: 70% CEA Positions: All CEAs Full Out RCS Boron concentration: 730 ppm Total Power Coefficient: H0.02 delta rho'/- percent power Isothermal Temperature -0.010 percent Coefficient: delta rho / degree F Inverse boron worth: 70 ppm / percent delta rho Which one (1) of the following is the: final RCS boron concentration required to increase Tave by 5 (five) degrees F ' WITHOUT changing reactor power?
- a. 723.0 ppm
- b. 724.5 ppm
- c. '725.0 ppm
- d. 726.5 ppm ANSWER: 023 (1.0) g1ps B
.i
- d. (1.0) calculation: ((-0.010 x 5)*(70)} = - 3.5 , 730 - 3.5 = l 726.5 ppm
REFERENCE:
- 1. SONGS Reactor Theory P. 176, 196 KSA 004000A404 3.2/3.6 001010K518 3.2/3.6 l
l
1 l QUESTION 024 (1.0) i The unit is operating at 100% power in Mode 1 when a spurious Containment Spray-Actuation Signal (CSAS) occurs. 1 Which one (1) of the following correctly describes what occurs as l a result of the CSAS?
- a. Containment Spray pumps receive an auto start signal,
. pump-recirculation valves open. l I
- b. Containment spray pumps do NOT receive an auto start signal, pump suction and recirculation valves open.
- c. Containment spray pumps do NOT receive an auto start signal, spray header isolation valves open.
- d. Containment spray pumps receive an auto start signal, spray header isolation valves open.
ANSWER: 024 (1.0) gips B
- c. (3.0) ,
REFERENCE:
- 1. SD-SO23-740, p 30, 33 KSAs 013000K105 4.1/4.4 l
1 l l l
' ~
c L . QUESTION 025 ( 1, 0 ) .
' Technical Specification 3.1.1.1 requires a specific shutdown margin be maintained during modes 1, 2, 3, and 4. !
b l Which one (1) of the following is the MINIMUM Shutdown Margin required-by Technical Specifications 3.1.1.1 for mode 17 !
- a. 15.15.% delta k/k.
b.. 5.15 % delta k/k.
- c. 4.15 %_ delta k/k.
- d. 3.0 % delta k/k. 4 ANSWER: 025. (1.0) gips =B
- b. [1.0)
REFERENCE:
- 1. SO23-3-3.29 6-5
- 2. SONGS 2/3 Technical Specification 3.1.1.1 l KSA 001010K535 1.3/3.6 l
I i 1 1
1 l l QUESTION 026 (1.0) j The reactor protection _ system requires a minimum 2 (two) channels of reactor trip signals'to cause a loss of power to the CEDM coils. See attached drawing. 4 Which one of the following combinations of the trip circuit will cause a loss of power to all of the CEDM coils?
- a. SSR K1 and SSR K2 energized. !
~b. SSR K1 and SSR K2 deenergized.
- c. SSR'K3 and SSR K2 energized.
- d. SSR K3 and SSR K2 deenergized.
ANSWER: 026 (1,0) glps B
- d. (1.0)
REFERENCE:
SD-SO23-710 p 60 KSA 001000K105 4.5/4.4 001000K202 3.6/3.7 i L l l l
A A' 4 480 v REACTOR TRIP 4eo v . Bus ' sus 2 STATUS , 120 VAC M 820 VAC ift7AL G ' VITAL i BUS 2 G gg3 3
,, . I POWER POWER SUPPLY TCB 9 y i
, 3
~ t 5 1 TCB2 TCB 6 TCB _3 _ _ _ TCB_7 ' _ _ ns I x n2 - - -______ _ _ r . z '[ PHASE CUARENT [ l i I I I I i I .I I I I I . ,x_L i _L c 1 7 (w) K
-- - - -- . U----_ __
4 (w) 1 x x r 3.. . 1 4 POWER POWER SUPPLY SUPPLY I20 VAC 820 VAC VITAL , CEDM U/V VITAL ~ nirs t BUS 4 1
. g.#- =,.- .~o +%,-
.s .s y -w .
n - * . - -
+.ru--+- -- . - - - - + w- _. ___.______r__._
1 1
. _ _ l QUESTION 027 (1.0) 1 l
Which ons"(1-) of the"following excore nuclear instrumentation L channels is REQUIRED to be used during a " Shutdown from Outside the Control Room," SO23-13-12?
- a. A linear power channel,
- b. A logarithmic power channel,
- c. A control channel.
L
,_ d. _ A,_startup channel.
ANSWER: 027 (1.,0L.91ps B .,_ _, _
- d. (1.0) or b. j
REFERENCE:
- 1. SD-SO23-470 p-4 KSA 015000K403 3.9/4.0 k
e- *
- g e d
-w t
t
% d-
\
i
)
i QUESTION 028 (1.0) . What is the purpose of providing Temperature Shadowing Compensation for the excore detector inputs to the core protection calculators-(CPCs)?
- a. Account for reduced neutron leakage _from the core at i lower temperatures. i i
- b. Account for reduced neutron-leakage from the core at !
higher temperatures. !
- c. !
Compensates for increased detector output 1due to lower temperature in detector well. ; 4
- d. $
_ Compensates for increased neutron detector outputidue-to higher temperature in detector well. ANSWER: '028 (l'.0) gips B
- a. [1.0)
REFERENCE:
- 1. SD-SO23-710 KSA 015000K502 2.7/2.9 6
i
.j i
l QUESTION 029 (1.0) ! The unit is operating with charging ~and letdown in service. ! Which one (1) of the following plant conditions or signals will I cause letdown to automatically isolate? -i
- a. Containment pressure 1.5 psig.
- b. VCT level high. .
- c. High~ letdown flow. ;
d.- High letdown temperature. ANSWER: 029 (1.0) g1ps R
- d. [1.0)
REFERENCE:
- 1. SD-SO23-390 i
KSA 004010A205 4.1/4.3 013000K111 3.3/3.8 i i
)
I i 4 l P
-^--
QUESTION 030 (1.0) Plant operating instruction SO23-3-2.1, CVCS Charging and Letdown, cautions operators to establish letdown immediately after initiatilig charging. What is the reason for ensuring that lotdown is established immediately after initiating charging?
- a. Prevent thermal shock of the VCT.
b. Prevent an inadvertent boron dilution accident.
- c. Prevent thermal shock to the non regenerative heat exchanger. )
- d. Prevent thecmal shock to the RCS loop inlet nozzles.
ANSWER 030 (1.0) g1ps B
- d. (1.0)
REFERENCE
- 1. SO23-3-2.1 p 5, 4.7 KSA 004000K511
, 3.6/3.9
QUESTION 031 (1.0) i The reactor operstor reports to you that both of the control room 1 l airborne monitors RE-7824-1 and RE-7825-2 have failed low. The I&C technician indicates that it will take five days to replace the detectors. What action must be taken (in accordance with Technical Specification 3.3.2 and 3.3.3.1) within one (1) hour? i
- a. Return the monitors to service, or be in Hot Standby within six (6) hours.
- b. Initiate and maintain operation of control room emergency air cleanup system.
- c. Establish a portable control room gaseous process radiation monitoring station.
d. Perform a surveillance test on the control room emergency air cleanup system. ANSWER: 031 (1.0) gips B
- b. (1.0)
REFERENCE:
1
- 1. SONGS 2/3 Technical Specifications 3.3.2 and 1.3.3.1
- 2. SD-SO23-700 p 37 KSA 072000A202 2.8/2.9 072000 GEN 2.1/3.4 I
QUESTION 032 (1.0) A reactor trip hr.s occurred on Unit 2. The cause of the trip was i an overfeeding of steam generator E-089. The trip also resulted in a reactor t?ip override (RTO) and a high level override (HLO). Present stear generator levels are the same as when the trip occurred: E-089 high, 92%, and E-088 is at normal .perating level. What will be the flow demand signal sent to the main feedwater regulating valve for E-089 (2FV-1111), with both the Reactor Trip i Override (RTO) and the High Level Override (HLO) signals in? '
- a. 0%
- b. 5%
- c. 25%
- d. 50%
ANSWER: 032 (1,0) g1ps R
- a. (1.0)
REFERENCE:
- 1. SD-SO23-250 p 56 - 89 KSA 059000A203 2.7/3.1 059000K418 2.8/3.0 I
i l I 1
QUESTION 033 (1.0) The following plant conditions exist immediately prior to a reactor trip: Pressurizer pressure 1850 psia Pressurizer level 29% Containment pressure 1.5 psig Steam Generator E-088 Pressure 705 pain Steam Generator E-008 Level 27% Which one (1) of the following will have caused a Main Steam Isolation Signal (MSIS) to be generated?
- a. Steam Generator E-088 Pressure 705 psia.-
- b. Steam Generator E-088 Level 27% and Pressurizer 1 Pressure 1850 psia.
- c. Steam Generator E-088 Level 27% and Containment pressure 1.5 psig,
- d. Pressurizer Pressure 1850 psia.
ANSWER 018 (1.0) g1ps R
- a. [1.0)
REFERENCE:
- 1. SD-SO23-720 p15, 16 KSA 059000K419 3.2/3.4 013000K115 3.4/3.8 '
013000A105 3.4/3.6
4 QUESTION 034 (1.0) A loss of all AC power has occurred in Unit 2. Which one (1) of the following auxiliary feedwater valves will be , able to be electrically controlled using DC power? See attached drawing.
- a. Auxiliary Feedwater pump P-504 discharge bypass valve 2HV-4762.
- b. Auxiliary Feedwater pump P-140 discharge valve 2HV-4705.
- c. Auxiliary Feedwater pump P-504 discharge valve 2HV-4712.
- d. Auxiliary Feedwater Pump P-141 discharge valve 2HV-4713.
ANSWER: 034 (1.0) g1ps B
- b. (1.0)
REFERENCE:
- 1. SD-SO23-780 p 5, 6 KSA 061000A203 3.1/3.4 4
SIMPLIFIED DIAGRAM OF THE AUXILIARY FEEDWATER SYSTEM MINI-FLOW TO CST T-121
~~
HV-4713 p HV-4731*
- MFN-N hs b-h+ ' s
' l a l! MINI-FLOW LO LC]( 4763 :- CST T-121 Z P-141 f40 TOR . o
, ""yXe uv_4715, FAOM
. Z .-
CST T-121 N " o r 8 ' OUTSIDE ! ' INSIDE CONTAINMENT CONTAIPNENT c a usT To - ATMOS ,g ,g Z. sv- g ! s : 4705 d KOO7 : P 140 I
/ sv- '*-
4700 47s2 : gy_ Ns3 uv-4730 FROM ; : MAIN B200 @M"Ns STEAM y 1 -
-{ ,
85' Fnos 4 CST T-121 to _
;Mi-N X LO LC)[k HV-4712 l ,' ' k :! N ' ,j ,"*
HV-471M, T . 140 TOR Z P-504 ' MINI-FLOW TO 4 CST T-121 FAOM Q u g CST T-121 LO . O m - m n e rA r7
. ( t, .
. R -
O
- _ - _ _ _ _ . . _ . . _ . . _ _ . . , ~ _ .- ... .- _- - . _ _ _ _ _ _ _ _ . _ _ _ _ - _ _ - . .
i QUESTION 035 (1.0) Which one (1) of the following methods is used to limit the flow of the TURBINE DRIVEN auxiliary feedwater pump to 1000 GPM in the event of a feedline break inside the containment?
- a. Discharge valves close to 40% open on a high flow signal from FE-4720 or FE-4725.
- b. A cavitating venturi is installed on the discharge of the pump.
- c. Discharge valves are preset to open to limit flow from the pump _to 1000 GPM.
- d. The flow signal from FE-4720 or FE-4725 will limit ,
steam flow to the turbine. ANSWER: 035 (1.0) epe2 B
- b. (1.0)
REFERENCE:
- 1. SD-SO23-780 p. 5
- 2. P&ID 40160A-15 KSA 000054A204 4.2/4.3 4
l 1 1 i
QUESTION 036 (1.0) Unit 2 has tripped and the following conditions are present: ' Pressure Level ! Steam Generator E088 700psig 10 % l l Steam Generator E089 625psig 10 % ) Which one (1) of the following statements correctly describes the response of the Auxiliary Feedwater system?
- a. Auxiliary feedwater will initiate to SG E088 ONLY.
- b. Auxiliary feedwater will initiate to SG E089 ONLY.
- c. Auxiliary feedwater will initiate to BOTH steam generators.
- d. Auxiliary feedwater will not be initiated.
ANSWER: 036 (1.0) gips B
- a. (1.0)
REFERENCE:
- 1. SD-SO23-780 p 81 KSA 061000K414 3.5/3.7 1
1 l QUESTION 037 (1.0) i Which one (1) of -the following Engineered Safety Feb.tures I Actuation Systems (ESFAS) signals are generated by an AUTOMATIC Safety Injection Actuation signal (SIAS) but not by a MANUAL ; SIAS? 4 i a. Containment Cooling Actuation Signal (CCAS). 1 b. Main Steam Isolation Signal (MSIS). !
)
- c. Main Feedwater Isolation Signal (MFWIS). j
- d. Non Critical Loop CCW Isolation signal.
ANSWER: 037 (1,0) g1ps B
- a. (1,0)
REFERENCE:
- 1. SD-SO23-720 p 8 KSA. 013000K103 3.8/4.1 i 022000K403 3.6/4.0 w
r A t t
QUESTION 038 (1.0) Unit 2 is in the process of a Hot Restart, at End of Life (EOL) in accordance with plant operating instruction SO23-5-1,3.1. An Estimated Critical Position (ECP) bas been determined. The reactor operator has followed the procedure in pulling rods to the ECP but does not have indication that the reactor is critical. Which one (1) of the following conditions will cause the actual critical rod CEA position to be HIGHER than the Estimated Critical Position (ECP)?
- a. Delaying the time of startup from 18 to 22 hours after the reactor tripped from steady state high power.
- b. Misadjustment of the steam bypass controller to be 50 psig higher than the normal no load setting.
- c. An inadvertent boron dilution of the reactor coolant system (RCS) during the rod withdrawal.
- d. Reactor coolant temperature is lower than assumed in the Estimated Critical Position calculation.
ANSWER: 038 (1.0) gips B b.
REFERENCE:
- 1. SONGS Reactor Theory p 181
- 2. SD-SO23-175 p 37 l
KSA 001010A207 3.6/4.2 l 1 l
j l l QUESTION 039 (1.0) l The plant is operating at 100% power with the following reactor coolant pump (RCP) seal conditions for RCP P001: Vapor seal-cavity pressure 75 psi Upper seal cavity pressure 75 psi , Middle seal cavity pressure 2250 psi USE THE ATTACHED DRAWING TO ANSWER THIS QUESTION. Which of the following correctly describes the RCP P001 seal failure?
- a. Lower seal failure only.
- b. Middle seal and vapor seal failure only.
- c. Vapor seal and Upper seal failure only.
- d. Upper seal and lower seal failure only.,
ANSWER: 024 (1.0) gips R d.
REFERENCE:
- 1. SO23-13-6 p6
- 2. SD-SO23-360 p 19 KSA 003000K602 2.7/3.1 003000A404 3.1/3.0 003000A109 2.8/2.8 003000A201 3.5/3.9
b ' O O
- l:
.l NUCLEAR GENERATION SITE ABNORMAL OPERATING INSTRUCTION S023-13-6 UNITS 2 AND 3 piVISi9N O PAGE 6 0F 9 e !
ATTACHMENT 1 TCN 0-A SEAL FAILURE ANALYSIS FROM CAIITY PRESSURE INDICATIONS NOTE: Table Assumes Nominal Values as follows: RCS Pressure at 2250 psia, Controlled Bleed-off Flow at 1.5 gps, VCT Backpressure at 75 psia. f NOTE: Failure of the vapor seal (Seal #4) may be indicated by little or no CB0 flow and the vapor seal
; cavity pressure approaching Containment Pressure.
NOTE: Differential pressure of less than 100 psid between seal cavities is an indication of associated seal i failure. ML SEAL 5 UPPER ANO MIDDLE ANO UPPER AND FAIL OR NORMAL UPPER SEAL MIDDLE SEAL LOWER SEAL MIDDLE SEAL LOWER SEAL LOWER SEE C80 CHECK FAILtRE FAILtRE FAILURE FAILtRE FAILURE FAILURE VALVE CLOSE0 I I ' LOW PRES 5URE
~
_. 1d I I I I I I 75 2250 - VAPOR SEE WITY I I I I l i J UPPER SEAL ' 800 7s - 114.5 75 2250 75 2250 - UPPER SE E W ITT I MIDDLE SEAL 4 I I l 1525 - 11h2.5 2250 75 2250 - - MIDDLE SEE WIM i I l._ l LOWER SEAL l I I__- ,23. - I I T._ - -
- ==Y ,REs5=E
! I I I I I I i _ - - - - - . - - _ . _ _ _ _ _ -...__.----__.---__---.__---_---------_a. _ . - - _ a e + , . - <m _ e-- -----n . - _ _ ---- _w r_.w- s w- a,_a wu.---w- - , _ _ _ _ _ _ _ , _ _ .~ - - - ----___--__x_---a,_u=.a
QUESTION 040 (1.0) The Area Radiation Monitoring System perform specific isolation functions. generates a signal to Which one (1) of the following isolation functions can be generated by the Area Radiation Monitoring System?
- a. Containment purge isolation.
- b. Containment isolation.
- c. Blowdown isolation.
- d. Fuel building ventilation isolation.
i.!!3WER: 040 (1.0) 91ps B a.
REFERENCE:
- 1. SD-SO23-690, p 29 KSA 072000K401 3.3/3.6 072000K403 3.2/3.6 9
i i QUESTION 041 (1.0) ; The unit la operating at 60% reactor power, at 4:00 AM you observe that Tave has decreased below 520 F and cannot be restored. Which one (1) of the following times correctly describes the time f at which the unit must be in HOT Standby if Tave can not be restored within Technical Specification limits?
- a. 4:30 AM ,
- b. 5:30 AM 6:30 AM
- c. *
- d. 10:30 AM ;
ANSWER: 041 (1.0) g2ps B
- a. ,
REFERENCE:
- 1. SONGS 2/3 Technical Specifications 3.1.1.4 KSA 002000K510 3.6/4.1 002000A109 3.7/3.8 i
1 l l i
i QUESTION 042 (1.0) i i Due to a sudden turbine load reduction Group 6 regulating CEAs were manually driven into the core to maintain Tave within the ; prescribed operating band. As a result of the CEAs being driven I in the following core power conditions exist: 1
~
Upper Half power 45 % Lower Half power 55 % Which one (1) of the following is the correct Axial Shape Index ! (ASI) for the given core power conditions? '
- a. + 0.2
- b. - 0.2 ,
- c. + 0.1 5 i
- d. - 0.1 ANSWER: 042 (1. 0) gips B c.
(55 - 45) / (55 + 45) = 0.10 l
REFERENCE:
- 1. SONGS 2/3 Technical Specification 3.2.7 & Definition KSA 015020K503 3.3/3.7 !
i t t k b .
QUESTION 043 (1.0) Component Cooling Water (CCW) has been lost to a Reactor Coolant Pump (RCP) while in Mode 1. You have entered OI SO23-13-6, RCP Seal Failure, and AOI SO23-13-7 Loss of CCW. / SWC. ; What is the MAXIMUM time that a Reactor Coolant Pump can be operated without component Cooling Water (CCW)?
- a. 3 minutes
- b. 5 minutes
- c. 15 minutes
- d. 30 minutes I
ANSWER: 043 (1.0) g3ps~B ! kW
REFERENCE:
- 1. SO23-13-7 p3 KSA 008030K302 4.1/4.2
, h' 7
P
QUESTION 044 (1.0) Component Cooling Water Pumps P-024 and P-026 are operating, P-025 is aligned to Train A and is in standby. A Safety Injection Actuation Signal (SIAS) and a Loss of Voltage Signal (LOVS) (on both safety related busses for greater than five seconds) have been received. Which one of tne following correctly describes the CCW pump responses when the busses are reenergized?
- a. P-024 and P-026 start P-025 does not start.
- b. P-025 and P-026 start P-024 does not start.
- c. P-024 and P-025 start P-026 does not start.
- d. All CCW pumps start.
ANSWER: 044 (1.0) g3ps B b.
REFERENCE:
- 1. SD-SO23-400 KSA 008030K201 2.9/3.0 008030K405 2.7/2.9 l
l .
QUESTION 045 (1.0) j operating Instruction S023-3-2.6, Shutdown Cooling System (SDCS), cautions the operator not to allow the SDCS and interconnecting , piping to exceed a specific temperature and pressure. What is the maximum pressure and-temperature that the SDCS shall not exceed?
- a. 375 deg F and 375 psia I
- b. 350 deg F and 400 psia !
)
- c. 375 deg F and 400 psia
- d. 350 den t' and 375 psia ANSWER: 045 (1.0) g3ps B d.
