ML20141C671
| ML20141C671 | |
| Person / Time | |
|---|---|
| Site: | 05561425 |
| Issue date: | 05/08/1997 |
| From: | Atomic Safety and Licensing Board Panel |
| To: | |
| Shared Package | |
| ML20141C667 | List: |
| References | |
| 97-725-02-SP, 97-725-2-SP, SP, NUDOCS 9705190142 | |
| Download: ML20141C671 (217) | |
Text
{{#Wiki_filter:- _ _ _ _ __ __ . _ _ _ _ _ . _ _ _ _ _ __. - . _ _ _ . l UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION . l BEFORE THE PRESIDING OFFICER ; Before Administrative Judge: G. Paul Bollwerk, III i (Thomas D. Murphy, Special Assistant) ' l i i In the Matter of )
) Docket No. 55-614-25-SP ,
FRANK J. CALABRESE JR. ) l
)
(Denial of Senior Reactor ) ASLBP No. 97-725-02-SP ; Operator's License) )
) . ;
1 l H E A RIN G FILE 10 C.F.R. f 2.1231 MAY 8,1997 NRC STAFF l b v 9705190142 970500 - PDR MISC 9705190137 PDR
i l l CN b CALABRESE HEARING FILE CONTENTS , Item Pages
- 1. NRC Form 398. Personal Qualifications Statement - Licensee 2 (the license application)
- 2. Form ES-303-1. Operator License Examination Report, for Mr. 8 Calabrese (see pages 3. 4. 6. and 7)
- 3. Form ES-303-1. Operator License Examination Report, for the 8 B0P operator during tile operating test (see pages 7 and 8)
- 4. Scenario Number Three used during the operating test 22
- 5. Scenario Number Two used during the operating test (see page 22 20 of 21)
- 6. Examiner's field notes (see page 3) 4
- 7. BOP examiner's field notes (see page 3) 4
- 8. Proposed denial letter frou Region I dated December 2. 1996 2 ,
/^\ 9. Applicant's request for a regrade dated December 19. 1996, and 3 Q 2-page attachment
- 10. Memo from Glenn Meyer to Stu Richai'ds dated January 16. 1997. 4 reviewing the applicant's contentions (see page 3)
- 11. Memo from Stu Richards to Tom Peebles. Paul Steiner, and Elden 1 Plettner dated January 30. 1997. establishing the appeal panel
- 12. Memo from T. Peebles. E. Plettner, and P. Steiner to Stu 6 Richards dated February 14. 1997. reviewing the applicant's contentions (see pages 4 and 5 of attachment)
- 13. Letter from Bruce A. Boger. NRC. to Calabrese, dated March 3. 7 1997, denying the appeal (see pages 4 and 5 of attachment)
- 14. Applicant's hearing request dated March 14. 1997 32
- 15. E0P flow charts 104. 112. and 113 (sheet 1) for Susquehanna 3 SES 2
- 16. Excerpted pages from Susquehanna " Operations Policies and Work 10 Practices" procedure OP-AD-001 (see page 34) 1 This document was requested from PP & L but not produced in time.
p) V Therefore. the copy in the file was obtained from the NRC regional or resident office. It may not match the one in effect at the time of the examination: however, it is unlikely that the guidance has changed.
i ; 1
. CALABRESE HEARING FILE CONTENTS .
Item Pages' l 2 i 2
- 17. Susquehanna Basis fc- E0P 100 (see page 5) 2 2
I 18. Susquehanna Basis for E0P 104 (see pages 1 - 4 and 17 - 21) 9 { 3 L 19. Susquehanna Basis for E0P 112 (see pages 1 - 5 and 13)' 6 2
- 20. Susquehanna Basis for E0P 113 (see pages 32 and 33) 3 !