REFERENCE:
- 1. SO23-3-2.6 p6, 4.1.1 KSA 005000K401 3.0/3.2 4
i QUESTION 046 (1.0) Unit 2 is in Mode 2 (startup). One of the Logarithmic Power Level channels has failed HIGH. 1 i What are the Technical Specification 3.3.1 required actions for a failed Logarithmic power channel in Mode 27 .
)
- a. Restore the channel to operable condition within one hour or be in Mode 3 within 2 hours.
- b. Place the channel in the bypass condition within two '
hours or be in Mode 3 within 6 hours. '
- c. Restore the channel to operable condition prior to ,
entry to Mode 1.
- d. Place the channel in the bypass condition within one hour, then continue to Mode 1.
ANSWER 046 (1.0) epe2 B d.
REFERENCE:
- 1. SONGS 2/3 Technical Specification 3.3.:-
KSA 000032K301 3.2/3.6 000032G003 2.6/3.3 1 l l l l l '
l l QUESTION 047 (1.0) l Emergenc Actions,yrequires operating Instruction that S023-12-1, Reactor Coolant Standard System Post Trip (RCS) average l temperature, Tave, be checked to be between 545 degrees F and 555 degrees P. Which one (1) of the following correctly describes the basis for ' the requirement that RCS Tave is within these limits?
- a. To ensure that main s'.eam isolation signal does not occur. !
- b. To ensure that the steam generators are removing RCS heat. )
- c. To prevent uncontrolled filling of the steam generators.
- d. To prevent an imbalance in steam generator cooldown.
ANSWER: 047 (1.0) epe 2 B ,
- b. *
REFERENCE:
- 1. SO23-12-1 p 7, 8 ,
CEN-152 p 2-16 KSA 000007K106 3.7/4.1 L e 1
I t QUESTION 048 (1.0) Unit 3 is in Mode 5 with the RCS partially drained for maintenance. One LPSI pump is OOS, the operating LPSI pump trips. The core exit temperature is 124 degrees F. The reactor was shutdown 40 days ago from 100% full power. (See attached SO23-13-15 attachment 3.) Which one (1) of the following correctly identifies the predicted time that should elapse before boiling occurs in the core?
- a. 35 minutes.
- b. 38 minutes.
- c. 40 minutes.
- d. 42 minutes.
ANSWER 048 (1.0) epe2 B c.
REFERENCE:
- 1. SO23-13-15 Attachment 4, REV 2 KSh 000025G002 3.9/3.9 l
- . . _ - - - - . . - - . . - . - _ _ . - - - - - _ ~ - ._- - - -.. ..
NUCLEAR GENERATION SITE ABNORMAL OPERATING' INSTRUCTION $023o13-15 ! UNITS 2 AND 3 ' REVISION 2 PAGE 35 0F 39 ATTACHMENT 4 i TCN 2 -1 REACTOR CORE HEATI)P TIME AFTER A LOSS OF COR COOL .NG
- NOTE: This table' is based on RCS level at the Center of the Hot leg after a
- trip with an infinite operating history at 100% power. (Ref. A01 080) -
- 1. 'F Record RX Core Exit Temeerature at Time Zero (loss of SDC flow): ,
- 2. Record RX Core Heatuo Rate for Time After Trip (table below): 'F/ min.
- 3. Calculate Time to reach 200'F/ ALERT: (If in Mode 4, then 350'F/$1TE AREA EMERGENCY.)
(200'F -
) + =
1 (or350'F) Temperature Reactor Core Time to 200'F at Time Zero Heat Up Rate [or350'F)
- 4. Calculate Time to reach 212'F (8 oiling): (Mark N/A if in Mode 4.) #
(212*F -
) + =
Temperature Reactor Core Time to 212*F at Time Zero Heat Up Rate 5. ( Notify the Shift Superintendent /STA of ALERT (or SITE AREA EMERGENCY) and Boiling times. TIME AFTER DECAY HEAT REACTOR CORE TIME FROM TIME FROM TRIP LOAD (days) - HEATUP RATE l'20'F TO 145'F T0 (MWth) B0ll (min) ('F/ min) B0ll (min) 0 209 72.0 1.3 0.9 1 17 5.8 16.0 11'.5 2 5.0 14 . ,, . . 19.0 .. 13.4 3 13.0 4.6 20 14.5- i 4 12.0 4.3 22 16.0 ; 5 11.5 4.0 23- 16.7 10 - 9.7 3.3 28 20 15 8.5 2.9 32 23 20 7.6 2.6 35 26 30 6.9 2.4 38 28 40 6.3 2.2 43 31 50 5.7 2.0 47 34 80 4.9 1.7 54 39 100 4.2 1.5 63 45 200 3.2 1.1 83 61 ' 300 2.9 1.0 93 67 { FILE DISPOSITJON: File per S0123 0 25. T15 2.wp5 ATTACHMENT 6 PAGE 1 0F 1 1
QUESTION o49 (1.0) The control room has been evacuated in accordance with SO23 2, Shutdown from Outside of the Control Room. l The following plant conditions exist: $ Pressurizer Pressure 360 psia Pressurizer Level" ~~ ~ ~- ~ ~ ~ ~ LI-0103 indicates 25 % Which one (1) of the following is the correct pressurizer level at the Essential Plant Parameter Monitoring panel? See attached sheet SO23-13-2 Attachment 4. ' a.. 30 %
- b. 35 %
- c. 38 % .
- d. 43 %
ANSWER 049 (1,0) epel B Mb REFERENCE
- 1. SO23-13-2 p 30 l KSA 000068A128 3.8/4.0 j 000068K318 4.2/4.5 l
l ps I \ rt,( l l
,- . -_ -- - - ~~ ~ ~ ~ ~ ~ ~ ~ ~ ^ ~~^
l NUCLEAR GENERATION SITE UNITS 2 AND 3 ABNORMAL OPERATING INSTRUCTION S023 13 2 i REVISION 2 PAGE 30 of 206 i ATTACHMENT 4 E~l TCN
^
LI-0103 COMPENSATION CHART 2250 1100 360 psia psia psia AMBIENT 100 ~
/ / / /
/ / / / ...
/ / / _/
eo / // / -
/ / / /
/ / / /
* / //
~
/
60- / // / d / // / . D / // /
////
((' j 5 40 (([ /
$ /// /
///
f/ i 20- M
///
f/ M/ O !! 0 20 40 60- 80 l$o l INDICATED LEVEL - % [ L - T2 2.wp4 ATTACHMENT 4 PAGE 6 0F 6 l
QUESTION 050 (1.0) Steam Generator Tube Rupture EOI SO23-12-4 cautions the operators to maintain Post Accident Pressurizer Pressure and Reactor Coolant System Temperature within cooldown limits. Which one (1) of the following correctly describes why maintaining Post Accident Pressure and Temperature Limits at the lower end of the band will help mitigate the consequences of a steam generator tube rupture?
- a. Reduces -the amount of feed water inventory required for the affected steam generator.
- b. Reduces the. amount of feed water inventory required for the unaffected steam generator,
- c. Reduces the amount of Icakage from the Reactor Coolant System to the steam generator,
- d. Reduces the amount of leakage from the steam generator to the Reactor Coolant System.
ANSWER: 050 (1,0) epe2 B c. l
REFERENCE:
- 1. SO23-12-4 p5 ESA 000037K306 3.6/4.1 l
l 1 i
l 1 QUESTION: 051 (1.00) The plant is operating at power. One (1) Circulating Water (CW) Pump has just tripped due to an electrical fault; three (3) CW Pumps are still running. Which of the following is MAXIMUM power limit for three (3) pump operation?
- a. 95% *
- b. 86%
- c. 75%
- d. 66%
ANSWER: 051 (1.00) g2ps B C.
REFERENCE:
SO23-2-5, 4.3.1 (page 4 of 68) KA: 075000A202 2.5/2.7 (Site specific importance.) 075000K304 1.9/2.1 (Site specific importance.) k l t I I e l l l
QUESTION: 052 (1.00) Which of the following symptoms is a slowly developing indication of the Circulating Water INTAKE screens starting to be choked with shells or debris?
- a. Decreased pump amps due to decreased flow
- b. Decreased pump amps due to decreased pump differential pressure
- c. Increased pump amps due to increased pump differential pressure
- d. Increased pump amps due to increased flow ANSWER: 052 (1.00) g2ps R C.
REFERENCE:
SO23-2-5, Attachment 9, 2.0 Notes (page 59 of 68) KA: 075000A201 3.0/3.2 075000K604 1.5/1.6 075000K504 1.4/1.6 1 l i t i
QUESTION: 053 (1,00) j The plant is operating at full power and all systems / components are operable. Which one of the following components would require a power reduction to 75% in accordance with SO23-5-1.7 " Power Operations," if it were taken out of service?
- a. One (1) Main Feedwater Pump '
- b. One (1) Steam Generator Atmospheric Dump Valve
- c. One (1) Heater Drain Pump
- d. One (1) Condensate Pump i
ANSWER: 053 (1.00) g2ps R a.
REFERENCE:
SO23-5-1.7, Attachment 6 i KA: 039000K305 3.6/3.7 I
l l QUESTION: 054 (1.00) ! i The plant is operating at full power and one (1) Heater Drain Pump has tripped due to an electrical fault. ! Which of the following components may be damaged if operations are' continued in this condition (for more than an hour)?
- a. Main Turbine blades
- b. Main Condenser tubes
- c. Heater Drain Tank level control valves
- d. Main Feedwater Pump impeller ANSWER: 054 (1.00) g2ps B b.
REFERENCE:
SO23-2-3, 4.3 (page 3 of 74% KA: 039000K106 3.1/3.0 ,
P QUESTION: 055. (1.00) The plant was operating at power when a Main Steam Isolation Signal (MSIS) and containment Isolation Actuation Signal (CIAS) were actuated. Which of the following is the MINIMUM action that must be taken to allow the MSIVs to be opened from the.concrol Room?
- a. Both the CIAS and the MSIS must be reset l-
- b. Only the MSIS must be reset
- c. Only the CIAS must be reset i
- d. The MSIVs can be opened using OVERRIDE l
ANSWER: 055 (1.00) g2ps B I
- a. !
4
REFERENCE:
SD023-160 page 20 of 78 I KA: 039000K405 3.7/3.7 ;
+
t
i I'
.l l
QUESTION: 056 (1.00) The plant is shutdown and the Unit Auxiliary Transformers (UATs) have been lined up and are supplying off-site power. l Which of the following component alignments must be utilized?_
- a. Main Generator output breakers must be closed i
- b. Main Generator disconnects must be insta,lled ,
- c. Reserve Auxiliary Transformer must be cleared
- d. Main Transformer must be cleared ANSWER: 005 (1.00) g2ps B a.
REFERENCE:
SO23-6-5 Attachment 5, page 1 of 7 KA: 062000K104 3.7/4.2 t
--v - , - -
- QUESTION: 057 (1.00)
A large LOCA occurred about 20: minutes ago while the plant was
- operating at power.
Which one of the following actions must be MANUALLY-performed after a valid Recirculation Actuation Signal (RAS) has been received?
- a. -The Low Pressure Safety' Injection Pumps must be tripped
- b. The High' Pressure Safety Injection Pumps must be .l tripped
- c. The Containment Emergency. Sump isolation valves must be opened t
- d. The Refueling Water Storage Tank solation. valves must be shut ANSWER: 057 (1.00) g2ps B d.
REFERENCE:
SD023-740, page 8 of 74 KA: 006020A304 4.2/4.3 000011A111 4.2/4.2 006000K409 3.8/4.1 006020K403 3.2/3.6 006020K402 3.3/3.6 l-b t 4
,- -. ,-- - r, - ,-
QUESTION: 058 (1.00) Which of the following is the MINIMUM Shutdown Cooling System flow rate that can ensure adequate core cooling? , I j
- a. 1000 gpm
- b. 1600 gpm
- c. 2200 gpm
- d. 2800 gpm ANSWER: 058 (1. 00)- g2ps R 1
c.
REFERENCE:
SO23-3-2.6, 4.2.9 KA: 0060'0A401 3.7/3.6 4 e Y 4 T
QUESTION: 059 (1. 00) j Where does the Shutdown Coolir;g System (SDC) Low Temperature over Pressure relief valve discharge te?
- a. Containment Emergency Sump: '
- b. Auxiliary Building Equipment D: aln Sump
- c. Refoeling Water Storage Tank
- d. Reactor Drain Tank ANSWER: 059 (1.00) g2ps R a.
REFERENCE:
SD-SO23-740, page 38 of 74 KA: 006050A404 4.2/4.3 I f i b
QUESTION: 060 (1. 00) An RTD used in the Reactor Coolant System Hot Leg has failed duc to an open circuit in the hsat sensing portion of the detector. , l Which of the following are the correct symptoms for this failure?
- a. Delta-T goes high, T-ave goes low l l
- b. Delta-T goes low, T-ave goes high j
- c. Delta-T goes high, T-ave goes high
- d. Delta-T goes low, T-ave goes low ANSWER: (1.00) g2ps R c.
REFERENCE:
SO23-5-2.15, ARP50A05, page 14 KA: 002000K512 3.7/3.9
QUESTION: 061 (1.00) Plant startup and heatup is in progress. Which of the following Shutdown' Cooling System (SDCS) conditions are a prerequisite to alignment of Containment Spray _for automatic initiation?
- a. SDCS must be in operation
- b. SDCS must be operable and capable of maintaining the RCS heatup rate within administrative limits
- c. SDCS must be operable and capable of maintaining minimum RCS recircult. tion flow
- d. SDCS must be removed from service ANSWER: 061 (1.00) g2ps B d.
REFERENCE:
SO23-3-2.9 6.1, page 5 of 42 KA: 026000K101 4.2/4.2 i l i l l
f 0C / QUESTION: 062 (1.00) Which one of the'following safety Injection Tank (SIT) valves receives an automatic CLOSE signal.from the Safety Injection Actuation System (SIAS)? ,
- a. SIT Fill Valves (HV-9341, 9351, 9361 and 9371)' .;
- b. SIT Fill / Drain Valves (HV-9342, 9352, 9362, and 9372)-
- c. SIT Vent Valves (HV-9345, 9355, 9365 and 9375)
- d. SIT Nitrogen Valve (HV-9344, 9354, 9364 and 9374) '
ANSWER: 062 (1.00) g2ps B a.
REFERENCE:
SD-SO23-740, pages 27 - 29 KA: 006000K412 3.2/3.5 l - l i i i
. QUESTION: 063 (1. 00)
Which of the following is the MINIMUM coincidence logic required to trip the Main Turbine on MECHANICAL OVERSPEED?
-a. Two (2) out of four (4) overspeed trip signals
- b. One (1) out of four-(4) overspeed' trip signals
- c. Two (2) out of two-(2) overspeed trip r,ignals
- d. One (1) out of two (2) overspeed tr.i.p signals ANSWER: 063 (1.00) g3ps R d.
REFERENCE:
SD-SO23-180, p. 81 KA: 045000K413 2.6/2.8 < l e 4 d i a 4 e i s - e c - -w-
A i QUESTION: 064 (1.00) Electrical power has been lost to 480 VAC busses B04 and B06. How many Pressurizer heater banks have LOST POWER?
- a. 2 -
l
- b. 3
- c. 4 1
- d. 6 '
ANSWER: 064 (1.00) g2ps B-a.
REFERENCE:
SO 23 DWG 30118, 30120 KA: 010000K201 3.0/3.4. _l r
\ , ,,. - - ,- ~ --
'I QUESTION: 065' (1.00)
Which of.the following thermocouple (TC) conditions are used by- ; QSPDS to determine that Reactor Vessel level is BELOW a heated-unheated thermocouple pair?
- a. .The heated Tc indicates greater than 700 F
- b. The unheated TC indicates greater than 500 F
- 0. The-heated - unheated Tc temperature difference is i greater than 90 F i
- d. The heated - unheated Tc temperature difference is. .
greater than 200 F ANSWER: 065 (1.00) g2ps B
- d. ,
REFERENCE:
SD-SO23-820, page 52 KA: 002000K107 3.5/3.7 i I
+
4 I l-1 l l l
l l QUESTION: 066 (1.00) l Unit 2 is conducting a plant cooldown,-RCS temperature is 180 'F. Which of the following is the MAXIMUM cooldown rate allowed by 1 SO23-5-1.5 " Plant Shutdown from Hot Standby to Cold Shutdown"?
- a. 100 degrees F / hour
- b. 75 degrees F / hour
- c. 25 degrees F / hour
- d. 8 degrees F / hour ANSWER: 066 g2ps B b.
REFERENCE:
SO23-5-1.5, page 4 KA: 002000G010 3.4/3.9
QUESTION: 067 (1.00) l The following conditions are present: l l Channel "A" Low DNBR trip is bypassed RCP P001 (loop 1A) flow input to CPC Channel "B" has failed LOW l Which of the following conditions will result in a reactor trip?
- a. Pressurizer pressure input to Channel "C" CPC fails low
- b. Pressurizer pressure input to channel "B" CPC fails high
- c. -Loop 1 Hot Leg temperature input to Channel "A" CPC fails low
- d. Loop 1 Hot Leg temperature input to Channel "B" CPC fails high ANSWER: 067 (1.0) g2ps B a.
REFERENCE:
SD-SO23-710, paga 59 - 61 KA: 012000K50'. 3.3/3.8 012000K607 2.9/3.2 I 9 i
QUESTION: 069 (1.00) Operating Instruction SO23-3-1.7, Reacto Coolant Pump (RCP) Operations requires that the operator ensure that the temperature .; difference between the steam generator (S/G) and reactor coolant- ' system (RCS) cold leg (T S/G - T cold) be less than 20 degrees F ; prior to starting a Aeactor Coolant Pump. Which one of the following correctly states the reason.for minimizing the delta T between the steam generators and RCS when ' starting a RCP? a. Prevent lifting of steam generator safety valves,
- b. Prevent excessive heat transfer from the S/G to the RCS.
- c. Prevent RCP seal ^ damage.
- d. Prevent S/G Tube Thermal Shock.
- ANSWER 069 '(1.0) gp2ps R b.
REFERENCE:
- 1. SO23-3-1.7 KA 002020k501
, 3.2/3.6 p
L s
?
l l [ l
i l QUESTION: 070 (1.00) Immediately after a Reactor and~ Main Turbine trip at Unit 2,'an operator inadvertently opened the Main Generator output breakers BEFORE the automatic breaker trip signal occurred. 1 1 Which of the following describes the effects of this action on ! the 6.9 KV and NON-1E 4 KV busses? a.. They will be deenergized until manual action is'taken to re-energize them
- b. 6.9 KV busses will be deenergized, NON-1E 4 KV busses will ba pcwered from the Emergency Diesel Generators
- c. They will be energized from the Auxiliary Transformer d.- They will be energized from the Reserve Auxiliary ,
Transformer ANSWER: 070 (1.0) g3ps.B d.
REFERENCE:
l KA: 045000K301 2.8/3.1 i l-l
=
1 1
QUESTION: 071 (1.0) The Standard' Post Trip Actions,.SO23-32-1 (Step 5) requires the operator to " Verify RCS inventory control - satisfied." Which one (1) of the following Pressurizer level indications satisfies this criteria for Reactor Coolant System (RCS) Inventory Control, after a reactor trip event?
- a. 75 %, slowly decreasing trend.
- b. 65 %, slowly increasing trend. '
- c. 25 %, slowly decreasing trend.
- d. 15 %, slowly increasing trend.
ANSWER: 071 (1.0) epel B d.
REFERENCE:
SO23-12-1 p. 5 i KA 000007K103 4.0/4.6 t { i l l
QUESTION: 072 (1.0) l Standard Post Trip Actions, SO23-12-1, (Step 5) requires the operator to " verify RCS Heat Removal criteria - satisfied. " ' Which one (1) of the fol.'.owing conditions satisfy the criteria for adequate RCS Heat Removal Criteria?
- a. Steam generator levels 30 %.
- b. One main Feed water pump operating Reactor Trip Override NOT actuated. .
- c. Auxiliary Feed Water flow-300-gpm.
- d. Core Exit Saturation Temperature 15 degrees F.
ANSWER: 072 (1,0) epel B c.
REFERENCE:
SO23-12-1 p. 7 KSA 000074K103 4.5/4.9 000074K311 4.0/4.4 1 1
l QUESTION: 073, (1.0) Standard Post Trip Actions, SO23-12-1, (Step 4) requires the l cperator to " verify Vital Auxiliaries functioning. properly." 1 What is the correct electrical configuration of the 6.9 kV and non-1E 4 kV buses, under post trip conditions?
- a. 6.9 kV and non-1E 4 kV busses are energized from'the Unit Auxiliary transformer. l
- b. 6.9 kV busses are deenergized; non-1E 4 .V k buses are:
energized from the Auxiliary transformer.
- c. 6.9 kV and non 1E 4KV busses are energized from the Reserve Auxiliary Transformer.
- d. 6.9 kV buses are deenergized; non 1E 4KV busses are energized from Reserve Auxiliary Transformer. t ANSWER: 073 ( l'. 0 ) - epel B c.
REFERENCE:
SO23-12-1 p. 4 KA 000007G010 4.2/4.1 I ! a
QUESTION: 074 (1.0)- Unit 2 was operating at 100 % power when a-reactor trip occurred. I Which one (1) of the following plant conditions vill require yLa to close the Main Steam Isolation Valves?