- 21. NUREG-1021. Rev. 7 (Supplement 1). Operator Licensing Examiner 28
- Standards. ES-303 f 4
- 22. NUREG-1123, Knowledge and Abilities Catalog for Nuclear Power 23 ;
Plant Operators: Boiling Water Reactors, Chapter 1 and .; excerpts from Chapter 4 j i Total Pages 209 4 I
) \ !
l 2 5 i I l
1,
. -)
8 SIC FORM 358 U.S. NUCLLAR REGULATORY CouMISSaON AM' WOWED sv angit estL 31MM1000 DATI RECEfVED ESTIM ATED SURDEN PER RE T C Y wlTH THIS OfROMATION b W 3 .31. GS 36' y Cou.ECTION REQUEST: 1 MOUR. Mtc RfoulRES YMit DFORMATION TO ENeuRE THAT APPUCANTB/UCENSEES MEET ALL THE RFOUIREMENTS Folt i
- 1. QUAUFICATION STATEMENT-UCENSEE
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TO REMAIN VAUD,THIS FORM MUST NOT BE ALTERED ' Q , E 7a'm,% C Q " C8 " "*""'"' ""D
- 1. APPUCANT'S FULL NAME (Last, hrst, Msddlel AND ADOPESS (kecludeElf Code) 4. TYPE OF APPUCATION (Checa oppisce6se bonesj g l HOT l l COLD e'NEW l f. WAlVER REQUESTED (Justr4 on Remsel CALABRESE. FRANK J, JR -
- b. RENEWAL 1. WRffTEN (Categoryl 698 S KENNEDY DRIVE -
- c. uPGRAoE __.
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MCADOO PA 18237 1 d. MumuNrr <Auruo ro
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- 2. CTHZENSHIP 3. SIRTH DATE 2.SECOND 6 OTHER asowTN DAv vtAR '
- s. UNrfTO STATES 3. THIRD UN E TA MINATION SECTON tif AWCA8Lf) 5 !8
--" j b OTHERr$a= &s 1l 1 1J 9 l l
- 6. TYPE OF UCENSE APPUED FOR 6. PREVIOUS UCENSE(S) HELD e OPERATOR e. DOCKET NUMBkR RO sao b. UCENSE NUMBER l-gaer w d FACILITY DOCKET NUMBER I I I X b. SENIOR OPERATOR
.. uum D Sno te s. rues uend,e,, n61425 X OP-10944 0 18 115 91,7 S387, 50-388 t
7 NAME. AND ADDRESS OF APPUCANT'S EMPLOYER (include 2/P Codel 10. CURRENT POSETION AT FACILTTY l e PLANT SUPERINTENDENT L AUXILIARY UNIT OPER-Pennsylvania Power & Light Company -
- b. AS$lSTANT PLANT SUPERINTENDE NT ATOR/ TRAINEE / TURBINE BUILDING 4 OUIPME NT P.O. Box 467 -
OPERATOR ,=0= uCluS. Bsrwick, PA 18603 - 'D "*7 8'
- d. STAFF ENGINEER OTHER (Speer&J
[ C. NAWlE OF APPUCANT*S FACIUTV l f ACIUTY DOCKET NUMBER _
- e. SHIFT TECHNICAL ADVISOR /SHlfT ENGINEER Aonistnnt Cit nettichnnn n RFC 90_ % 7 f. INSTRUCTOR 11ni t Ettnv.