- a. Condenser Vacuum decreasing to 7" Hg.
- b. Only one circulating water pump operating._
c.- Turbine speed 2000 rpm and not decreasing.
- d. Both main feedwater pumps trip.
ANSWER: 074 (1.0) epel B C.
REFERENCE:
SO23-12-1 p. 4 000074A208 J PW
QUESTION: 075 ( 1. 0) . Abnormal Operating Instruction, SO23-13-12, " Loss of Boron Concentration Control / Inadvertent Dilution," cautions that a , reduced letdown temperature will cause.a dilution effect in the RCS. - Why will a reduced letdown temperature cause an inadvertent dilution? !
- a. The Purification Ion Exchanger's boron saturation will NOT be in equilibrium. :
- b. The Purification Ion Exchanger filter media will. j release strainer, F-100.
CATION material which will block the letdown' ; c.- The Purification Ion Exchanger's boron concentration - will be super-saturated with boron. ! d.- The Purification Ion Exchanger filter media will i release strainer, ANION F-100.material which will block the letdown ANSWER 075 (1.0) epel B a.
REFERENCE:
i SO23-13-12, p. 2; SD-SO23-390 j KA 000024K104 2.8/3.6 1 i 1
QUESTION: 076 (1.0) The unit is in Mode'2 and you are emergency borating in accordance with SO23-13-11, Emergency Boration of Reactor Coolant System. Both boric acid makeup pumps trip,.and fail, and can rot be restarted. Both the RWST and BAMU tanks are above their minimum Technical Specification level. Which one (1) of the following correctly describes the preferred gravity feed method of completing the emergency boration, in accordance with SO23-13-11?
- a. From the Refueling Water Storage Tank (RWST) to the suction of the charging pumps. ,
- b. From the Boric Acid Makeup Tank (BAMU tank) to the suction-of the charging pumps .
- c. From the RWST to the LPSI pumps then to the suction of the charging pumps.
- d. From the Boric Acid Makeup Tank to the HPSI pumps then to the suction of the charging pumps.
ANSWER 076 (1.0) epe 1 B b.
REFERENCE:
SO23-13-12, p. 4; SD-SO23-390 KA 000024A201 3.8/4.1 I t
QUESTION: 077 (1.0) Unit 2 is in MODE 1 when the " Air Compressor Control Trouble Alarm" initiates. 'The Nitrogen Backup system begins-to supply the Instrument Air headers: What is the Instrument Air header-pressure at which the nitrogen back up system supply will begin to supply the instrument air header?
- a. 85 - 90 psig
- b. 75 - 80 psig
- c. 65 - 70 psig
- d. 55 - 60 psig 8 ANSWER 077 (1.0) epe2 B c.
REFERENCE:
SO23-13-5; LP: 2A0705 pp. 3&4 000065EK304 3.0/3.2 000065K308 3.7/3.9 i l l l 1 1 l 1 l l
QUESTION: 078 (1.0) Unit 2 is in MODE 1 when the " Air compressor control Trouble Alarm" initiates. The Nitrogen Backup Air system is maintaining the loads on the Instrument Air headers. Per SO23-13-5, " Loss of j Instrument Air": ' When would you expect to commence an " orderly plant shutdown to Hot Standby?" l l
- a. When normal system supply pressure cannot be restored.- l
- b. Immediately_upon trouble alarm confirmation.
- c. When one air operated valve starts to drift from its required position.
- d. One hour after the trouble alarm was received.
ANSWER 078 (1.0) epe2 B a.
REFERENCE:
SO23-13-5; LP: 2A0705 pp. 3-5 KA 000065A205 000065K308 3.7/3.9
i l QUESTION: 079 (1.0) l A spurious Safety Injection Actuation Signal (SIAS) and Containment Purge Isolation Signal (CPIS) has occurred. Operators in the Fuel Handling building notice pool temperatures rising. ) Why would Spent Fuel Pool temperatures rise after this event? l
- a. A SIAS would isolate the Component Cooling Water Non-critical loop supplying the Spent Fuel Pool Heat Exchangers.
- b. A SIAS would isolate the Nuclear Service Water Non-critical loop supplying the Spent Fuel-Pool Heat Exchangers.
- c. A CPIS would isolate the Nuclear Service Water cooling the Transfer Canal, subsequently heating the Spent Fuel Pool,
- d. A CPIS would isolate the Component Cooling Water Non-critical loop cooling the Transfer Canal, subsequently heating the Spent Fuel Pool.
ANSWER 079 (1.0) epe3 B a.
REFERENCE:
SO23-13-20; LP: 2A0720; SD-SO23-430, p.94 KA 000036A104 t l l l
E QUESTION: 080 (1.0) During refueling evolutions in MODE 6, you receive a " Reactor Vessel Pool Upper / Lower Seal Low -Pressure Alarm. " What is the NORMAL Reactor Vessel' Pool Seal supply source?
- a. Instrument Air
- b. _ Main Nitrogen Bank c.. Nitrogen Cylinders
- d. Site Service Air supply ANSWER 080 (1.0) epe3 B a.
REFERENCE:
SO23-13-20; LP: 2A0720; SD-SO23-430, p.94 KA 000036A104 3.1/3.7 1
QUESTION 081 (1.00) A Loss of Forced Circulation has occurred as a result of a Loss of Offsite Power. You are attempting to establish natural circulation in accordance with Emergency Operating Instruction SO23-12-7, Loss of Forced Circulation.
- Which one (1) of the following conditions will require a
" Response Not Obtained" action when performing verification of Natural Circulation?
- a. Loop delta T at'30 degrees F, and increasing slowly,
- b. Core Exit Saturation Margin at 10 degrees F, and increasing slowly.
- c. T hot / Representative Core Exit Thermocouples delta T at 5 degrees F, and increasing slowly.
- d. Reactor Vessel Plenum level above 82 %, and constant.
ANSWER: 081 (1.00) epe 1 B b.
REFERENCE:
- 1. EOI S023-12-7, Page 9, para. 10.
KSA 000015K101 4.4/4.6 l l 1 l l l
t I (; QUESTION 082- (1.00) } Which one (1) of the following is an entry condition for the Emergency Boration of the RCS Abnormal Operating Instruction SO23-13-11?
- a. Failure of one (1) CEA to drop following a reactor trip.
- b. Tave _15 degrees below Tre f due to a main .feedwater system transient.
- c. Shutdown margin of 4.15 % 'clelta K/K during_ Mode 5 Operation.
- d. Group 6 regulating CEAs below Power Dependent Insertion Limit.
ANSWER: 082 (1.00)-epel B d.
REFERENCE:
- 1. LESSON PLAN # 2A0711 OBJECTIVE 1.0
- 2. S023-13-11;0-2 KA 000024K301 4.1/4.4 m er
QUESTION 083 (1.00) Which one (1) of wi.e following conditions would require a immediate reactor and turbine trip per the Abnormal Opereting Instruction S023-13-7, entitled " Loss of Component Cooling Water or Salt Water Cooling"?
- a. Loss of CCW to the CEAs.
- b. One stage of a Reactor Coolant Pump (RCP) seal has failed on three (3) RCPs.
- c. A RCP thrust bearing temperature at 225 degrees F. ,
- d. One train of CCW fails on loss of CCW pump.
ANSWER: 083 (1.00) epel B
- c. 1
)
REFERENCEt
- 1. SO23-13-7 Page 4 KA 000026G010 4.0/4.2 000026K303 4.0/4.2 I
l b
QUESTION 084 (1.00) SONGS 2/3 EOI S023-12-5, " Excess Steam Demand Event", ellows operators to terminate or throttle High Pressure Safety A.7jection (HPSI) when. specific conditions are met. 1 Which one (1) of the following conditions meetc the criteria to terminate or throttle HPSI?
- a. Pressurizer level 35 %.
- b. Core Exit Saturation Eargin greater than 15 degrees F.
- c. Pressurizer pressure greater than 400 psia.
- d. Reactor Vessel plenum level 61 %.
ANSWERt 084 (1.00) epel B a.
REFERENCE:
- 1. EOI S023-12-5, PAGE 39 KA 000040A205 4.1/4.5 I
i
l QUESTION 085 (1.00) ( EOI S023-12-5, " Excess Steam Demand Event" provides criteria for determining that an excess steam demand is isolated. Which one (1) of the following is a valid indication that excess steam demand is isolated?
- a. Pressurizer pressure is decreasing. I 1
- b. Steam Generator indicated steam flow decreases to zero. l
- c. RCS Tc in each loop increasing.
- d. Both steam generator levels inc.reasing.
ANSWER: 085 (1.00) epel B : c.
REFERENCE:
- 1. EOI S023-12-5, PAGE 5, AND 15.
KA 000040G011 4.1/4.3
1 QUESTION 086 (1.00) Abnormal Operating Instruction S023-13-10, " Loss of condenser ! Vacuum" provides operator guidance for use of HS-2808B, Heat Treat Override Switch. 4 Which one (1) of the following is the MAXIMUM time that HS-2808-B may be placed in OVERRIDE during a loss of condenser vacuum? ;
- a. 5 minuta -
- b. 15 mir .itoa. ,
- c. 30 mint e
- d. 60 minutes.
1 ANSWER: 086 (1.00) epel B '
- c. '
REFERENCE:
i
- 1. AOI S023-13-10, PAGE 2 KA 000051A202 3.9/4.1
QUESTION 087 (1.00) One of the major objectivits of SO23-12-8, Station Blackout, is to establish a source of 4 K'1 electrical power. Which one (1) of the following is another major objectivc of SO23-12-8? a. Maintain possible. Pressurizer level greater than 5 % as long as
- b. Maintain steam generator levels less than 30 t.
c. Maintain shutdown for as long margin greater than 6.15 % delta K/K as possible. d. Maintain possible. single phase natural circulation as long as ANSWER: 087 (1.00) epel B d. REFEREHCE:
- 1. LESSON PLAN 2E0720 PAGE 5 KSA 000055G003 4.3/4.6
QUESTION 088 (1.00) Which one (1) of the following correctly describes why MSIS initiation will aid in mitigating the effects of Station Blackout?
- a. Prevents Steam Generator Safety Actuation. ,
- b. Prevents Damage to the High Pressure Feedwater heaters. l
- c. Prevents damage to main condenser.
- d. Prevents over feeding of steam generators.
ANSWER: 088 (1. 00) epel B c.
REFERENCE:
- 1. LESSON PLAN 2E0720 PAGE 6 KA 000055K302 4.3/4.6 ,
000055K204 3.7/4.1 '- i r b
QUESTION 089 (1.00) During Station Blackout, EOI S023-12-8, a CAUTION requires load reduction of Batteries D3 and D4 within thirty (30) minutes of the Station Blackout in order to extend battery life. Which one (1) of the following is the reason for extending Batteries D3 and D4 load life?
- a. Ensures RCP controlled bleedoff isolation valve HV-9216 operable for up to 8 hours.
- b. Ensures ADV operable on S/G with AFW flow up to 8 hours. '
- c. Ensures MSIV operable on S/G with AFW flow up to 8 hours.
- d. Ensures SDC suction valve operable from control room for up to 8 hours.
ANSWER: 089 (1.00) opel B d.
REFERENCE:
- 1. EOI S023-12-8 PAGE 3
- 2. LESSON PLAN 2E0720 PAGE 7 KSA 000055K302 4.3/4.6
QUESTION 090 (1.00) Loss of coolant accident procedure, SO23-12-3, requires shifting to hot leg injection two hours after initiation of safety injection. What is the reason for shifting to hot leg injection at this time , after a loss of coolant accident? i
- a. Ensure safety injection flow to the core in the event that the break is located in a cold leg.
b p. Return into solution boric acid which has become plated at the top of the core.
- c. Equalizes safety injection flow through the core to allow more even cooling,
- d. Ensures safety injection flow to the core in the event that the lower portion of the core has melted.
- ANSWER 090 (1.0) epc2 R b.
REFERENCE:
- 1. SO23-12-3 KA 0000llK313 3.8/4.2
QUESTION 091 (1.00) Operating Instruction S023-6-17, " Class 1E 120 VAC Vital Bus Power Supply System operation" requires that if High in-rush current loads are to be energized from the vital bus; then the Vital Bus supply should be transferred to the alternate source, and thea transferred back to the Vital Bus (inverter) once the high in-rush loads have been energized. Which one (1) of the following provides the reason.for this requirement?
- a. Avoids operation of fast acting current limiting feature of the inverter,
- b. Avoids reduction of inverter life expectancy due to transient overloading.
- c. Prevents feedback into the alternate source.
- d. Prevents failure of the standby inverter.
ANSWER: 091 (1.00) epel B f a.
REFERENCE:
- 1. OI S023-6-17, PAGE 3 KSA 000057A101 3.7/3.7
I QUESTION 092 (1.00) Which one (1) of the following plant staff positions is responsible for coordinating the fire fighting effort for a fire in the Unit 2 non-1E 4 KV electrical switch gear?
- a. Shift Superintendent.
- b. Unit 2 Control Room Supervisor.
- c. Unit 2 Control Operator.
- d. Unit 3 Control Operator.
ANSWER: 092 (1.00) epel B a.
REFERENCE:
- 1. SO23-13-21, PAGE 6
- 2. LESSON PLAN 2A0721, PAGE 5 KA 000067K304 3.3/4.1 i
i i
- - - - - - - - --- -)
QUESTION 093 (1.00) AOI S023-13-16, " Loss of. Containment Integrity", provides operator actions if containment integrity is lost. Which one (1) of the following conditions would NOT require implementation of the actions in SO23-13-16 if refueling operations were in progross?
- a. ONE door in each air lock becoUes inoperable.
- b. Equipment hatch being held closed by TWO (2) bolts.
- c. Containment Purge Isolation valve becomes inoperable,
- d. Containment Emergency Cooling Unit containment -
Isolation Valve becomes inoperable. ANSWER: 093 (1.00) coel B
)d1 &
REFERENCE:
- 1. LESSON PLAN 2A0716, PAGE 6 KA 000069G011 4.0/4.2
' t
(' N
i QUESTION 094 (1.00) Two of the three reactor trips initiated by the failure of Linear Power instrument JI-0002A, B, C, or D, are: l i Linear Power Level - High Local Power Density -(Lod) lii$ k What is the Third Reactor Trip affected by the loss of a linear i power instrument? '
- a. Log Power level - High
- b. DNBR - Low -
- c. Axial shape Index - Low
- d. Planar Radial Peaking Factor - High ,
- ANSWER 094 (1.0) epe2 B b.
i
REFERENCE:
- 1. SO23-13-18, P 2 & Attachment 1 -
- 2. SONGS 2/3 Technical Specificatior. 3.3.1 KA 000007K203 3.5/3.6 000007A105 4.0/4.1
t-QUESTION 095 (1.0) The failure of a Pressurizer narrow range pressure instrument , PT-0101- 1, 2, 3, or 4, will affect the " Pressurizer Pressure - High" Reactor Trip. What are the other TWO (2) Reactor Trips which will be affected by the loss of a Narrow Range Pressurizer Pressure Instrument?
- a. Pressurizer Pressure - Low (CCAS), and -
Pressurizer Pressure - Low RPS '
- b. Pressurizer Pressure - High (SIAS/CCAS), and :
Pressurizer Pressure - Low (RPS) '
- c. Local Power Density - Low DNBR - High ,.
- d. Local Power Density - High DNBR - Low '
- ANSWER 095 (1.0) epe2 B d.
REFERENCE:
- 1. SO9#-13-18, P2 & Attachment 1
- 2. SONGS 2/3 Technical Specification 3.3.1 j KA 000007K203 3.5/3.6 000007A105 4.0/4.1 i
i
QUESTION 096 (1.0) Which of the following choices is a means of determining whether a Reactor Protection System Channel failure involves single or redundant functional unit channels?
- a. Check the COLSS diagnostic printout.
- b. Check the Critical Functions Monitoring System (CFMS) diagnostic printout.
- c. Verify the Core Protection Calculator (CPC) matrix .
Power Supply indicating lights on the apron section of the CPC panels.
- d. Verify Logic Matrix Power Supply indicating lights inside the door on back of the PPS panel.
- ANSWER 096 (1.0) epe2 B
- d. or C, REFERENCE
- 1. SO23-13-18, p2 KA 000033G009 2.9/3.0 I
f' W l
QUESTION 097 (1.0) An inadvertent Safety Injection Actuation Signal (SIAS) occurred while Unit 2 was at 100 %. power. SO23-13-17, Recovery From Inadvertent Safety Injection / Containment Isolation, has been entered. This procedure requires you to override and stop the charging pumps. What action is required before 'n charging pumps can be restarted?
- a. Ensure that letdown is reestablished.
1
- b. Ensure that the pressurizer backup heaters are placed in override.
l
- c. Reset the-Reactor Trip Override.
- d. Reset the Safety Injection Actuation Signal.
- ANSWER 097 (1.0) epg2 R d.
REFERENCE:
- 1. SO23-13-17 p 3 '
KA 006050A201 3.9/4.2 i i i
QUESTION 098 (1.0) i The Steam Bypass Control System. (SBCS) receives a Reactor Coolant System Average Temperature (Tave) signal. l Which one (1) of the following correctly describes what the Tave signal to the SBCS is used for?
- a. Control the modulation of the Group "X" valves. ;
- b. Blocks the modulation of the Group "Y" valves. I o
- c. Block Quick open feature of SBCS. ,
- d. Provide Quick opening signal for SBCS.
- ANSWER 098 (1.0) g3ps R c.
REFERENCE:
- 1. SD-SO23-175 "
KA 041020K501 2.9/3.2 ' i i b i ( . i [ , l l l i l l 1 i i I i j
QUESTION 099 (1.0) The control room has been evacuated due to fire, you have been instructed to transfer the source range neutron flux monitor power from train "B" to train "A". Which one (1) of the following correctly describes where you l would obtain the key to operate the Train "A" source range ' neutron flux monitor transfer switch? 4
- a. Train "C" source range neutron flux monitor transfer '
switch located in Train "B" switchgear room..
- b. Train "B" source range neutron flux monitor transfer switch located in Train "B" switchgear room. j
- c. Safe Shutdown Locker, Control Room Supervisors Bag.
- d. Safe Shutdown Locker, Assistant Control Room Operators-Bag.
- ANSWER 099 (1,0) ope 2 R b.
REFERENCE:
- 1. SO23-13-2 KA 000032K201 2.7/3.1
+
d i t i
QUESTION 100 (1.0) ! SO23-13-13, Misaligned control Element Assembly (CEA), cautions the operator to limit the withdrawal of a misaligned or dropped , CEA to 5 inches per minute (equivalent to 6 steps per 10 seconds). Which one (1) of the following correctly describes why the withdrawal rate is limited to 5 inches per minute?
- a. To prevent Axial Shape Index fluctuations. ,
- b. To prevent Excessive Xenon transients in the area of the CEA.
- c. To minimize potential for fuel damage in the area of '
the misaligned CEA. !
- d. To minimize Quadrant Power Tilt fluctuations.
i
- ANSWER 100 (1.0) epe2 R
- c. ,
REFERENCE:
- 1. SO23-13-13 KA 000003K304 3.8/4.1 4
?
}
MM T1G. WEY ( 5'<0)
/ 'o,,
'E a UNITED STATES O@/(/AMf2-i ;l NUCLEAR REGULATORY COMMISSION ftEGON V
' ' " l e...*p wALYur"cn"tsKI$ALirNi Ises Senior Reactor ._ Operate-l WRITTEN EXAMlHATION ANSWER SHEET / POWER REACTOR: SONGS EXAMINATION DATE:
Region:
/ Reactor Type: CE V Passing Grade: 80%
Time LTmit: 4 hrs. - CANDIDATE'SNAME(pleaseprint): ' e MULTIPLE Example: .01:CH91CE A X C D MARK THE CORRECT ANSWER WITH AN X:
.01: A B C D .26: ABCD
.02: A B C D .51: ABCD .76: A B C D
.27: ABCD .52: ABCD
.03: A B C D .28: ABCD .77: ABCD
.04: A B C D .53: A B C D .78: A B C D
.29: ABCD .54: ABCD .79: ABCD
.05: ABCD .30: ABCD
.06: ABCD .55: ABCD .80: A B C D
.31: ABCD .56: ABCD .81: ABCD
.07: ABCD .32: ABCD
.08: ABCD .57: ABCD .02: ABCD
.33: ABCD .58: ABCD .83: ABCD '
.09: A B C D .34: ABCD
.10: A B C D .59: ABCD .84: ABCD
.35: ABCD
.11: ABCD .60: ABCD .85: ABCD
.12: A B '; D .36: ABCD .61: A B C D .86: A B C D
( .13: AdCD
.14: A B C D
.37: ABCD
.38: ABCD
.39: ABCD
.62: ABCD
.63: ABCD
.64: ABCD
.B7: ABCD
.88: ABCD t
.15: ABCD .40: ABCD .89: ABCD i
.16: A B C D .65: ABCD .90: A B C D
.41: ABCD .66: ABCD .91: ABCD
.17: ABCD .42: ABCD
.1B: ABCD .67: A B'C D .92: ABCD
.43: ABCD .68: ABC3 .93: ABCD
.19: ABCD .44: ABCD .69: ABCD
.20: A B C D .45: ABCD .94: 'ABCD
.21: A B C D .70: ABCD .95: ABCD i .46: ABCD .71: ABCD .96: ABCD
.22: ABCD .47: ABCD .72('A'8'C D'
.23: ABCD .48: ABCD .97: ABCD
.24: ABCD .73: ABCD .98: ABCD '
' .49: ABCD .74: ABCD .99: ABCD
.25: ABCD .50: ABCD .75: ABCD l .100: ABCD '
Total Points Available: Score: Overall Grade: % NOTE: : 1. Check that you responded to ALL questions, and that all responses are correctly marked ON THIS ANSWER SHEETl 2. Sign this answer sheet and turn in ALL exam materials to the NRC EXAMINER, l All work done or received on this examination is my own. I have neither given aid: ; Candidate's Signatu E b
i ES-201 ! Rev 5 01/01/89 i i ATTACHME T 2 (continued) ' i
. Enclosure 3
-PROCEDURES FOR THE ADMINISTRATION OF WRITTE'!
- 1. t Check identification badges. )
- 2. - Pass-out examinations.and all handouts. ' Remind applicants not t
examination until instructed to do so. ,
=. . -
READ THE FOLLOWING INSTRUCTIONS VERBATIM: During the administration of this, examination the following rules app + 1. and could result in more severe penalties. Cheating on the - t
- 2. .
After the examination has been completed, you must sign the statem . ' received or given assistance in completinthe cover sheet in done af ter you complete the examination. g the examination. This must be .have n i READ THE FOLLOWING INSTRUCTIONS l .
- 1. .
, Restroom leave. trips are to be limited and only one a " l room to avoid even the appearance or possibility of cheating.
- 2. .
o Use black ink or dark pencil only to facilitate legible reproductions , 2 the examination cover sheet. Print your name in the blank provid 4. e S. Fill in the date on the cover sheet of the examination (if neces . You separate may sheet write your answers on the examination question page or on of paper. ; THE BACK SIDE OF THE PAGE. USE ONLY THE PAPER PROVI ' 6. If you write your answers on the examination question page and more space to answer a specific question, use a separate sheet of the paper provided and insert it directly after the specific WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION DO NOT PAGE questio 7. section of your answer sheets whether you use pages or separate sheets of paper. Initial each page. f 8. ( Before sheet, intluding you turn in your examination, any additional pa consecutively number each an 1 on the examination question page. ges inserted when writing your answers . ( Examiner Standards s 14 of 18 ,
- . , - c..mo - --. , .~w.-- - , . . - - . . -y - - , --
ES-201 I' Rev 5 01/01/89 , (? ATTACHMENT 2 (Continued) ' ' Enclosure 3 (Continued) i
- 9. i l
If youi are1.04,6.10) number allow sp(ac.e,- using separate sheets, number each answer as' I e for grading. and skip at least 3 lines between answers to 10. 11.
. Write "End of Category _" at the end of your answers to a i w
Start each category on a new page. i 12 ' Write "Last Page" on the l'ast answer sheet, 13. Use abbreviations only if the Avoid using symbols such as <y are commonly used in facility _ literature. error resulting in an incorrect answer.or Write > signs it to out, avoid a simple transposition .,
- 14. i The point question. value, for each question is indicated in p'arenthese e The amount NOT an indication of theofdepth blankofspace answeron an examination question p required.
15. 16 Show all calculations, methods, or assumptions used r. to obta Partial credit may be given. i (3 (
/ 17.
AND-00 multiple choice NOTquestions. Proportional grading will be applied. LEAVE ANY ANSWER BLANK. Partial cred not be given on ' that is provided may count against you. Any Fo additional wrong information worth one point and asks for four responses,r example, if a question is t points, paints.and you give five responses, each ofeach which isof worth your 0.25 rresponse O.20 If one of your five responses is incorrect. 0.20 will be deducted and your total credit for that question will 1.00 even though you got the four correct answers. . be 0 80 in t s ead of ! 18. 19. If the intent of a question is uaclear,.,ast. questionsonly. of the exa When examination turning questions in your e examination, assenible the completed n with ex turn in all scrap pape,r.xamination aids and answer sheets. In addition, '
- 20. To pass the examination, greater and at least 70% you in each must achieve an ovnall grade of 80% or category.
21. There is a time limit of (6) hours for completion of the examina
- 22. (or some other time if less than the full examinatio .
When you are done and have turned in your examination , area (DEFINE THE AREA).
- 23. is still in progress, your If you are found license may be in this denied area, while the examina or revoked 5 ,
Ensure that all information you wish to have evaluated . r of as pa your answer is on your answer sheet. Scrap paper will be dipo k.) without review immediatly following the examination. Examiner Standards 15 of 18
l QUESTION: 001 (1.0) According to 10 CFR 20, which ONE (1) of the following correctly lists the requirements which must be met if the 1-1/4 Rem /Qtr whole body dose limit is to be exceeded?
- a. Cannot exceed 3' Rem /Qtr whole body, 5(N-18) Rem whole ;
body . lifetime, and NRC Form 4 completed
- b. Cannot exceed 5 Rem /Qtr whole body, 5(N-18) Rem whole body lifetime, and NRC Form 4 completed i
- c. Cannot exceed 3 Rem /Qtr whole body, 3(N-18) Rem whole body lifetime, and NRC Form 4 completed
- d. Cannot exceed 3 Rem /Qtr whole body, 3(N--18) Rem whole body lifetime '
ANSWER: 001 (1.0) pwg B
- a. (1.0)
REFERENCE:
- 1. . 10 CFR CH. 1 (1-1-89), Part 20.101 '
KSA 194001K103 2.8/3.4 i , S l l l l l l
QUESTION: 002 (1.0) SONGS Procedure SO123-0-10, you to complete Shift Relief Status sheets." Operations Shift Relief", requires Which ONE (1) of the following correctly de. scribes when the Shift Relief Status sheets should be completed?
- a. One hour prior to shift turnover
- b. Continually during the shift
- c. During shift turnover
- d. 1-2 hours following shift relief ANSWER: 002 (1.0) pwg B
- b. [1.0) or a,
REFERENCE:
1. S0123-0-10 " Operations Shift ReliefH, TCNO-15, p.4 KSA 194001A106 3.4/3,4 \
QUESTION: 003 (1.0) SO 123-0-15 " Control Room Access and Conduct", describes the control room manning and responsibilities during a Loss of Coolant Accident (LOCA). Which ONE (1) of the following personnel, by title, is responsible, per the procedure, for the control room command function, during a LOCA?
- a. The SRO officially stationed in the control room
- b. The senior operations manager on site d
- c. The senior control operator
- d. The shift superintendent .
ANSWER: 003 (1.0) pwg B
- d. (1.0) or 4.
REFERENCE:
- 1. SO123-0-15 " Control Room Access and Conduct", TCNO-3, p.7 KSA 194001A109 2.7/3.9 4
i = l 4
I I I i QUESTION: 004 (1.0) ] 1 In accordance with SO 123-7-9.9, " Radiation Exposure Permit Program", an extended Radiation Exposure Permit (REP) is issued to cover work involving relatively low radiological hazards. i Which ONE (1) of the following general area dose rates, from the selections given is the MAXIMUM for an area to be considered ( relatively low radiological hazard? a :
- a. 50 Mr/Hr
- b. 75 Mr/Hr 1
- c. 95 Mr/Hr
- d. 150 Mr/Hr ANSWER: 004 (1.0) pwg B
- c. (1,0)
REFERENCE:
1. SO123-7-9.9 " Radiation Exposure Permit Program", P.15 ' KSA 194001K104 3.3/3.5 l 4
)
l I 1 1
i QUESTION: 005 (1.0) S0123-0-11," Narrative Logs", requires specific entries be made in the Control operators log when specific types of systems or components are declared out of Service (00S). San Onofra's unit 2 Shutdown Cooling System (SDC) has been declared 00S. Which ONE (1) of the following symbols must be . entered in the margin of the Control Operators Log?
- a. Red E
- b. Black E
- c. Red OOS ,
i
- d. Black OOS !
ANSWER: 005 (1.0) pwg B
- b. (1.0) er 4.
REFERENCE:
- 1. SO123-0-11 " Narrative Logs", TCN1-1, p.12 KSA 194001A106 3.4/3.4 !
i
)
l l
\
h l l l
)
i
QUESTION: 006 (1.0) A system alignment.is being performed that requires independent verification. stem valve inAn itsinitial operator required hasposition. throttled placed a throttled rising Which ONE (1) of the following best describes the actions a SECOND. operator should take to perform an independent verification?
- a. Shut the valve completely and place in throttled position
- b. Open the valve completely and place in throttled position c.
Observe that the valve is in an intermediate position d. Operate an intermediate the valve only as necessary to verify it is in position i I ANSWER: 006 (1.0) pwg B
- c. (1.0)
REFERENCE:
- 1. SO123-0-23 " Control of System Alignments",TCNO-7,p.10 KSA 194001K101 3.6/3.7 t
~ - - ~ ^ ^ ^^
QUESTION 007 (1.0) Which ONE.(1) key.being taken outside of the protected area?of the following personnel can auth
- a. The shift superintendent
- b. The Central Alarm Station officer i
- c. The security shift commander
- d. The SRO operations supervisor ANSWER: 007 (1.0) pwg B
- c. (1.0)
REFERENCEt
- 1. SO123-0-27 " Key Control", p.4 KSA 194001K105 3.1/3.4 e
l QUESTION: 008 (1.0) Valve Operation procedure, 80123-0-13.1, allows the SRO operations. a manual globesupervisor
~ valve.. to approve the use of a leverage device on Which ONE .
leverage device that can be authorized?(1) of the- following is the maximum l,length
- a. As long as the handwheel radius
- b. 2 times as long as the handwheel radius
- c. 2-1/2 times as long as the bandwheel radius
- d. 3 times as long as the handvheel radius i i
ANSWER: 008 (1. 0) pwg B
- c. [1.0)
REFERENCE:
l i 1. SO123-0-23.1 " Valve Operation",TCNO-7,p.3 KSA j 194001A112 3.1/4.1 I
\
M Na
QUESTION: 009 (1.0) , I
- f. San Onofre Unit 2 is scheduled to change from Mode 3 to Mode 2 at 10:00 AM on Monday.-A Return to Service System Alignment Requirement SAR)
Mode change.(At 8:00 AM on Monday the mode change is postponed until 11:00 AM on Tuesday. i Which ONE (1) of the following is required? i
- a. Re-evaluation and approval of the SAR
- b. Determine plant impact and initiate an LCO Action Requirement (LCOAR) c.
Initiate corrective action under Shift Supervisor Acceleraced Maintenance (SSAM)
- d. Initiate a Site Problem Report (SPR)
\
ANSWER: 009 (1,0) pwg S '
- a. (1. 0 ] '
REFERENCE:
- 1. SO123-0-23,TCNO-7,p.15 KSA 194001K101 3.6/3.7 KSA 294001K101 3.6/3.7 4
1 1 l QUESTION: 010 (1.0) SO23-13-21 describes the. duties ar.$ responsibilities of personnel during a fire. A fire is reported inside the protected area of San Onofre Unit 2. Which ONE ' 1 of the following correctly describes the duties of the common (co)ntrol Operator (41), after obtaining a copy of the zone evaluation?
- a. Remain in the control room, and coordinate the fire !
t fighting effort
- b. Remain in the control room, and act as technical advisor c.
Proceed to the scene, and act as fire brigade leader
- d. Proceed to the scene, and act as technical advisor ANSWER: 010 (1.0) pwg B
- d. (1.0)
REFEPPNCE:
- 1. SO23-13-21 " Fire, TCN1-5,p.6 KSA 194001K116 3.5/4.2 Y
I 1
I QUESTION: 011 (1.0) ' 1 l It'is 2:00 AM.in the morning on Sunday.. San Onofre Unit 2 is in ' Made 1 with minimum shift. manning.-One (1)' Reactor Operator (RO) assigned to Unit 2 Control room becomes sick and leaves the site. Which ONE (1) of the following is the maximum time that Unit 2 shift may be undermanned before action must be taken? E
- a. One (1) hour i
- b. Two (2) hours
- c. Three (3) hours
- 0. :c2r (4) hourc ANSWER: 011 (1.0) pwg B
- b. (1.0)
REFERENCE:
- 1. SO123-0-30, TCN 0-7,P.3 l KSA 194001A103 2.5/3.4 1
l R l l l l i l l 1
\
QUESTION: 012 (1.0) Which ONE (1) of.the following components should be marked with an orange triangular. shaped emblem adjacent.to its controls? ia. Steam Bypass Control System Valve 8423
~
- b. Train.A Chargin Pump P190
- c. Containment Spray Pump P190 ',
- d. RCP 9318 Controlled'Bleedoff Containment Isolation Valve HV-ANSWER: 012 (1.0) pwg B i b.jfI(1.0)
REFERENCE:
- 1. SO23-0-17, TCN 10-14, p. 4 )
KSA 194001K101'3.6/3.7 i I et i t 4 L/
g 1 1 I QUESTION: 013 (1.0) Which ONE (1) of the following statements is correct concerning the work authorization;""Long Term Permission"?
- a. Issued for activities where the probability of the hazards are riot likely to change b.
Issued for change activities'which-involve during the work activity hazards that.could ' c. Issued for activities where the probability of the hazard is of great risk
- d. Issued by the Shift Superintendent only i f a clearance exists ANSWER: 013 (1.0) pwg S
- a. (1.0)
REFERENCE:
-i
- 1. SO123-0-21, Rev. 1, TCN 1-6, p.58 KSA 194001K102 3.7/4.1 G
l
QUESTION: 014 (1.0) SO 123-0-21 " Equipment Status Control", lists maintenance that requires a Work. Authorization Record (WAR) and maintenance that can be authorized verbally without a WAR. Which ONE.(1) without a WAR?of the following tasks can be authorized verbally
- a. Removal of lagging.on bt. , acid heat traced pipe
- b. Packing adjustment of valves requiring inservice tests i
- c. Vibration analysis t
- d. Lubrication of control-room emergency air system ANSWER: 014 (1.0) pwg B
- c. [1.0)
REFERENCE:
1. SO123-0-21, " Equipment Status Control"TCN 1-6, pp.34,35 KSA 194001A112 3.1/4.1 i e i
Ps , QUESTION: 015 (1.0) gd-hg b4 lN a SO123-0-21 " Equipment. Status Control" dictates which equipment asy'be~ tested with'thF circuit in a non test position. Of the following equipment, which OME (1) SHALL be. tested under an "In Test" position? Work Authorization with the circuit in the test
- a. Solenoid Operated Valves' i l
- b. 120 Volt Power Circuits t
- c. Power Receptacles i
- d. ACB Breakers ANSWER: 015 (1.0) pwg B 'i'.
- d. (1.0)
REFERENCE:
i' 1. SO123-0-21 " Equipment Status Control",Rev.1.TCN1-6,p.36 KSA 194001K107 3.6/4.1 ' a k i l N t s J l 1 iJ
.)
I QUESTION: 016 (1.0) San Onofre Unit 2 is in Mode 1. The following Reactor Coolant System (RCS);. chemistry is reported.
-!1 Chloride .10 PPM Fluoride.20 PPM Dissolved Oxygen .05 PPM
.)
Which ONE (1) of the following correctly lists the parameter or : parameters operation? that are out of specification for steady state
- a. Dissolved Oxygen, Fluoride
- b. Dissolved Oxygen, Chloride
- c. Chloride only
- d. Fluoride only 4 ANSWER: 016 (1.0) pwg S
- d. (1.0)
I
REFERENCE:
- 1. Technical Specification 3.4.6 KSA 194001A114 2.5/2.9 KSA 294001A114 2.5/2.9 f
J t I b
5 QUESTION:'017 (1. 0)-: The following. plant conditions exist: Unit 2- Mode?4, 345 F You are'on shift at' Unit 2.. Unit 2 experiences. severe ground-i s haking due to a major earthquake in the vicinity. As a, result of damage, the plant can be maintained at a temperature.of 350 F, j but cannot be cooled down for at least 12 hours until repairs can < be made. -The seismic monitoring alarmas apparently were rendered ! inoperable by the earthquake.. Which'ONE (1) of the following is the correct emergency class, at a MINIMUM, that you should declare as an operator in UNIT.27 .;
- a. Unusual Event &
- b. Alert
- c. Site Area Emergency
- d. General Area' Emergency i ANSWER: 017 (1.0) pwg B
- b. (1.0)
REFERENCE
- 1. SO23-8-1," Recognition and Classification'of Emergencies", TCN7-1,p.35 ,
KSA 194001A116 3.1/3.5 t
=- ,
HUCl. EAR GENERATION SITE .
' UNITS 2 AND 3 EPIP S023-VIII-1 ,
4 REVISION 7 PAGE 40 0F 60 .
- TCH '7 -/
(. DISASTER ATTACHMENT 2"
.. i UNUSUAL EVENT. -
TAB E1 -
- 1. For Modes 1-6: .
l Fire within the protected area which is not brought under control - within 10' minutes after verification. NOTE 1:
-A fire will be determined to be under control by the '
fire department incident commander at the scene. Emergency declaration should proceed unless.this T) 9 N j determination has been made within the indicated time. 2: See Event Codes 02-4 and D3-4 for Control Room evacuation. ,
- 2. For Modes 1 4: .
An earthquake Activation" alarm.causing activation of 61C21, " Seismic Recording System NOTE: Declare an Unusual Event for units 1, 2 and 3, if any unit receives a valid seismic trigger alarm and that unit, or any other unit, is in modes 1-4. {
- 3. F.or Modes 1-2:
A natural disaster, including hurricane, tornado, tsunami or flooding causing inoperability of, any system listed in Attachment 4 -to .the applicable Technical Specification. , , extent that reac
- 4. For Modes 1-2:
w A manmade disaster, including explosion, train derailment,. aircraft or missile impact, or toxic or flammable gas release,. causing i inoperability of any system listed in Attachment 4 to the extent that-reactor shutdown Technical has been initiated as.specified by the applicable Specification. 4 NOTE: Occurrence of. any natural or manmade disaster should be immediately evaluated for impact on plant. components or operation. Emergency classification of these occurrences should be made .under applicable El event codes if the -Emergency Coordinator judges the impact to be significant or to warrant emergency notification of offsite authorities, even though the explicit criteria of the event code may not be met. ATTACHMENT 2- PAGE 35 0F 49 .
NUCLEAR GENERATION SITE UNITS 2-AND 3 -EPIP . 5023 Vill-1
) . ?- - x: - .= =.- .=-. - REVISION 7
- ~- = u = ; .PAGE 41 0F 60
.. .- -. TCN r] - I_
4
=a.
DISASTER
- . m._ . .. ,,,-,. ., . ._ . . . -
a : . = a m o :. c . _-- . r _.. ATTACHMENT - , - ~ - - - ..,_,. . 4
- . 3 131'.._.. . ..c s . . . ifAB .E2 L For Modes 1-6:
An earthquake greater than .339 ground acceleration. by al arm, initiation y-.,.. of-61C22, " Operating Basis Earthquake Acceleration"This is
.. . ~ - ~ . _ . . - - - - . c. .. .
2.- For Modes 4 6:. . L A natural disaster. including hurricane, tornado. tsunami'or' flooding i causing the loss of ability to. achieve or maintain cold shutdown. I 1 NOTE: See Event Code E31 for loss of hot! shutdown capability.
. . . . . . . . {
- 3. For Modes 4-6:-
A manmade disaster, including fire, explosion, aircraft or missile impact, or toxic or flammable gas release causi to achieve.or.. maintain. cold shutdown..-. . . _._ ng the loss of ability ( . NOTE 1: See Event Code E3-2 for loss of hot shutdown-capability. 2:
, See Event Codes 02-4 and.03-4 for Control' Room-evacuation, t
=
NOTE: The loss of ability to achieve or maintain cold shutdown is based on the following:
.i: . :
For Modes 4-6: i.
,. (a) Loss of shutdown margin - actual percent shutdown less than 1% ak/k.