V- TDONAL FAC3UTY DOCKETS (Mutts-unit bcenses) g SENIOR CONTROL ROOM OPERATOR 50-388 -
- h. CONTROL ROOM OPERATOR
- 11. EDUCATION
- o. Moon sCoiOOL s. MAJOR AAE.AIS) OF STUDY ,",,"l* 'f,, ffl DEGREE CODES d VOCADONAL/TECHPMCAL "Y" IMi[ l (To be used tar
- HIGHEST " ' " * " # " ~
K GRAouATg E NGINE ERING 4f tELOSI nar cee=* Of GREE
- obteoned)
] GEo E OulvatE NCy oy,, UCERT nCArt EI""" " T9 & O1- % Y j NO 2. ASSOCIATE 3.DACHELOR h ggd 4. MASTE R ccuaor 6. DOCTORAL
- 12. FACILITY OPERATOR TRAINING PROGRAM e INPO ACCAfDITED OPtn ATOR TRAe4NG D CIRTWE D 80h N,RC 4 *sWtAA T80N TACiUTY PnoontM fuAt is sAsto uPoN A syatiMs YES NO rea resc.A reon OR50RM 41 nova o smutation r Aciufv 15 wRc Ap* YES NO APraoAcM 70 TR aid #NQ { USED es tHe OP(RAf DR TRA@NC, PMOvRAM y
1s. TnAmsNO tssNCE LASt APPUCAHON- SEEIN5MUCHONS) 14. LKPERIENCE IDO NOT DOUBLE COUNT - SEE INSTRUCHONSI l s otwAevon a uJWA e umrn amo vi an = imT 7rst in or mas enav To asum w I- NUCLE A3 POWER PLANT FUNDAMENTALS U 1-RO l 2 - Ptt*T SYSTE MS 2. E OOWSPWO l CLASSROOM 3.tWSPPWS i 0"SE CVATION i 4.ERSCRW 3 - OPE RATING PRACTICI 6 OTHER (Spechi CONTROL ROOM OPERAnONS ON $HIFT i FOSSIL i SIMULATOR OPE RATING (includes ClassroomJ 6 OPf RATOR SIMULATOR NAMES T SUPERVISOR U MSQIl"bH"UA SEb g . PLANT STMF j
- b. I mmmyg,w, 9 . OTHE R (SescWI PnaonAms ttes vio l l YTS l lNO COMMERCIAL NUCLEAR finclucfmp Reseerri b7est Reactori j
_gjaiQlM%tQN1 M*Awawag'TT%s F~ 10 RE ACTOR OPE RATOR fleceased) [ \ l
} 11. St NIOR OPE RATOR lticensed) 12 . SHIFT SUPE R VISOR (tscensed; h" rg Tvw on seert m contnot noou
{ ;- , , 13. ST AFF/SHsFT E NGINEE R iteceased.' 14 OPERATOR (Nonferensed)
- 16. PtANT STAFF 16 OTHER (Specr&l
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- 15. EXPEftIENCE DETAILS
- a. POSITION TIT 11 1 FROM TO b. FACILITY c. DUTIES G .
d
- 14. FOR ftENEWALS ONLY ts. DATE AND MSIILT OF MOST
- a. HOURS OPERATED FACIUTY: RECENT FACILITY PASS Fall HEOUAUNCATION EXAM
- 11. COMMENTE ISpecify the itern number to which you are elaborating. Attach addttional sheets if necessary.1 1.0 This is an address change since last application.
O < 1 1
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- 15. NRC FOflM 398. CERTIFICATION OF MF. DICAL e6amanaldATION SY FAC8tJTY UCENSEE. 88 ATTACHED YES I l
ANY FALSE STATEMENT OR OMISSION H THIS DOCUMENT, INCLUDING ATTACHMENTS MAY BE SUBJECT TO CML AND CRIMINAL SANCTIONS. toe. t certW under penemy of pequry that the 6nformation 6n thes document end attachmenes le true and correct. I further cendy that I have notated my currere employer of 0) all prewous employers;(2) eny 6netoisce where 4 have been tested ty a Health and Human Services (HHS) Ceruhed Drug Teeteng Lecoratory or e Licensee's testmg facitety for alcohol or a controlled substance. and the test
. reevRs escoeded the susoff 6evete estatAshed pursuant to 10 CFR Part M (31 any instense where I have been arrested for the sale, use or possession of a controlled subetance described in 10 CFR Part M and 44) any resenne for removal or revocahon of unescorted socess at a nuclear fac4hty,I e6so authertre the NRC to submq the results of enemmations to my employers for use in propanng retroeune programs, as necouary. jj ICANT I c-SIGNATURE 2-AA O. ' ~ , .