0.8 (b) Inoperability of steam generators or shutdown cooling system resulting in: Uncontrolled RCS temperature increase to
> 200*F , / Inability to reduce or maintain RCS temperature 1 . _.
to < 200*F
~
ATTACHMENT 2 PAGE 36 0F 49 -
.i'
NUCt. EAR GENERATION SITE UNITS 2 Afl0 3? EPIP' S023-Vill 1 s- REVISION 7 PAGE 42 0F 60 a .a.:- -u ..- . - . . . . . . - TCH 7-/ DISASTER ATTACHMENT 2 SITE AREA EMERGENCY TAB E3
~
l t
- 1. For, Modes 1-3:
A natural'disaster, including earthquake, hurricane, tornado, tsunami l . or flooding causing the loss of ability to achieve or maintain hot shutdown.-
}
- 2. For Mod 6: 1 -3 :'
A.. manmade _ disaster,.Jncluding. f. ire... explosion, aircraft or missile impact, or toxic or flammable gas release causing the. loss of ability to achieve or maintain hot shutdown. NOTE: See Event Codes 02-4 and D3-4 for Control Room evacuation. NOTE: The" loss .of ability to achieve or maintain hot' shutdown is based on the following: ! For Modes 1-3: (a) Shutdown margin < 1% ak/kl l i
.....,..._.._____.......3;....._ t (b) Inoperability of steam generators or shutdown-cooling system resulting in: .
i
,. Uncontrolled RCf temperature increase to
> 350*F l
l E Inability to < 350'F to maintain or reduce RCS tempt.rature ! ATTACHMENT 2 PAGE 37 0F 49 I
NUCLEAR GENERATION' SITE-UNITS 2-AND 3 EPIP
.>-- S023-Vill 1
;....... - .... . REVISIO
...-w......- PAGE 43 0F 601
.TCN 3 ~N /' 71 . . . . _
^
. .U. _ _..._ _.._ _ ..,. . ._ DISASTER
_ ATTACHMEHL2
--GENERAL.- EMERGENCY _
... TAB E4
- 1. For Modes'l 4:
For t.oss 'of' Coolant Accident sequences: - i (a) The loss of 2 of 3 fission' the following 3 conditions: product barriers indicated by any 2 of (1) Any loss of coolant accident requiring or resulting_ in ECCS! actuation ' (2) The loss of containment integrity .
-(3)- Probable significant fuel damage. )
4 i, p)_A,n i,ncreased_ po, tent,ia,1 for_ loss of the third fission product barrie.r. NOTE 1: i Significant the following: fuel-damage may be indicated by any of
.. (a) Sample analysis of the RCS, ' indicating the re-lease of fission products-to the primary coolant
> Technical Specification 3.'4.7 limits .
(b) Indications listed in E01 safety-function status
- checks for core heat . removal ,
(c) Indications listed.in Chemistry Procedure
.. e S0123-111-8.8, Parameter Sampling. Alternate Methods of Post-Accident 2:
The by any 'sss of containraent integrity can be determined of the following: i 4 (a) Known breach of containment (b) components Status indication of containment isolation (c) A decrease in containment pressure not due to containment spray or cooling systems.
^
ATTACHMENT 2 PAGE 38 0F 49 I
J NUCLEAR GENERATION SITE-UNITS 2 AND-3 EPIP 5023-VI11-1
>: REVISION 7 PAGE 44 0F 60 ,
TCH 'I -l DISASTER = ATTACHMENT 2 GENERAL EMERGENCY.- TAB'E4 l
^
2.- -For Modes 1-4: For Steam Generator Tube Rupture sequences: (a) A steam generator tube rupture requiring or resulting in ECCS-l actuation ' (b) Either one'of the following 'two conditions, with an increased potential for the second: (1) Potential significant fuel damage i (2) . An active or potential flowpath for release of fission product gases to the atmosphere. t NOTE: Significant fuel. damage may be indicated by any of ~ the following: (a) Sample analysis of the RCS, indicating the re-lease of fission products to the primary coolant
> Technical Specification 3.4.7 limits t
(b) Indications listed in E01 safety function status checks for' core heat removal
-(c)-Indications listed in Chemistry Procedure-S0123-III-8.8, Alternate Methods of LPost-Accident Parameter Sampling.
s-
~
ATTACHMENT 2 ~ PAGE 39 0F 49 i
.:n QUESTION: 018 (1.0)
A LOCA,has n o,ccurred..andA he.following plant. conditions exist. . Reactor Coolant Pressure 1850 psia Containment Pressure ~ 3.5 psig Steani'GenerEtoY ~Prssiuies ~" ' "810 psig
' ' - ' " - ~ * ^
Pressurizer level- 10%. Which one of the plant conditions will cause an Engineered i Safety Features Actuation Signal-(ESFAS)?
- a. Reactor Coolant Pressure i
- b. Containment Pressure
- c. Steam Generator Pressures d.fPressurizer level.
i ANSWER: 018 (1.0) epe2 B
- b. (1. Oj _ _ _ . . ,
Reference:
- 1. SD-SO23-720 p. 6
- 2. SD-SO23-740 p. 10' KSA 000011A104 4.4/4.4 l
1 1 1 l i
Y QUESTION 019 (1.0) Which one (1) of the.following is the MINIMUM plant conditions ; that will cause a Containment Spray Actuation Signal (CSAS)?
- a. Containment Pressure 3.25 psig'and a Pressurizer level of 10%.
- b. Containment Pressure 8.3 psig'and a Pressurizer Pressure -
of 1800 psia.
- c. Containment Pressure 12.25 psig and a pressurizer level of 10%. t l d. Containment Pressure 16.3 psig and a pressurizer pressure-l of 1800 psia.
- ANSWER: 019 (1.0) Ope 2 B
- d. (1.0)
REFERENCE:
l l 1. SD-SO23-720 p. 13 l KSA 000009A113 4.4/4.4 I f
QUESTION 020 (1.0) The unit is operating at 100% power with charging pump P-191 in- ' operation and with charging pump P-192 selected as first backup when the NON'1E Uninteruptable Power Source (UPS) is lost. Which one (1) of the following statements correctly describes the response of the Chemical and Volume Control System? a.- Charging pump P-192 will start due to a simulated low pressurizer level.
- b. Charging pump P-191 will stop and charging pump P-192' will start to maintain-pressurizer level,
- c. Charging pump P-191 will stop due to a simulated high pressurizer level,
- d. Charging pump P-192 and P-190 will start due to a simulated low pressurizer level.
ANSWER 020 (1.0) glps B
- c. (1.0)
REFERENCE:
- 1. SO23-3-2.1 p 5 (4.11)
KSA 004000K101 i i ,
i QUESTION 021 (1.0) The unit is operating at power with the CEAs in Manual Sequential. Which of the following PRE-TRIPS would prevent you from withdrawing the.CEAs?. ,
- a. 1 of 4 - Low pressurizer pressure,
- b. 2 of 4 Low pressurizer pressure,
- c. 1 of 4 Low DNBR. i
- d. 2 of 4 . Low ~ DNBR.
ANSWER 021 (1.0) g1ps B.
- d. (1.0) or c, !
REFERENCE:
- 1. SD-SO23-510, p 38' KSAs 001000K104 3.2/3.4 001000K105 4.5/4.4 l
l l 1
.l l
N 4
_. . ._. , . - . _ ._. . . . . . .. . . .~. 1 l QUESTION 022 (1.0) ! i It-is desired-to raise Tave 5 degrees by changing boron concentration. , The following plant conditions exist:' Reactor Power: 70% a CEA Positions: All CEAs Full Out
'RCS Boron concentration:' I 730 ppm
' -Total Power Coefficient' O.02 delta rho / percent power- , Isothermal Temperature -0.010 percent < Coefficient: delta rho /. degree F Inverse boron worth:: 70 ppm / percent delta rho 1 Which one (1) of the following is the final RCS boron concentration required to increase.Tave by 5:(five) degrees F WITHOUT. changing reactor power?.
- a. 721.0 ppm
- b. 723.5 ppm
- c. ~725.0 ppm I
- d. 726.5 ppm ANSWER
- 022 (1.0) g1ps B
- d. [1.0) calculation: ((-0.010 x 5)*(70)) = - 3.5 726.5 ppm
, 730 - 3.5 =
REFERENCE:
1
- 1. SONGS Reactor Theory P. 176, 196
- KSA 004000A404 3.2/3.6 i 001010K518 3.2/3.6 L
l J
1 l i i i 1 023 (1.0) 1 QUESTION The unit is operating at 100% power in mode 1 when a spurious containment spray Actuation Signal (CSAS) occurs. Which one (1) of the following correctly describes what occurs as a result of the CSAS? ,
- a. Containment Spray pumps receive an auto start signal, pump recirculation valves open.
- b. Containment spray pumps do NOT receive an auto start ;
signal, pump suction and recirculation valves open. '
- c. Containment spray pumps do NOT' receive an auto start signal', spray header isolation valves open.
- d. Containment spray pumps receive an auto start signal, spray header isolation valves open. i 1
ANSWER: 023 (1.0)~g1ps B
- c. (1.0) l
REFERENCE:
l 1. SD-SO23-740, p 30, 33 KSAs 01300,0K105 4.1/4.4 L I w
I 1 1 QUESTION 024 (1,0) i 1 Technical Specification 3.1.1.1-requires a specific shutdown margin be maintained during modes 1,-2, 3, and'4. Which one (1) of the following is the MINIMUM Shutdown Margin required by Technical Specifications 3.1.1.1 for mode 1? >
- a. 15.15_% delta k/k.
- b. 5.15 %~ delta k/k. t
- c. 4T15 % delta k/k.
- d. 3.0 % delta k/k.
ANSWER: 024 (1.0) g1ps B
- b. (1.0)
REFERENCE:
- 1. SO23-3-3.29 6-5 2.. SONGS 2/3 Technical Specification 3.1.1.1-KSA 001010K535 3.3/3.6 l
I l l l i 1
QUESTION-025 (1.0) ! Which one of the following conditions is the basis for the ' Shutdown Margin requirements for Modes 1-4 as specified in.
' Technical Specifications?
l
.a.. BOL, Full load operating Tave,-Loss of Feedwater. ,
- b. BOL, No loadJ operating. Tave, Main Steam Line Break. ,
- c. EOL, Full Load operating Tave, Loss of.Feedwater.
- d. EOL, No load operating Tave, Main Steam Line Break.
ANSWER:: 025 (1.0) gips S .
- d. [1.0)
REFERENCE:
- 1. SO23-3-3.29
- 2. SONGS 2/3 Technical Specification Bases 3/4.1.1.1. ;
KSA 001000K508 3.9/4.4 u
.~
t
I 1 1 QUESTION 026 (1.0) The reactor protection syster requires a minimum 2 (tw )) channels of reactor trip signals to cause a loss of power to the CEDM coils. See attached drawing. Intich one of the following combinations of the trip circuit will cause a 3oss of power to all of the CEDM coils?
- a. SSR K1 and SSR K2 energized.' '
- b. SSR K1 and SSR K2 deenergized,
- c. SSR K3 and SSR K2 energized.
- d. SSR K3 and SSR K2 deenergized.
ANSWER: 0 (1.0) gips B
- d. (1.0)
REFERENCE:
SD-SO23-710 p 60 KSA 001000K105 4.5/4.4 ; 001000K202 3.6/3.7 ! 4
, e-,~ , v ,
O .~- m .O
- Q '. ._- ..;
r*
- 480 v REACTOR TRIP 4
sus i 4eo v STATUS , sus 2 '2 820 VAC M M VITAL -120 vaC - BtJS 2 G g , ygynt 4 gy 3
? g 2
MI .l , POWER .M2 1 POWER SUPPLY TCB 9 ' SUPPLY , z -- 4 r r I TCB 2 TCB 6 Tce3 r 2 - - - - - - - - - - TCa7 I . 2 (w-) x2 --------- x3 (,) r z
' PHASE CURRENT
- I I I
[ I I I 4 l- :l l l .I 1 . . I , . i y TCBI TCB S - TC8 4 . x (w}xi ---------- TCB S K4 (/ y I-4 z
' I z ,
t I {. POWER s ' POWER-SUPPLY I SUPPLY 120 VAC 'l 2 '3 4
- VITAL B20 VAC -
CEDM U/y VITAL' [ AtlS I : sus 4 1 I
<- - - - + .wsv. e a-.--- - % 'nEg-_. .a s-,-
+y - - . , _ - - . + , - , - , , , _ _ _ _ _ _ .
- n. _ ..
QUESTION 027 (1.0) Which one (1) -of th'e 'following excore nuclear instrumentation channela is required to be used during a " Shutdown from outside the Control Room," SO23-13-2?
- a. -A linear power channel,
- b. A logarithmic power channel,
- c. A control channel.
, d. _ A startup channel.
1 ANSWER: 027 (1.0) gips B-
- d. [1.0) or b.
REFERENCE:
- 1. SD-SO23-470 p 4 KSA 015000K403 3.9/4.0 e
l
QUESTION 028 (1.0)- l What'is the purpose of-providingLTemperature Shadowing ~ l - Compensation for_the excore detector inputs to the core-protection calculators (CPCs)? a .' Account for reduced neutron leakage from the core at
-lower temperatures.
t
- b. Account for-.reducedLneutron leakage from the core at higher temperatures.
L
- c. Compensates for increased detector output'due to lower &
temperature in detector-well. . i
- d. Compensates.for increased neutron detector output due to. higher temperature in detector well.
- ANSWER: 028 (1.0) 91p3 B
- a. [1.0]
REFERENCE:
- 1. SD-SO23-710
. KSA 015000K502 2.7/2.9 4
l i l l
l ! QUESTION-029 (1.0) ! Plant operating. instruction SO23-3-2.1, CVCS Charging and ! Letdown, cautions operators to establish letdown immediately. after initiating charging. i What is the reason for ensuring that letdown is established immediately after initiating charging?. ! 1 a.- . Prevent thermal shock of the VCT. '
- b. >
Prevent an inadvertent boron-dilution accident.
- c. Prevent thermal shock to the~non regenerative heat exchanger.
i d. Prevent thermal shock to the RCS loop' inlet nozzles. ' I, ANSWER: 029 (1,0) gips B
- d. (1.0)
REFERENCE:
i
- 1. SO23-3-2.1 p 5, 4.7 KSA 004000K511- 3.6/3.9 i
.l i
4
- ~ QUESTION 000 (1.0) The reactor operator reports to you that both of tha control room airborne monitors RE-7824-1 and RE-?825-2 have failed low. The I&Cdetectors. the technician' indicates that it will take five days to replace What action must be taken (in accordance with Technical Specificati,4, 3.3.2 and 3.3.3.1) within one (1) hour? a. Return the monitors to service, or be in Hot Standby within six (6) hours.
- b. j Initiate and maintain operation of control room emergency air cleanup system.
c. Establish a portable control room gaseous process radiation monitoring station. d. Perform a surveillance emergency test on the control room air cleanup system. k ANSWER: 030 (1.0) glps B
- b. (1.0)
REFERENCE:
- 1. SdNGS 2/3 Technical Specifications 3.3.2 and 3.3.3.1
- 2. SD-SO23-700 p 37 KSA 072000A202 2.8/2.9 072000 GEN 2.1/3.4
8 QUESTION 031 (1.0) A loss of all AC power has occurred in Unit 2. Which one '(1) of the following auxiliary feedwater valves will be able-to be electrically controlled using DC power? See attached drawing.
- a. Auxiliary Feedwater pump 2HV-4762.
P-504 discharge bypass valve
- b. Auxiliary Feedwater pump P-140 discharge valve 2HV-4705.
- c. Auxiliary Feedwater pump P-504 4712. discharge valve 2HV-
- d. Auxiliary Feedwater Pump P 141 discharge valve 2HV-4713.
ANSWER: 031 (1.0) g1ps B
- b. (1.0)
REFERENCE:
- 1. SD-SO23-780 p 5, 6 KSA 061000A203 3.1/3,4 2
E r l
- 3. /u' e.
_ SIMPLIFIED DIAGRAM OF THE AUXILIARY FEEDWATER SYSTEM a MINI-FLOW TO CST T-121 e 1 __ HV-4713 p HV-4731',
!Mi-N-M ,$, - . , a i
!l n- P&" 55 '8 MINI-FLOW LO LC)[ d5 f-MOTOR
^
P-141 cst T-121 l b _! 1 "Wv 4706AW a HV-471d FWM 2 ~ l CST T-121 N ,, , 6 eM Q f t '
- OUTSIDE IrGIDE
=
LO # " LO #I EXHAUST TO = Amos "2"4 '"3 s Jo';Xe ,! KOO7 P 140 I
,_ -- h% _
% E4" uv n 3aj MAIN r
8200.@ _4716 k
=
- a W
STEAM rJ H24 y ucs - "v- 'Ca ~ . e FADM Q CST T-121 to l
,'M M ' '
,m p TO S/G d2 LD HV-4712 '
HV-471y T . > MOTOR Z P-504 MINI-FLOW
~
TO M T-121 FROM Q g , CST T-121 LO NH24 D , sw-mo-av,2Ar7 (f. p i . O.
l I i , QUESTION 032 (1.0) i Which one (1) of the following limits the flow of the TURBINE DRIVEN auxiliary feedwater pump to 1000 GPM in the event of a feedline break '_inside the containment? j
- a. Discharge valves close to 40% open on a high flow !
signal from FE-4720 or FE-4725.
- b. A cavitating venturi is installed on the discharge of the pump. ,
+
- c. Discharge valves are preset to open tE limit flow from.
the p. ump,,to 1000 GPM. , ,
- d. The flow signal from FE-4720 or FE-4725 will limit steam flow to the turbine.
ANSWER: 032 (1.0), epe2 B ,
- b. [1.0) ,
REFERENCE:
- 1. SD-S023-780 p. 5 ;
- 2. P&ID 40160A-15 KSA 000054A204 4.2/4.3 I 4
f, l. l t l l l 1
QUESTION 033 (1.0) Unit 2 has tripped and the following conditions are present:
, . Pressure Level Steam Generator E088 700 psig 10 %
Steem Generator E089 625 psig 10 % Which one (1) of the following statements correctly describes the response of the Auxiliary Feedwater system?
- a. Auxiliary feedwater will initiate to S/G E088 ONLY .
- b. Auxiliary feedwater will initiate to S/G E089 ONLY.
l
- c. Auxiliary feedwater will be initiated to BOTH S/Gs.
- d. Auxiliary feedwater will be initiated to NEITHER S/Gs.
ANSWER: 033 (1.0) g1ps B '
- a. (1.0)
REFERENCE:
- 1. SD-SO23-780 p 81 KSA 061000K414 3.5/3.7 l
r 9 4 l r
. . . . , , , . . - . . . , - , - . , . . _ _ , _ . , - .r. , -
QUESTION 034 (1. 0) Which one (1) of the following Engineered Safety Features i Actuation Systems (ESFAS) signals are generated by an AUTOMATIC Safety Injecti'on Actuation Signal (SIAS) but not by a MANUAL SIAS? i i
- a. . Containment Cooling Actuation Signal (CCAS).
- b. Main Steam Isolation Signal (MSIS). '
'c. Main Feedwater l
Isolation Signal (MFWIS).
- d. No'n Critical Loop CCW Isolation signal.
ANSWER: 034-(1.0) gips B
- a. (1.0)
REFERENCEt
- 1. SD-SO23-720 p 8 KSA. 013000K103 3.8/4.1 022000K403 3.6/4.0 f
J 1
- - . - - , - - , , - --,,-.,-.n , , , -- - - - - - - -
r QUESTION 035 (1.0) Unit 2 is in the process of a Hot Restart, at End of Life (EOL) in.accordance with plant operating instruction SO23-5-1.3.1. An Estimated- Critical Position (ECP) has been determined. The reactor operator has followed the procedure in pulling rods to ; the ECP but does not have indication that the reactor is critical. Which one (1) of the following conditions will cause the actual critical rod CEA position to be HIGHER than the Estimated ' Critical Position (ECP)? I
- a. Delaying the time of startup from 18 to 22 hours after the reactor tripped from steady state high power,
- b. Misadjustment of the steam bypass controller to be 50 psig higher than the normal no load setting.
c. ' An inadvertent system boronthe (RCS) during dilution of the reactor coolant rod withdrawal.
- d. Reactor coolant temperature is lower than assumed in the Estimated Critical Position calculation.
ANSWER: 035 (1.0) gips B
- b. ,
REFERENCE:
1, SONGS Reactor Theory p 181
- 2. SD-SO23-175 p 37 KSA 001010A207 3.6/4.2 a - - -
QUESTION 036 (1.0) The Area Radiation Monitoring System generates a signal to i perform specific isolation functions. Ubich one (1) of the following isolation functions can be i generated by the Area Radiation Monitoring System?
- a. Containment purge isolation.
- b. Containment isolation.
- c. Blowdown isolation.
- d. Fuel building ventilation isolation.
ANSWER: 036 (1.0) gips B
- a. i
REFERENCE:
- 1. SD-SO2$-690, p 29 KSA 072000K401 3.3/3.6 072000K403 3.2/3.6 l
l
- - + - , - ~ . ~ , , .
1 i QUESTION 037 (1,0) Technical Specification 3.9.2 requires that a specific number of source range neutron flux monitors be in service while in Mode 6. What actions must be taken if ONE (1) of the source range neutron flux monitors required by. Technical Specification 3.9.2 becomes ; inoperable?
- a. Suspend all fuel movements and emergency borate.
- b. With one required monitor inoperable fuel movement can continue for up to 12 hours if a boration flow path is established.
- c. Suspend all operatio.as that involve core alterations; fuel movement in reactor vessel is NOT allowed.
- d. Suspend all core alterations that add positive reactivity; only core unloading operations are allowed.
ANSWER: 037 (1.0) gips S c.
REFERENCE:
- 1. SONGS 2/3 Technical Specification 3.9.2 KSA 015020SG05 3.3/3.8 i
1 ---v - .-_i -.- - ~ - - . - - . .
r a QUESTION 038 (1.0) ' The' unit is operating at 60% reactor power. At 4:00'AM you observe that Tave has decreased to below 520 F and cannot be restored.
~~
Which
'at which onethe(1) u.of the following times correctly describes the time nit must be in HOT Standby if Tave can not be l
+
restored within Technical Specification limits?
- a. 4:30 AM
- b. 5:30 AM i
- c. 6:30 AM 1 i
- d. 10:30 AM' ANSWER:'038 (1.0) g2ps B
- a. i 4
REFERENCE:
- 1. SONGS 2/3 Technical Specifications 3.1.1.4 KSA 002000K510 3.6/4.1 002000A109 3.7/3.8 I
B
i
)
l QUESTION 039 (1.0) i Technical Specification 3.5.4 states that the maximum boron concentration ppm. of the Refueling Water Storage Tank (RWST) is 2800
]
1 Which one (1) of the following correctly states the Technical Specification in the RWST? bases for this maximum limit on boron concentration ; l
- a. Ensure that boron does not precipitate in the core '
following a LOCA.
- b. Prevent boron precipitation in RWST.
- c. Prevent precipitation of boron in the sump causing less borated water to be recirculated back to the core.
d. Ensures minimum shutdown boron concentration during a main steam line rupture. ANSWER: 039 (1.0) g2ps S a. REFERENCE i
- 1. SONGS 2/3 Technical Specification bases 3/4.5.4 KSA 00600dG006 2.9/4.0 020000K101 3.7/3.9 1
4
, r--- , -,- - - -
r yv-
QUESTION 040 (1.0) Due to a sudden turbino load reduction Group 6 regulating CEAs ' were manually driven into the core to maintain Tave within t'.ie prescribed operating banC. As a result of the CEAs being dr3ven in the following core power conditions exist: , Upper Half power 45 %
- Lower. Half. power 55 %
Which one (1).of the following is the correct Axial Shape Index (ASI) for the given core power conditions?
- a. + 0.2
- b. - 0.2
- c. + 0.1
- d. - 0.1 ANSWER: 040 (1.0) gips B
- c. ,
(55 - 45) / (55 + 45) = 0.10
REFERENCE:
- 1. SONGS 2/3 Technical Specification 3.2.7 & Definition KSA 015020K503 3.3/3.7 t
- w-- yw .. g m, , --- --
F QUESTION 041 (1.0) i COLSS has been computed.by.the out of service with Axial Shape Index (ASI) being CPCs. Which one (1) of the following is the correct ASI Technical J Specificatior. 3.2.7 limit when COLSS is out of service?
- a. -0.15 $ ASI S +0.15
- h. -0.20 $ ASI $ +0.20 ,
- c. ~0.25 s ASI s +0.25
- d. -0.27 s ASI s +0.27 ANSWER: 041 (1.0) gips S b.
REFERENCE:
- 1. SONGS 2/3 Technical Specification 3.2.7 KSA 015020K503 3.3/3.7 015020G005 3.3/3.8
\
e n
i i i l QUESTION 042 (1.0) l J Component Cooling Water (CCW) has been lost to a Reactor Coolant l Pump (RCP) while in Mode 1. You have entered OI SO23-13-6, RCP i Seal Failure, and AOI SO23-13-7 Loss of CCW / SWC. What is the MAXIMUM time that a Reactor Coolant Pump can be operated without Component Cooling Water (CCW)?
- a. 3 minutes '
- b. 5 minutes
- c. 15 minutes
- d. 30 minutes ANSWER: 042 (1.0) g3ps B
REFERENCE:
- 1. SO23-13-7 p3 KSA 008030K302 4.1/4.2 s
[ ( l I l
) 1
$ \'
l l
QUESTION 043 (1.0) i Component Cooling Water Pumps P-024 and P-026 are operating, P-025 is aligned to Train A and is in standby. A Safety Injection Actuation Signal (SIAS) and a Loss of Voltage Signal LOVS) (on both safety related busses for greater the's five secon(ds) have been received. l Which one of the following correctly describes the CCW pump ! responses when the busses are reenergized?
- a. P-024 and P-026 start P-025 does not start.
- b. P-025 and P-026 start P-024 does not start.
- c. P-024 and P-025 start P-026 does not start.
- d. All CCW pumps start.
ANSWER: 043 (1.0) g3ps B b.
REFERENCE:
- 1. SD-SO23-400 KSA 008030K201 2.9/3.0 008030K405 2.7/2.9 l
l l
., " * ~
l l l QUESTION 044 (1.0) Operating Instruction S023-3-2.6, Shutdown Cooling System (SDCS), cautions.the operator not to allow the SDCS and interconnecting piping to_ exceed a specific temperature and pressure. What is the maximum pressure and temperature that the SDCS shall not exceed?
, a. 375 deg F and 375 psia
- b. 350 deg F and 400 psia
- c. 375 deg F and 400 psia
- d. 350 deg F and 375 psia ANSWER: 044 (1.0) g3ps B d.
REFERENCE:
- 1. SO23-3-2.6 p6 , 4.1.1 KSA 005000K401 3.0/3.2 i
1 I l QUESTION 045 (1.0) Unit 2 is in Mode 2 (startup). One of the Logarithmic Power Level ' l channels has failed HIGH.. What are the Technical Specification 3.3.1 required actions for a s failed Logarithmic Powett channel in Mode 2?
- a. Restore the channel to operable condition within one hour or be in Mode 3 within 2 hours,
- b. Place the channel in the bypass condition within two-hours or be in Mode 3 within 6 hours.
- c. Restore the channel to operable condition prior to entry to Mode 1.
- d. Place the channel in the bypass condition within one hour, then continue to Mode 1. '
ANSWER 045 (1.0) epe2 B d.
REFERENCE:
- 1. SONGS 2/3 Technical Specification 3.3.1 '
KSA 000032K301 3.2/3.6 000032G003 2.6/3.3 < I I (
1 QUESTION 046 (1.0) Emergency Operating Instruction SO23-12-1, Standard Post Trip Actions, requires that Reactor Coolant System (RCS) average temperature, Tave, degrees F. " be checked to be between 545 degrees F and 555
\
Which one (1) of the following correctly describes the basis for the requirement that RCS Tave is within these limits?
- a. To ensure that main steam isolation signal does not occur.
- b. To ensure that the steam generators are removing.RCS heat
- c. To prevent uncontrolled filling of the steam generators.
- d. To prevent an imbalance in steam generator cooldown.
ANSWER: 046 (1.0) epe 2 B ' b.
REFERENCE:
- 1. SO23-12-1 p 7, 8 !
CEN-152 p 2-16 KSA 000007K106 3.7/4.1 i i { P I
l l QUESTION 047 (1.0) Unit 3 is in Mode 5 with the RCS partially drained for maintenance. One LPSI pump is OOS, the operating LPSI pump trips. The core exit temperature is 124 degrees F. The reactor was shutdown 40 days ago from 100% full power. (See attached SO23-13-15 attachment 3.) Which one (1) of the following correctly identifies the predicted time that should elapse before boiling occurs in the core?
- a. 35 minutes.
- b. 38 minutes,
- c. 40 minutes.
- d. 42 minutes.
ANSWER 047 (1.0) epe2 B c.
REFERENCE:
- 1. SO23-13-15 Attachment 4 KSA 000025G002 3.9/3.9
l NUCLEAR GENERATION SITE ABNORMAL OPERATING INSTRUCTION S023 13o15 UNITS 2 AND 3 REVISION g !
- PAGE 35 0F 39 ;
ATTACHMENT 6 TCN .2 -1 i i ( . REACTOR CORE HEATUP TIMR AFTER A LOSS OF CORE COOL NG l NOTE: This tabis'is based on RCS level at the Center of the Hot Leg after a trip with an infinite operating history at 100% power. (Ref. A01 080) 1. Record RX Core Exit Temnerature at Time Zero (loss of SDC flow): 'F, I l 2. Record RX Core Heatun Rate for Time After Trip (table below): 'F/ min.
- 3. t Calculate Time to reach 200'F/ ALERT: (If in Mode 4, then 350*F/ SITE AREA :
EMERGENCY.) ! (200'F -
) + =
(or 350'F) Temperature. Reactor Core Time to 200*F at Time Zero Heat Up Rate (or350'F] !
- 4. Calculate Time to reach 212'F (Boiling):
, (Mark N/A if in Mode 4.) # f (212*F -
) + =
Temperature Reactor Core Time to 212'F ' l
- l at Time Zero 1 Up Rate
( 5. Notify the Boiling Shift Superintendent /STA of ALERT (or SITE AREA EMERGENCY) and times. i TIME AFTER DECAY HEAT REACTOR CORE TIME FROM TIME FROM TRIP LOAD HEATUP RATE 120'F TO 145'F TO
! (days) _~ (MWth) l ('F/ min) HQll (min). BOIL (min) :
0 209 72.0 1.3 0.9 l 1 17 5.8 16.0 11.5 2 14 . 5.0 . . 19.0 13.4 3 13.0 4.6 20 14.5 i 4 12.0 4.3 22 16.0 5 11.5 4.0 23 16.7 10 - 9.7 3.3 28 20 15 8.5 2.9 32 23 20 7.6 2.6 35 26 30 6.9 2.4 38 28 4 40 6.3 2.2 43 31 i 50 5.7 1.0 1 47 34 80 4.9 1.7 54 39 100 4.2 1.5 63 45 200 3.2 1.1 83 61
- 300 2.9 1.0 93 67
{ FILE DISPOSITJON: File per 50123 0 25. T15 2.wp5 ATTACHMENT 6 PAGE 1 0F 1 1 4
j l l QUESTION 048 (1.0) 1
)
The control room 2, Shutdown fromhas been'~ev&buated Outside of the controlinRoom. accordance with S023 j The following plant conditions exist: ! Pressurizer Pressure 360 psia # Pressuriz er LeveT~~ ~"*~~~~ ~ LI-0103,indi, cates 25 % ,, _ Which one (1) of the following is the correct Pressurizer level at the Essential Plant Parameter Monitoring panel? See attached sheet SO23-13-2 Attachment 4.- -
- a. 30 % ,
- b. 35 %
- c. 38 % -- -
- d. 43 %
ANSWER 048 (1.0) epel B __ .. [.b REFERENCE- . . . -
- 1. SO23-13-2 p 30
- KSA 000068A128 3.8/4.0 t
i 000068K318 4.2/4.5 b
+
w w =wy- + w w -ww-
____ , . . - - - - - - - - - ~ ' - ~ ~ ~ ^ ^ ^ ~ ~ ~ ~ ~ ~ ' ~ ~ ' ~ ~ ~ ~ ~ ~ ~ NUCLEAR GENERATION SITE i l-t UNITS 2 AND 3 ABNORMAL OPERATING INSTRUCTION S023 13 REVIS10H 2 I PAGE 30 of 206 i ATTACHMENT 4'2'7 TCH \ l ((- LI 0103 COMPENSATION CHART 4
. . . l l
\
2250 l100 360 ; 100- -. Psla psia p ela AMBIENT l / / /
---.--. / / / / ..
l ! / _
/ / /
\ e0 -
/ / / / i 1 -
/ J ;
,j 7 ,/ .,/
/ //
/
60_ ! // / . d / // / - ' ( C / // / ' l ) {( j
$ 40 [// /
////
i \
/[/
///
i f/ . 20-i f
#/
M/ i o /// l 0 2'O 40 60 80 100 [ ] . INDICATED LEVEL - % i L l( i T2 2.wp4 ATTACHMENT 4 PAGE.6 0F 6
QUESTION 049 (1.0) ' Steam Generator Tube Rupture EOI S023-12-4 cautions the operators i to maintain Post Accident Pressurizer Pressure and Reactor Coolant System Temperature within cooldown limits. Which one (1) of the following correctly describes why maintaining Post Accident Pressure and Temperature Limits at the lower end of the band will help mitigate the consequences of a steam generator tube rupture? . . - - . . I
- a. Reduces-the amount-of. feed water inventory required for the af fected .st.ea..m. generator. . ...
- b. Reduces-the amount. of. feed water inventory. required for
- the,u,n,affected steam generator, , , ,
- c. Reduces-the.. amount of leakage from the Reactor Coolan';
System to the steam generator.
- d. Reduces the.. amount.of leakage.from.the steam generator to the Reactor Coolant System.
ANSWER: 049 (1.0) epe2 B c.
'-~
REFERENCE:
*- ~
- 1. SO23-12-4 3... 5 KSA 000037K3 06.. ,3.. 6/4.1 eGeh .g . ge g.
1 1 l i l 1 i l
QUESTION: 050 (1.0) The plant is operating at power. One (1) Circulating Water (CW) Pump.has just tripped. due to an electrical fault; three (3) CW Pumps are stil1 running. Which of the following is MAXIMUM power limit for three (3) pump operation?
- a. .95%
i b. 86%.
- c. 75%
- d. 66%
1 ANSWER: 051 (1.00) g2ps B C.
REFERENCE:
SO23-2-5, 4.3.1 (page 4 of 68) KA: 075000A202 2.5/2.7 (Site specific importance.) 075000K304 1.9/2.1 (Site specific importance.) 4 I f
.m , ._,
t QUESTION: 051 (1.00) The plant is operating at full power and one (1) Heater Drain Pump has tripped due to an electrical fault. Which of the f611owing components may be damaged if operations are continued in this condition (for more than an hour)?
- a. Main Turbine blades
- I
- b. Main Condenser tubes
- c. Heater Drain Tank level control valves
- d. Main Feedwater Pump impeller ANSWER: 051 (1.00) g2ps B b.
REFERENCE:
SO23-2-3, 4.3 (page 3 of 31) KA: 039000K106 3.1/3.0 i l I l l l l
i QUESTION: 052 (1.00) ' The plant was operating at power when a Main Steam Isolation Signal (MSIS) and Containment Isolation Actuation Signal (CIAS) were actuated. Which of the following is the MINIMUM action that must be taken to allow the MSIVs to be opened from the Control Room?
- a. Both the CIAS and the MSIS must be reset
- b. only the MSIS must be reset
- c. Only the CIAS must be reset -
4
- d. The MSIVs can be opened using OVERRIDE ANSWER: 052 (1.00) g2ps B a.
REFERENCE:
SD023-160 page 20 of 78 KA: 039000K405 3.7/3.7 i l l
QUESTION: 053 (1.0) The plant is shutdown and the Unit Auxiliary Transformers (UATr.) have been lined up and are supplying off-site power. Which of the following component alivn tents must be utilized?
- a. Main Generator output breakers must be closed b.
Main Generator disconnects must be installed
- c. Reset'Je Auxiliary Transformer must be cleared
- d. Main Transformer must be cleared ANSWER: 053 (1.00) g2ps B a.
REFERENCE:
SO23-6-5 Attachment 5, page 1 of 7 KA: 062000K104 3.7/4.2 5
){
I QUESTION: 054 (1.0) A large LOCA operating pecurred about 20 minutes ago while the plant was at power. Which one of the following actions must be MANUALLY performed after a valid Recirculation Actuation Signal (RAS) has been received? a. The Low Pressure Safety Injection Pumps must be tripped
- b. The High Pressure Safety Injection Pumps must be tripped ;
c. The Containment Emergency Sump isolation valves must be opened i d. Theshut be Refueling Water Storage Tank isolation valves must l ANSWER: 054 (1.00) g2ps B d.
REFERENCE:
SD023-740, page 8 of 74 KA: 006020A304 4.2/4.3 000011A111 4.2/4.2 006000K409 3.8/4.1 006020K403 3.2/3.6 006020K402 3.3/3.6 e
QUESTION: 055 (1.0) Plant startup and heatup is in progress.
- Which of the following Shutdown Cooling System (SDCS) conditions are a prerequisite to alignment of Containment Spray for automatic initiation?
- a. SDCS must be in operation i
- b. SDCS must be operable and capable of maintaining the RCS heatup rate within administrative limits c.
SDCS must be operable and capable of maintaining minimum RCS recirculation flow d. SDCS must' be removed from service ANSWER: 055 (1.00) g2ps B d.
REFERENCE:
SO23-3-2.9 6.1, page 5 of 42 i KA: 026000K101 4.2/4.2
QUESTION: 056 (1,0) ve&L be deD. Which one.of the following Safety Injection Tank (SIT) valves receives an automatic CLOSE signal from the Safety Injection Actuation;.S.ystem (SIAS).? a .'
. SIT Fill Valves (HV-9341, 9351, 9361 and 9371) b.
SIT Fill / Drain Valves (HV-9342, 9352, 9362, and 9372) c.
. SIT Vent Valves (HV-9345, 9355, 9365 and 9375) d.
SIT Nitrogen Valve (HV-9344, 9354, 9364 and 9374) ANSWER: 05'6 (l'. 0 0 ) g2ps B a.
REFERENCE:
SD-SO23-740, pages 27 - 29 KA: 006000K412 3.2/3.5 NRs
l QUESTION: 057 (1.0) Which of the following conditions would meet the MINIMUM requirements for operability as described in Technical ) Specification.3.8.1.1 for the fuel oil system for one (1) i i Emergency Diesel Generator for Mode 1?
- a. Day. Tank level 325 !
oil transfer pump. gallons, and one (1) operable fuel !
- b. Day Tank level 325 gallons, and two (2) operable fuel oil transfer pumps. '
c. Day Tank level 275 gallons, and one (1) operable fuel ' oil transfer pump.
- d. Day Tank level 325 gallons, and two (2) operable fuel '
oil transfer pump. ANSWER: 057 (1.00) g2ps S a.
REFERENCE:
T.S. 3.8.1.1, SD-SO23-750, pages 65 - 70 i KA: 064000K608 3.2/3.3
QUESTION: 058 (1.0) l I Electrical power has been lost to 480 VAC busses B04 and B06. How many Pressurizer heater banks have -IDST POWER?
- a. 2 I
- b. 3 1
- c. 4
~d. 6 ANSWER: (1.60) g2;.c B ;
l a.
REFERENCE:
SO23 DWGu 30118, 30120 i KA: 010000K201 3.0/3.4 i 4 l > l h l-t- i !
QUESTION: 059 (1.0) Prior to using.a HPSI pump =to' fill a Safety-Injection Tank (SIT), SO23-3-2.7.1 " SIT Operation," requires that RCS pressure be greater than 1600 psig. l Which of one the following is the-basis for this requirement?
- a. This ensures adequate HPSI head to fill the SIT
- b. This prevents HPSI injection into the RCS i i
l
- c. This prevents contamination of HPSI and SIT fill lines
- d. This prevents excessive stress on the SIT
-t ANSWER: (1.0) g2ps S '
b.
REFERENCE:
SO23-3-2.7.1, page 8 KA: 006030K603 3.3/3.6 i 1 i l' l i,
% - - , .e. . ,, _ .y v. - -,
1 QUESTION: 060 (1.0) Which of the following thermocouple-(TC) conditions are used by QSPDS to determine that Reactor Vessel level is BELOW a heated-unheated thermocouple pair? l l L
- a. The heated Tc indicates greater than 700 'F l
- b. The unheated TC-indicates greater than 500 'F
- c. The heated - unheated Tc temperature difference is-i greater than 90 'F
- d. The Jieated -_ unheated Tc temperature difference is greater than'200 'F ANSWER: 060 -(1.00) g2ps B d.
REFERENCE:
SD-SO23-820, page 52 KA: 002000K107 3.5/3.7 9 i I s* i t
s I QUESTION: 061 (1.0) Unit 2 is conducting a plant cooldown, RCS temperature is 180 'F.- ) Which of the following is the MAXIMUM cooldown rate allowed by l SO23-5-1.5 " Plant Shutdown from Hot Standby to Cold Shutdown"? !
- a. 100 degrees F / hour
- b. 75 degrees F / hour 3
.c. 25 degrees F / hour 8 degrees F / hour ANSWER: 061 (1.0) g2ps B b.
REFERENCE:
SO23-5-1.5, page 4 - KA: 002000G010 3.4/3.9 1 4 i k 1
)
-QUESTION: 062 (1.0)
The following conditions are present: J Channel "A" Low DNBR trip is bypassed. RCP-P001 (loop 1A) flow input to CPC Channel "B" has failed LOW Which of the following conditions will result in a reactor trip?.
- a. Pressurizer. pressure input ~to-Channel "C" CPC-fails low
- b. Pressurizer pressure input to channel "B" CPC fails: >
l high
- c. Loop l' Hot Leg temperature input to channel."A" CPC:
fails low ,
- d. Loop 1 Hot Leg temperature input to channel "B" CPC fails high i
ANSWER: 062 (1.0) g2ps B 4
- a. ,
REFERENCE:
SD-S023-710, page 59 - 61 KA: 012000K501 3.3/3.8 012000K607 2.9/3.2 9 i l l L i l
' QUESTION: 063- (1.0)
Which of:the-following is a Technical Specification basis for requiring an RCS temperature limit of'520 'F during power s , operations?. ,
- a. Ensures the Moderator Temperature: Coefficient is LESS
, than zero _.
- b. Ensures. Reactor Pressure Vessel temperature is LESS
. than Reference Transition' Temperature
- c.- Ensures protective instrumentation is within'its riormal i operating range '
d. Ensures thermal. stress from cooldown will remain within the analyzed limits of the safety analysis ANSWER: 063 (1.0) g2ps S C.
REFERENCE:
T.S. B 3/4.1.1.4, page B 3/4 1-2 i KA:-002000G006 2.6/3.8
?
, i. i' . i i l' i 1 l .'
l- i i QUESTION: 064 (1. 00) ! i ' Technical Specification 3.7.1.5 requires the Main Steam Isolation : Valves (MSIVs) to be operable during power operations. The basis for-this requirement is to mitigate the consequences of a main steam line (MSL) rupture. l Which one of the following concerns during a main steam line ' rupture is a Technical Specification Basis for requiring MSIV operability?- - -
- a. Minimizing the RCS inventory loss
- b. Minimizing the Steam Generator' water level transient
- c. Minimizing the Radioactivity. release rate
- d. Minimizing the containment pressure transient -
ANSWER: 064 (1.0) g2ps S d.
REFERENCE:
T.S. Bases 3/4.7.1.5, page B 3/4 7-3 KA: 039000G006 2.2/3.1 I l 1 l l I _. . __--
QUESTION: 065 (1.0) Which.of the following signals from the Turbine Protection System s is used by the Reactor Protection System to' initiate-a reactor trip on Loss of Load?
- a. Hydraulic oil pressure from the Main Turbine Unitized Actuators
- b. Steam pressure 'from the= Main Turbine first stage steam *
, chest-
- c. Valve position indication of.the Main Turbine Stop l
Valves l l d. Breakisr position of the Main Generator 22 KV output { breakers l ANSWER: 065 (1.0) g3ps B I
- a.
l l
REFERENCE:
SD-SO23-710, " Loss of Load Trip" KA: 045000K411 3.6/3.9 l-1 I I l 1
\
QUESTION: 066 (1.00) l l l L Immediately after a Reactor and Main Turbine trip at Unit 2, an ., operator inadvertently opened the Main Generator output breakers o ( BEFORE the automatic breaker. trip signal occurred. Which of the following describesLthe effects ~of'this action on the 6.9 KV and.NON-1E 4'KV busses?
- a. They will be'deenergized until manual action is taken to re-energize them
- b. 6.9'KV busses will be deenergized, NON-1E-4'KV busses will be powered from the Emergency Diesel Generators c.,
They will be energized from the Auxiliary Transformer
- d. They will be energized from the Reserve Auxiliary Transformer ANSWER: 066 (1.0) g3ps B d.
REFEREh6 : KA: 045000K301 2.8/3.1 h l l I
QUESTION: 067 (1.0) The-Standard Post. Trip Actions, SO23-12-1 (Step 5) requires the. operator .to ~ '! Verify RCS inventory control . - satisfied. " Which one_(1? of the following Pressurizer level-indications satisfies th:.s criteria for, Reactor Coolant System (RCS) Inventory control, after a reactor trip event?
- a. 75 %,:-slowlr decreasing trend.
- b. 65 4, slowly increasing trend.
- c. 25 %, slowly decreasing trend.
- d. 15 %, slowly increasing trend. !
ANSWER: 067 (1. 0) . epel B d.
REFERENCE:
SO23-12-1 p. 5 000011A103 4 i 9
, , - , , - -. . , , , - , , -..n... . ,. ,
l l l l QUESTION: 068 (1.0) Standard; Post Trip Actions,.SO23-12-1, (Step 5) requires the operatorLto " verify RCS Heat Removal-criteria - satisfied." Which one (1) of the following conditions' satisfy the criteria-for. adequate. RCS.. Heat Removal: Criteria? )
.: = ~ ' : .
- a. Steam generator levels 30 %.
- b. One main Feed water pump operating Reactor-Trip Override NOT actuated,
- c. Auxiliary Feed Water flow 300 gpm.
- d. Core Exit Saturation Temperature 15 degrees F.
ANSWER:.068 (1.0) epel B l c.
REFERENCE:
SO23-12-1 p. 7
~
KSA 000074K103 4.5/4.9 000074K311 4.0/4.4 l L 1 1 l l l
_ QUESTION: 069 (l'. 0) _ Standard Post Trip Actions, SO23-12-1, (Step di _ requires the-
. operator to " verify Vital Auxiliaries functioning properly."
What is the correct electrical configuration of the 6.9 kV and non-1E 4-kV buses, under post trip conditions?
- a. 6.9 kV and non-1E 4 kV' busses are energized from the Unit Auxiliary transformer.
'b. 6.9 kV busses are deenergized; non-1E 4 kV buses are energized from the Auxiliary transformer.
- c. 6.9 kV and non 1E 4KV busses are energized from the Reserve Auxiliary Transformer.
- d. 6.9 kV buses are deenergized; non IE 4KV busses are energized from Reserve Auxiliary Transformer.
ANSWER: 069 (1.0) epel B c.
REFERENCE:
SO23-12-1 p. 4 000007G010' 4.2/4.1 t
^
l l.
- - _ , .-.. ~ -- .
QUESTION: 070 (1.0) Unit 2 was operating at 100 % power when a reactor 1: rip occurred. Nhich one (1) of the following plant conditions will require you to close the Main Steam Isolation Valves?
.a. Condenser Vacuum decreasing to 7" Hg. .
- b. Only one Circulating water pump operating.
- c. Turbine speed 2000 rpm and not decreasing.
- d. Both main feedwater pumps trip.
ANSWER: 070 epel B c.
REFERENCE:
SO23-12-1 p. 4 000074A208
~
l t 1 i l
t QUESTION: .071 (1.0) Abnormal Operating-Instruction, SO23-13-12, "Losslof Boron Concentration Control / Inadvertent Dilution," cautions that a reduced letdown temperature will.cause.a-dilution effect in the .] RCS. - Why will a. reduced letdown temperature causeEan inadvertent 4 dilution? l a.- The Purification Ian Exchanger's-boron saturation will ' NOT be in equilibrium.
- b. The Purification Ior.fExchanger: filterimedia will l release CATION-material which will block the letdown strainer, F-100.
t
- c. The Purification l Ion Exchanger's boron concentration-will be super-saturated:with boron.-
- d. The Purification Ion Exchanger filter media will 2 release ANION material which will block the letdown strainer, F-100. ,
1 1 4 ANSWER 071- (1.0) epel B i 2 a. i
REFERENCE:
o SO23-13-12, p. 2; 7 SD-SO23-390 KA 0000241:104 2.8/3.6 ' f
- l 1
1 t R i~
. , - . -. , ,.,,,,e, . - _y . _ . . - , - , . ,
QUESTION: 072 (1.0) The unit is-in-Mode 2 and you are emergency borating in accordance with: SO2 3-13-11,, Emergency Boration of Reactor Coolant System. Both boric acid. makeup pumps trip, and fail, and can-not l be restarted. Both the RWST and BAMU tanks are above their- I minimum TS levels.
~
Which one (1) of the following correctly describes the preferred gravity feed method of completing the emergency boration, in accordance with SO23-13-117
- a. From the Refueling Water Storage Tank (RWST) to the i suction of the charging pumps.
- b. From the Boric Acid Makeup Tank (BAMU tank) to the suction of the charging pumps .
- c. From the RWST to the LPSI pumps then-to the suction of the charging pumps.
d.- From the Boric Acid Makeup Tank to the HPSI pumps then to the suction of the charging pumps. ANSWER 072 (1.0) epe 1 B b.
REFERENCE:
SO23-13-12, p. 4, i SD-SO23-390 KA 000024A201 3.8/4.1 u o 4 6 4
--- ,, .-,-,.v- - - .
t QUESTION: 073 (1.0) Your.. shift has just experienced'an uncomplicated reactor trip due !
.tcr a Core Protection. Calculator- (CPC)' malfunction. Verification i of " Safety Function Criteria.- Satisfied," as specified by the l
Reactor Trip Recovery Procedure, SO23-12-2, has been initiated. What is the MAXIMUM time period, permitted by the procedure, between readings to verify. Safety Function Criteria Status?
*a.- 5 minutes
- b. 10 minutes
- c. 15 minutes
- d. 30 minutes ANSWER 073 (1.0) epe2 S c.
REFERENCE:
SO23-12-2; LP 2EO719 p. 6 1 000007G012 3.8/3.9
~
r _,.e- ,
QUESTION: 074 (1.0) You are the Unit 2 CRS. 'Your shift has just experienced an uncomplicated reactor. trip; due to a Main Turbine load drop. Your control Operators are+1n the process of reactor trip recovery. One of the operators reports that he is lowering the steam Generator feed rates to maintain Core Exit Saturation Margin l below'160. degrees.F. What is the reason for this action?
- a. To prevent overcooling the Reactor Coolant System and minimize Pressurized Thermal Shock.
- b. To prevent overfilling the Steam Generators and minimize Feed Ring thermal Shock.
- c. To prevent overcooling the Reactor Coolant System and minimize pressurizer Spray line Thermal Shock.
- d. To prevent overfilling steam generators and minimize feed nozzle Thermal Shock.
ANSWER 074 (1.0) epe2 S a.
REFERENCE:
SO23-12-2;. LP 2EO719 p. 6 000007K318 4.2/4.5 l I l l l l l 1
QUESTION: 075 (1.0)- You are the Unit 2 CRS. Your shift experienced an uncomplicated reactor 2. trip, and is recovering per recovery procedure SO23 one' of your. Control Operators reports that he cannot verify that both Steam Generator (S/G) levels are below 100% wide range 1evel indication, and subsequently actuates an MSIS, Why was the Control Operator procedurally required to actuate an MSIS?.'
- a. To prevent S/G carryover and subsequent main steam line damage. -
- b. To prevent excessive-RCS cooldown and subsequent.RPV thermal shock, c.
To prevent steam generator Main Steam Safety Valves from actuating,
- d. To prevent excessive RCS cooldown and excessive Pressurizer out surge.
ANSWER 075 (1.0) epe2 S a.
REFERENCE:
'l SO23-12-2; LP 2EO719 p. 12 KA 000007K302 4.2/4.4 l
l l
,. _ -~
- l L
QUESTION: 076 (1.0) As1the Shift Superintendent,.you and the CRS determine that Control Room Evacuation is.necessary. One of the four immediate actions ~.that are performed before evacuating is to: " Select manual and stop all charging pumps." What is the purpose for performing this immediate action?~
- a. Isolates the primary from CVCS - due to the loss of operator control over CVCS.
- b. . Prevents damaging the charging pumps, if VCT outlet closed.
- c. Isolates the primary from CVCS - due'to loss of operator control over PZR level.
- d. Prevents inadvertent RCS depressurization - if pressurizer spray valves inadvertently open.
ANSWER 076 (1.0) epel S b. l
REFERENCE:
i SO23-13-2[ LP 2A0702 p. 4 ? t j 000068K318 4.2/4.5 l l t l i -i l i i l l 1 l
l. QUESTION: 077 (1.0) l As the Shift Superintendent, you and the CRS' determine that' Control. Room Evacuation is necessary. The last of the four i immediate actions _that are performed before evacuating is to:
"Stop all RCP's".
What is the purpose for performing this immediate operator action?
- a. Prevents excessive Reactor System Cooldown if a Main Steam safety valve fails open.
- b. To prevent excessive Reactor System Cooldown in the event of overfeeding a steam generator, c.
To prevent excessive Reactor Coolant System inventory loss upon pressurizer level control system failure.
- d. To prevent excessive Reactor. Coolant System depressurization upon pressurizer prescure control system failure.
ANSWER 077 epel S d. REFERENCEi' SO23-13-2; LP 2A0702 p. 4 000068G005 3.6/3.8 l l i l l l l l l
QUESTION: 078 (1.0) . 1 An inadvertent Safety Injection Actuation Signal (SIAS) occurred while the Unit was at 100 % power. SO23-13-17, Recovery From Inadvertent Safety -Injection / Containment Isolation, has been i i entered. The procedure requires you to override and place the charging pumps in stop,. placing'the Unit in Technical Specification (TS) Action Statement 3.0.3. What one (1) of the following' actions is required within one (1) hour in order to exit TS 3.0.3?-
- a. Ensure that letdown is reestablished,
- b. Ensure that Backup Heaters are placed in Override,
- c. Reset the Reactor Trip Override.
- d. Reset the Safety Injection Actuation Signal.
- ANSWER 078 (1.0) epe2 S d.
REFERENCE:
- 1. SO23-13-17 p 3 KA 006050A201 3.9/4.2 l -
L l l l
l QUESTION: 079 (1.0) Unit 2 is in MODE 1 when the " Air Compressor Control Trouble Alarm" initiates. The Nitrogen Backup System begins to supply-the Instrument Air , headers: , hhat-istheInstrumentAirheaderpressureatwhichthenitrogen backup system will begin to supply instrument air header? l
- a. 85 - 90 psig l
l b. 75 - 80 psig
- c. 65 - 70 psig
-d.- 55 - 60 psig.
i ANSWER 079 (1.0) epe2 B c.
REFERENCE:
SO23-13-5; LP: 2A0705 pp. 3&4 000065K304 3.0/3.2 ' 000065K308 3.7/3.9 i i
l QUESTION: 080 (1.0) i Unit 2 is in. MODE 1 when.the " Air Compressor Control Trouble Alarm": initiates._ The Nitrogen Backup' Air system is maintaining the loads on the Instrument Air headers. Per SO23-13-5, " Loss of I Instrument Air":. When would you expect to commence an " orderly plant shutdown to Hot Standby?" l , a. When normal system supply pressure cannot be restored. I 1
- b. Immediately _ upon trouble alarm confirmation.
1
- c. When one air operated valve starts to drift from its i required position.
d. One hour after the trouble alarm was received. ANSWER 080 (1.0) epe2 B
- a. a l
REFERENCE:
SO23-13-5; LP: 2A0705 pp. 3-5 000065A205 000065K3 08- 3.7/3.9 l l -
QUESTION: 081 (1.0) A spurious safety Injection Actuation signal (SIAS) and Containment Purge Isolation Signal.(CPIS) has occurred. Operators in the Fuel-Handling-building notice pool temperatures rising. ; Why would Spent Fuel Pool temperatures rise.after this event?
- a. A SIAS'would isolate the Component Cooling Water Non-critical loo j Exchangers, p supplying the Spent Fuel Pool Heat i b.
A SIAS would isolate the Nuclear Service Water Non-critical loop supplying the Spent Fuel Pool Heat i i Exchangers. c. A CPIS would isolate the Nuclear Service Water cooling the Transfer Canal, subsequently heating the Spent Fuel Pool. d. A CPIS would isolate the Component Cooling Water Non-critical loop heating coolingFuel the Spent the Pool. Transfer _ canal ~, subsequently i ANSWER 081 (1.0) epe3 B a.
REFERENCE:
SO23-13-20; LP: 2A0720; SD-SO23-430, p.94 ( KA 000036A104 3.1/3.7 f i i: y .
i QUESTION: 082 (1.0)
.During refueling evolutions in MODE 6, Vessel Pool Upper / Lower. Seal Low ' Pressure Alarm. "you teceive a " Reactor What is the NORMAL Reactor Vessel Pool Seal supply source?
a.- Instrument Air _ q
- b. Main Nitrogen' Bank. i
- c. Nitrogen cylinders t
- d. Site Service Air supply i
ANSWER 082 (1.0) epe3 B a. REFERENCE * ! I SO23-13-20; LP: 2A0720; SD-SO23-430, p.94 KA 000036A104 3.1/3.7 1
QUESTION 08) (1.0) Which one (1) of the following is the BASIS for the Technical Specification limit.for a large misalignment (greater than or equal to 19 inches) of two or more Control Element Assemblies (CEAs)?
- a. Ejected CEA worth VMed in the Safety A :alysis may not '
be. conservative,
- b. Severity of an anticipated transient without a scram is significantly increased.
- c. Reactor protection system does not account for distortion of the core power distributio...
- d. Increases the severity of a loss of feed water incident. '
ANSWER: 083 (1.00) epel S L. *
REFERENCE:
1.T.S. BASES 3/4.1.3 Amend. No. 61 Page 3/4 1-3 KSA 000005K304 3.4/4.1 l l l i N
.n,
QUESTION 084 (1.0) A of Loss of Power. Offsite Forced You Circulation has occurred are attempting as a result to establish of a Loss natural circulation in'accordance with Emergency Operating Instruction SO23-12-72, Loss of Forced Circulation. Which one (1) of the following conditions will require a
" Response Not obtained" action when performing verification of ,
Natural Circulation? l
-a.
Loop delta T at 30 degrees F, and increasing slowly.
- b. Core Exit Saturation Margin at 10 degrees F, and increasing slowly.
- c. T hot / Representative Core Exit Thermocouples delta T at 5 degrees F, and increasing slowly. f
- d. Reactor Vessel pleaum level above 82%, and constant.
ANSWER: 084 (1.00) epe 1 B b.
REFERENCE:
- 1. EOI S023-12-7, Page 9, para. 10.
KA 000015K101 4.4/4.6 I i J
i QUESTION 085 (1.0) Which one (1.) of the following is an entry condition for the Emergency Boration.of the RCS. Abnormal Operating Instruction, SO23-13-11? ' - -
- a. Failure.of one (1) CEA to drop following a reactor trip.
- b. Tave 15 degrees below Tref due to a main feedwater system transient. [
- c. Shutdown. margin of 4.15 % delta K/K during Mode 5 Operation.
- d. Group 6 regulating CEAs below Power Dependent Int,ertion Limit.
ANSWER: 085 (1.00) epel B d. P REFERENCE '
- 1. LICENSE TEST NUMBER 300.
- 2. LESSON PLAN # 2A0711 OBJECTIVE 1.0
- 3. S023-13-1110-2 ,
l KSA 000024K301 4.1/4.4 i
- y. .. ._ _ , . , -.w... . - --
QUESTION 086 (1.0) Which.one (1).of.ttie.following conditions would requiro a immediate S023-13-7, Iristruction reactor and turbine trip per the Abnormal Operating or Salt Water Cooling"? entitled " Loss of Component Cooling Water
- a. LoEE'of CCW to the CEAs.
.b. One stage of a Reactor Coolant Pump (RCP) seal hat failed on three (3) RCPs. j
- c. A RCP thrust bearing twaperature at 225 degrees F.
- d. One train of CCW fails on loss of CCW pump.
.1 ANSWER: 086 (1.00) epal B
- c. !
REFERENCE:
1..SO23-30-7 Page 4 KSA ,0026J010 4.0/4.2 I KSA s00026K303 4.0/4.2 i y
.. d
-1 I
QUESTION OS7 (1.0) SONGS 2/3 EOI S023-12-5, " Excess Steam Demand Event", allows operators to terminate or throttle High Pressure Safety Injection (HPSI) when specific conditions are met. Which one (1) of the following conditions meets the criteria to terminate or throttle HPSI?
- a. Praceuriter level 35 %. 6
- b. Core Exit Savuration margin greater than 15 degrees F.
- c. Pressurizer pressure greater than 400 psia.
~
i
- d. Reactor Vessel plenum level ^1 %.
ANSWER: 087 (1.00) epel B a. i
REFERENCE:
- 1. EOI S023-12-5, PAGE 39 i
KSA 000040A205 4.1/4.5 l i i e t t l L i h l
. ~ , . _ __ _ _ _ _ _
QUESTION 088 (1. 0) EOI S023-12-5, " Excess Steam Demand Event" provides criteria for determining that an excess steam demand is isolated. Which one (1) of the following is a valid indication that the source of excess steam demand is isolated?
- a. Pressurizer pressure is decreasing.
b. Steam Generator indicated steam flow decreases to zero.
- c. RCS Tc in each loop increasing.
- d. Both steam generator levels increasing.
ANSWER: 088 (1.00) epel B c.
REFERENCE:
- 1. EOI S023-12-5, PAGE 5, AND 15.
KSA 000040G011 4.1/4.3 l
~
l l i
QUESTION 089 (1.0) , Abnormal Operating Instruction S023-13-10, " Loss of Condenser , Vacuum" provides operator guidance for use of HS-2808B, Heat Treat override. Switch. Which one (1.0) of the following is the MAXIMUM time that HS-2808-B may be placed in OVERRIDE during a loss of condenser vacuum? ;
- a. 5 minutes.
k
- b. 15 minutes.
- c. 30 miinutes,
- d. 60 minutes.
r ANSWER: 089 (1.00) epel B
- c. '
REFERENCE:
- 1. AOI S023-13-10, PAGE 2 KSA 000051A202 3.9/4.1
/
e
l 1 QUESTION 090 (1.0) i h One of the major obje.ctives of SO23-12-8, Station Blackout, '
- g. to establish a source of 4 KV electrical power.
' j s
Which one (1) of the following is another major objective of SO23-12-87 1 I
- a. Maintain pressurizer level greater than 5 % as long as possible.
- b. Maintain steam' generator levels less than 30%.
- c. Maintain shutdown margin greater than 6.15 % delta K/K i for as long as possible,
- d. Maintain single phase natural circulation as long as possible.
ANSWER: 090 (1.00) epel B
- d. .
REFERENCE:
P
- 1. LESSON PLAN 2E0720 PAGE 5 KSA 00005,5G003 4.3/4.6 l
- h k
.. ._ ._ _ . _ . _ . __ ~_.
i 1 l l 1 1 1 QUESTION 091 (1.0) 1 Which one (1) of the following correctly describes why MSIS j initiation will aid in. mitigating the effects of Station Blackout? l
- a. Prevents Stea:n Generator Safety Actuation. !
- b. Prevents Damage to the High Pressure Feedwater heaters, l j
- c. Prevents damage to main condenser. *
- d. Prevents over feeding of steam generators.
ANSWER: 091 (1.00) epel B c.
REFERENCE:
- 1. LESSON PLAN 2E0720 PAGE 6 KSA 000055K302 4.3/4.6 000055K204 3.7/4.1 i
we e s l l l 1
QUESTION 092 (1.0) During Station Blackout, EOI S023-12-8, a CAUTION requires load reduction of Batteries D3 and D4 within thirty (30) minutes of the. Station Blackout in order to extend battery life. Which one (1) of the following is the reason for extending Batteries D3 and D4 load life?
- a. Ensures RO! controlled bleedoff isolation valve HV-9216 operable for up to 8 hours.
- b. Ensures ADV operable en S/G with AFW flow up to 8 hours.
- c. Ensures MSIV operable on S/G with AFW flow up to 8 hours.
- d. Ensures SDC suction valve operable from control room for up to 8 hours.
ANSWER: 092 (1.00) epel B > d.
REFERENCE:
- 1. EOI S023-12-8 PAGE 3
- 2. LESSON PLAN 2E0720 PAGE 7 KSA 000055,,K302 4.3/4.6
QUESTION 093 (1.0) _A. Class 1E.120. volt vital AC bus output breaker has tripped. Which will one (1) of the following correctly describes how the bus be reenergized? J t
- a. Manual transfer to the alternate source of power. ;
- b. Manual transfer to the standby inverter.
c. Automatic power. bus transfer to the alternate source of e
- d. Automatic transfer to the standby inverter.
l ANSWER: 093 (1.00) epe 1 B ' a.
REFERENCE:
- 1. OI S023-6-17, PAGE 3 1 KSA 000057A101 3.7/3.7 000057K301 4.1/4.4 4
5 i 9
' '~
QUESTION 094 (1.0) Operating Instruction S023-6-17, " Class 1E 120 VAC Vital Bus Power. Supply System Operation" requires that if High in-rush current loads are to be energized from the vital bus; then the Vital Bus supply should be transferred to the alternate source, and then high transferred in-rush loads haveback beentoenergized. the Vital Bus (inverter) once the I Which one (1) of the following provides the reason for this requirement? t I
- a. Avoids operation of fast acting current limiting feature of the inverter.
- b. Avoids reduction of inverter life expectancy due to transient overloading.
c. Prevents feedback into the alternate source.
- d. Prevents failure of the standby inverter.
ANSWER: 094 (1.00) epel B a.
REFERENCE:
i
- 1. OI S023-6-17, PAGE 3 l
KSA 0000E7A101 3.7/3.7 l l l l 4 J p
QUESTION 095 (1.0) 1 l Which one (1) of the following plant staff positions is responsible for coordinating the fire fighting effort for a fire i in the Unit 2 nor.-1E 4 KV electrical switch gear? ! I
- a. Shift Superintendent.
- b. Unit 2 contrei Roo.a Oupervisor.
j
- c. Unit 2 Control Operator.
- d. Unit 3 control Operator.
ANSWER: 095 (1.00) epel B
- a. '
REFERENCE:
- 1. SO23-13-21, PAGE 6
- 2. LESSON PLAN 2A0721, PAGE 5 KSA 000067K304 3.3/4.1 L
I
- QUESTION 096 (1.0) l AOI S023-13-16, " Loss.of. Containment Integrity", provides operator actions if. containment integrity is lost.
. {
Which one (1) of the following conditions would NOT require implementation of the actions in SO23-13-15 if refueling operations were in' progress?
- a. ONE door in each air lock becomes inoperable.
- b. Equipment hatch being held closed by TWO (2) bolts.
- c. Containment Purge Isolation valve becomes. inoperable,
- d. Containment Emergency Cooling Unit Containment Isolation valve becomes inoperable.
ANSWER: 096 (1.0) epel B
,gf &b - '
REFERENCE:
- 1. LESSON PLAN 2A0716, PAGE 6 KSA 000069G011 4.0/4.2 u-
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i QUESTION 097 (1.0) Two of the three reactor trips initiated by the failure of Lineer Power. instrument JI-0002A, B, C, or D, ara: ' Linear Power Level - High Iccal Power Density - LOW What is the Third Reactor Trip affected by the loss of a linear power instrument?
- a. Log Power. level - High
- b. DNBR - Low
- c. Axial Shape Index - Low
- d. Planar Radial Peaking Factor - High '
- ANSWER 097 (1.0) epe2 B b.
REFERENCE:
- 1. SO23-13-18, P 2 & Attachment 1
- 2. SONGS'2/3 Technical Specification 3.3.1 KA 000007K203 3.5/3.6 000007A105 4.0/4.1 1
I I QUESTION 098 (1.0) i The failure of a Pressurizer narrow range pressure instrument i I PT-0101- 1,.2, 3,.or.4, will affect the " Pressurizer Pressure - High" Reactor Trip. i l What are the other TWO (2) Reactor Trips which will be affected l 1 by the loss of a Narrow Range Pressurizer Pressure Instrument? I
- a. Pressurizer Pressure - Low (CCAS), and 1 j
Pressurizer Pressure Low RPS i
- b. Pressurizer Pressure - High (SIAS/ccAS), and Pressurizer Pressurs - Low (RPS) >
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Local Power Density - Low DNBR - High
- d. Local Power Density - High DNBR - Low
- ANSWER 098 (1.0) epe2 B d.
REFERENCE:
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- 1. S00#-13-18, P2 & Attachment 1 l 2. SONGS 2/3 Technical Specification 3.3.1 KA 000007K203 3.5/3.6 000007A105 4.0/4.1 P
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QUESTION 099 (1.0) Which of the following choices is a means of determining whether a Reactor Protectfen System Channel failure involves single or ; redundant functiC")^. u. nit channels?
- a. Check the COLSS diagnostic printout.
- b. Check the Critical Functions Monitoring System (CFMS) diagnostic printout.
- c. Verify the Core Protection Calculator (CPC) matrix Power Supply indicating lights on the apron section of the CPC panels.
- d. Verify Logic Matrix Power Supply indicating lights inside the door on back of the PPS panel. .
- ANSWER (1.0) epe2 B
- d. or C, .
REFERENCE
- 1. S023-13-18, p2 KA 000033G009 2.9/3.0 t
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QUESTION 100 (1.0) Which one (1) of the following statements correctly describes one of the indications you would see for a " SLIPPED" Control Element Assembly? ~- a. CEAC CEA CRT had would indicate a solid line as far down as the slipped.
- b. CEA Rod Bottom light lit (on).
- c. A decrease in Plknt Monitoring System CEA position 1 indication.
- d. A Plant Monitoring System CEA position indication of zero (0). ,
- ANSWER 100 (1.0) S '
c.
REFERENCE:
- 1. SD-SO23-510 KA 000003A201
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r R0 EXAMINATION QUESTION: 005(1.0) S0123-0-11. " Narrative Loge,", requires specific entries be made in the Control > Operators lo when specific t.vpas of systems or components are declared Out of , Service (005 . San Onofre's Unit 2 Shutd vn Cooling System (SDCS) has been declared 00$. Which one (1) of the following .. mbols must be entered in the margin of the Control , Operators Log? ,
- a. Red E
- b. Black E ,
- c. Red 005
- d. Black 005 ,
ANSWER: 005(1.0) pwg B F
- b. [1.0)
REFERENCE:
- 1. 50123-0-11 " Narrative Logs", TCN 1-1, P. 12 SCE COMMENTS:
S0123-0-11 is 4mbiguous on this point. a and b could be choser qs:.ed on the . reference. , SCE REQUESTED RESOLUTION: 1 Accept both answers a and b ! NRC RESOLUTION Accept licensee comment; answers "a" and "b" will be accepted as correct.- 1 i I l 1 i l
1 i R0 EXAMINATION ! QUESTION: 018 (1.0) ! A reactor startur is being conducted in accordance with S023-5-1.3.1, Plant.
- Startup From Standby To Minimum Load, and 5023-3-1.1, Reactor Startup. j Which one (1) of the following is the maximum positive startup rate that shall not be exceeded while conducting a plant startup?
- a, '0.5 DPM '
- b. 1.5 DPM >
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- c. 2.5 DPM
- d. 5.0 DPM ANSWER: 018(1.0) gips R
- a. [1.0]
REFERENCE:
- 1. S023-5-1,3.1 p7 5
- 2. S023-3-1.1 P 29 KSA 001000K518 4.2/4.3 SCE COMMENTS:
Answers a and b could be chosen as correct answers. Reference S023-3-1.1
- (page 23 of 58), step 3.1.1.8 states, "Stop CEA withdrawal if startup rete indication exceeds 1.5 DPM." With this step in the procedure and in a position where it is not easily found, it is believed both answers should be accepted.
SCE REQUESTED RESOLUTION: , Accept answers a and b. 1 NRC RESOLUTION Accept licensee comment; answer "a" and "b" will be accepted as correct.
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R0 EXAMINATION QUESTIONS: 043 (1.0) ComponentCoolingWater(CCW)hasbeenlosttoaReactorCoolantPump(RCP) ! while in Mode 1. You have entered A01 S023-13-6,~ RCP Seal' Failure, and AOI S023-13-7 Loss of CCW/ SWC. ; What is the MAXIMUM time that a Reactor Coolant Pump can be operated without : Component Cooling Water (CCW)? 1
- a. 3 minutes
- b. 5 minutes i
- c. 15 minutes
- d. 30 minutes '
i ANSWER: 043 (1.0)g3psB . d.
REFERENCE:
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- 1. S023-13-7 p3 i KSA 008030K302 SCE COMMENTS:
i Answers a and d can be chosen as correct answers. S023-3-1.7 precaution 4.9 l ! on page 5 of 41 states, the Reactor Coolant Pump should not be operated for j greater than 3 minutes without 53 GPM Component Cooling Water flow on the ) Seal Water Exchanger or seal damage may result. Since seal cooling flow is not readily measured, inadequate flow is indicated by Seal Water CCW outlet l temperature greater than 120 degrees F." SCE REQUESTED RESOLUTION: Accept both answers a and d. l NRC RESOLUTION L ! Licensee resolution not accepted. Correct answer has been corrected to "a". The only correct answer is "a".
i I RO EXAMINATION i
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QUESTION: 049(1.0) ) The control room has been evacuated in accordance with 5023-13-2, Shutdown ! from Outside of the Control Room. . The following plant conditions exist: Pressurizer Pressure 360 psia : LI-0103 indicates 25% Which one-(1) of the following is the correct pressurizer level at the i Essential Plant Parameter L..itoring panel? See attached sheet S023-13-2 Attachment 4.
- a. 30% ,
c
- b. 35%
- c. 38%
- d. 43%
F ANSWER 049(1.0)epelB b.
REFERENCE:
- 1. S023-13-2 p. 30 KSA 000068A128 3.8/4.0 000068K318 4.2/4.5 SCE COMMENTS:
1 i Thecorrectanswershouldbe(a.)30%. When reading the attached graph, l 360 PSIA and 25% level, the correct answer is 3J%. ' SCE REQUESTED RESOLUTION: Change answer key to (a.). NRC RESOLUTION Concur with licensee comment: Response corrected to "a". r
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RO EXAMINATION QUESTION: 068 (1.0) Which of the following signals from the Turbine Protection System is used by the Reactor Protection System to initiate a reactor trip on Loss of Load? a Hydraulic oil pressure from the Main Turbine Unitized Actuators. ' I
- b. Steam pressure from the Main Turbine first stage steam chest.
- c. Valves position indication of the Main Turbine Stop Valves,
- d. Breaker position of the Main Generator 22 KV output breakers.
ANSWER: .068 (1.0) g3ps B l a.
REFERENCE:
5D-S023-710. " Loss of Load Trip" KA: 0450000K411 3.6/3.9 3 SCE COMMENTS: j a and c could be chosen as correct answers. SD-5023-710 REACTOR PROTECTION ! SYSTEM page 15 of 77 states the Turbine High Pressure Valves Close position i as a setpoint for the Loss of Load Trip. i SCE REQUESTED RESOLUTION: Accept both answer a and c. l i NRC RESOLUTION i Licensee conrnent not accepted. Both answers "a" and "c." were partially incorrect, no correct answer i provided. Question deleted. l l i
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RO EXAMINATION 1 QUESTION: 070 (1.00) Imediately after a Reactor and Main Turbine trip et Unit 2, an operator inadvertently opened the Main Generator output breakers BEFORE the automatic ; breaker trip signal occurred. Which of the following describes the effects of this action on the 6.9 KV ' l and NON-1E 4 KV busses?
- a. They will be de-energized until manur,1 action is taken to ;
re-energize them,
- b. 6.9 KV busses will be de-energized, NON-1E 4 KV busses will be powered from the Emergency Diestl Generators,
- c. They will be energized from the Auxiliary transformer. ,
- d. They will be energized from the Reserve Auxiliary Transformer.
ANSWER: 070 (1.0)g3psB d.
REFERENCE:
KA: 045000K301 2.8/3.1 SCE COMMENTS: . Answer d is the correct answer; however, (a) may be seen as the correct' ' answer by some of the license candidates. The automatic transfer-logic for opening the generator output breakers is a Unit 2/3 difference. However, this change was only completed in February of 1990 and the s simulator is not modeled to the plant at this time. Therefore, based on the time frame of completion of this change and the simulator not being modeled to unit 2 at this time we believe both answers a and d should be accepted. SCE RE0 VESTED RESOLUTION: l Accept both answers (a) and (d). NRC RESOLUTION Response "d" described as-built plant conditions prior to examination material request date. Question will be retained and answer "d" accepted as only correct answer, l
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SRO EXAMINATION QUESTION: 005(1.0) 50123-0-11. " Narrative Logs", requires specific entries be made in' the Control Operators log when specific types of systems or components are declared Out of Service'(005). San Onofre's Unit 2 Shutdown Cooling System (SDCS) has been declared 005. Which one (1) of the following symbols must be entered in the margin of.the Control Operators Log?
- a. Red E
- b. Black E
- c. Red 00S
- d. Black 005 ANSWER: 005(1,0) pwg B
- b. [1.0)
REFERENCE:
- 1. 50123-0-11 " Narrative Logs", TCN 1-1, P. 12 SCE COMMENTS: I i
S0123-0-11 is ambiguous on this point. a and b could be chosen based on the reference. SCE REQUESTED RESOLUTION: Accept both answers a and b. NRC RESOLUTION Accept licensee comment; answers "a" and "b" will be accepted as correct. I i 1 l
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SRO EXAMINATION QUESTION: 042 (1.0) Component Cooling Water (CCW) has been lost to a Reactor Coolant Pump '(RCP) while in Mode 1. You have entered A01 S023-13-6, RCP Seal Failure, and AOI S023-13-7 Loss of CCW/ SWC. What is the MAXIMUM time that a Reactor Coolant Pump can be operated without ComponentCoolingWater(CCW)?
- a. 3 minutes
- b. 5 minutes
- c. 15 minutes
- d. 30 minutes ANSWER: 042 (1.0)93psB d.
REFERENCE:
- 1. S023-13-7 p3 KSA 008030K302 4.1/4.2 SCE C0KMENTS:
Answers a and d can be chosen as corrcet answers. S023-3-1.7 precaution 4.9 on page 5 of 41 states, "the Reactor Coolant Pump should not be operated for greater than 3 minutes without 53 GPM Component Cooling Water flow on the Seal Water Exchanger or seal damage may result. Since seal cooling flow is not readily measured, inadequate flow is indicated by Seal Water CCU outlet temperature greater than 120 degrees F." SCE REQUESTED RESOLilTION: Accept both answers a and d. NRC RESOLUTION Correct answer has been corrected to "a".
1 , c- SRO EXAMINATI0N QUESTIONi 048(1,0) The control rcora has been evacuated in accordance~ with S023-13-2 Shutdown , from Outside of the Control Room. -l The following plant conditions exist:'
. Pressurizer Pressure 360 psia LI-0103 indicates' 25%
Which one'(1) of the- following is' the correct pressurizer level at the k Essential Plant' Parameter Monitoring panel? See attached sheet S023-13-2 Attachment 4
- a. 30%
- b. 35% "
- c. 38%
d- 43% ANSWER: 0% (1.0) epel 0 b.
REFERENCE:
- 1. S023-13-2 p. 30 KSA 000068A128 3.8/4.0' 000068K318 4.2/4.5 SCE COMMENTS:
The correct answer should be (a'.) 30%. When reading the attached graph, 360 PSIA and 25% level, the correct answer.-is 30%. SCE REQUESTED RESOLUTION: Change answer key to (a.). NRCJESOLUTION Concur with licensae comment: Response corrected to "a".
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SRO EXAMINATION QUESTION: 065 (1.0) Which of the following sionals from-the Turbine Protection System is used~ by the Reactor Protectic. System to. initiate a reactor trip on Loss of Load?- a Hydraulic oil pressure'from the Main Turbine Unitized Actuators,
- b. Steam pEessure from the Main Turbire first stage steam chest.
~
c.- Valves position indication of the Main Turbine Stop Valves,
- d. ~ Breaker position of the Main Generator 22'KV output breakers.
ANSWER:= 065 (1.0) 93ps 8 t a.
REFERENCE:
S0-S023-710. " Loss of Load Tri'" KA: 0450000K411 3.6/3.9 SCE COMMENTS: a and c could be chosen as correct answers. SD-S023-710 REACTOR PROTECTICH SYSTEM page 15 of 77 states the Turbine High Pressure Valves Close position as a setpoint for the Loss of Load Trip. i SCE REQUESTED RESOLUTION: Accept both answer a an6 c. NRC RESOLUTION Licensee comment not accepted. Both answers "a" and "c" were partially incorrect, no correct answer provided. Question deleted. t k
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QUESTION: 066 -(1.00) Imediately after a Reactor and Main Turbine trip at Unit 2, an operator.- j inadvertently opened the Main Generator outputbreakers.BEFORE the automatic breaker trip signal occurred. { Which of the following' describes'the effects of.this action on the 6.9 KV-and-NON-1E 4 KV busses? i
- a. .They will be de-energized until manual action isa taken-to re-energize them, b',
6.9 KV busses will be de-energized, NON-1E 4 KV busses will be j powered from the Emergency Diesel Generators,
- c. They will be energized'from the Auxiliary transformer.
- d. They will be energized from the Reserve Auxiliary Transformer. '
ANSWER: 066 (1.0)g3psB l d.
REFERENCE:
KA: 045000K301 2.8/3.1 SCE COMNENTS: Answer d is the correct answer; however, (a) may be seen as the correct answer by some of the license candidates. The automatic transfer logic for opening the generator output breakers is a Unit 2.3 difference'. However, this change was only completed in February of 1.990 and the' simulator is not modeled to the plant at this time. Therefore, based on the time frame of completion of this change and the simulator'not being modeled accepted. to unit 2 at this tine we believe both answers a and d should be l SCE RE0 VESTED RESOLUTION: Acceptbothanswers(a)and(d). HRC RESOLUTION J Response "d" describes as-built plant conditions prior to examination material request date. Question will be retained and # r "d" accepted
-as only correct answers.
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' Enclosure 4 SIMULATION FACILITY REPORT Licensee: Southern California Edison Company-San Onofre Nuclear Generating . Station:
Units 2 and-3 Docket-No.: 50361 and 50362 i Operating Tests Administered on:- June 18 ~through June 28, 1990 This form is used only to report observations. These observations do not constitute audit or inspectics findings and are not, without further verification and review,-indicative cf Tioncompliance withl10 r 7 50.45(b). These observations do not affect NRC certification or, approval. of--the simulation facility other than to provide informaticn which may be usedoin - future evaluations. No license action is required in' response to these - observations. During-the conduct of the simulator portion of the operating tests, the following items were observed:
- 1. Containment Pressure indications indicated below' minimum Technical l Specification LC0 limits. .The-indications-did not have an apparent-cause. Simulator operator was able to perform adjustments to>the instrumentation to correct for subsequent scenarios.E A permanent- ;
program change needs,to be performed in order to: eliminate the problem. '
- 2. ~
The Reactor Protection System (RPS) cabinets-are.not fully modeled in the simulator. Only' channels "A" and "C" have cabinets. Simulator RPS cabinets "A" and "C" do not_have the capability of bypassing trip signels, as in the actual plant. This prevents' fully testing the candidates skills in bypassing trip signals.
- 3. The simulator layout does not provide the NRC examiner with the capability to monitor telephone conversations between the simulated control room and the simulator operator.
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