l DATE Solus CHEC8C APPUCABLE DOX \\ [ .] h I serary ones the above named mdeved e has successiuny completed the 'facekey heensees requirements to em heensed se en Operator /semor operator pursuant to Tme 19. Code of Federas Regutahons. Port SS, and that the 6ndnndual has e need for en Operator /Seruor Operetor hcanse to perform hismer essigned datees and that the facility will too made availab8e for enemmation. I sesa certdy under penalty of pequry that the informataan 6n this dor.ument and attachments le true and correct. I e. RE NEWAL DNLv . I certefy that the above named mdenduel meets the approved requeltiecation program (wsth eacepisons noted es item f ?/ as roquared by sectson 60.64 h.t) of 10 CFR to, and J that horshe has thscharged h6 Aer heensed respons6tuhtees competenity and safeiy. I eine certify under peneety of p.< jury that the informanon in thee document and enachments le true and sorrect;. j TRAINING COOftDINATOft SENIOft MANAGEMENT ftEPftESENTATWE ON SITE PRNTED OR TYPED NAME AND TTni PRINTED OR TYPED NAME AND TITLE , EHe LOwthert/ Manager-Nuclear Training C.J. Kuczynski/ Plant Manater StGNATUME S4GNATURE 44% % AA vum14J '1hdar t lDATE Af.L 4
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jDATE er s. 2 WAlVER (Check or Comptere items, es opphrebAel N MEETSREQUIREMENTS l l DOES NOT MEET REQLi4REMENTS fEspram betwJ <
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ES-303 Operator License Examination Report Form ES-303-1 I V U.S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSE EXAMINATION REPORT APPLICANT'S NAME Acut J. & b Btesf DOCKET NUMBER 55- /s/#M h R EXAMINATION TYPE (INITIAL OR RETAKE) FACILITY NAME52ra u h m,( REACTOR OPERATOR >( HOT SENIOR REACTOR OPERATOR (SRO) INSTANT COLD FACILITY SRO UPGRADE y BWR DESCRIPTION <
)(
SR0 LIMITED TO FUEL HANDLING PWR - WRITTEN EXAMINATION
SUMMARY
WRITTEN BY Fcchh
- TOTAL EXAMINATION POINTS 92. #
GRADED BY Fcu.a.% l'y e.c. TOTAL APPLICANT POINTS ~73 % f DATE ADMINISTERED 30 /2_, / qg APPLICANT GRADE ~]SiW 3 80* %
! MERATING TEST
SUMMARY
x ADMINISTERED BY 7, C/mso DATE ADMINISTERED /o/.u -/oM;/'74 A. ADMINISTRATIVE TOPICS J B. CONTROL ROOM SYSTEMS AND FACILITY WALK-THROUGH s C. INTEGRATED PLANT OPERATIONS (SIMULATOR TEST) y EXAMINER RECOMMENDATIONS CHECK BLOCKS PASS FAIL WAIVE SIGNATURE DATE WRITTEN EXAMINATION K [jM i,//#'6 OPERATING TEST y 9 [ // Y _ yleft FINAL RECOMMENDATION )( h/M% #/Aff, LICENSE REC 0'MMENDATION ISSUE LICENSE SIG U - ON. C DATE DENY LICENSE
,v M,j j k.
4 p 0-Examiner Standards 9 of 27 Rev. 7, January 1993 J
Fwk z && one ES-303 2 Form ES-303-1 APPLICANT DOCKET NUMBER: 55-(,p/pg PAGEp 0F 8 A. . ADMINISTRATIVE TOPICS - EVALUATION COMENT PAGE (S OR U) NUMBER-
- 1. CONDUCT OF OPERATIONS g 4' f
- 2. - EQUIPMENT CONTROL g
- 3. RADIATION CON 1ROL g
- 4. EMERGENCY PLAN- f _
8.1 CONTROL ROON SYSTEMS SAFETY JPM GRADE FUNCTION (S OR U) QUE5 TION CP.ADE (S OR U) O SYSTEM /JPM TITLE SYSTEM GRADE
~(S OR U)
COMMENT PAGE NUMBER l
- 1. groga Afgg, j .:yi My K s .s a 2 As.fMe A'HL S/h wbbs E .s s s
- 3. fa.,.oy,-//yK MM RMTm rre X s s s .
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- 7.
B.2 FACILITY WALK-THROUGH ~ 1 beh &.W s As2P 2n~ S s .s 2 Fist Pssr x-ne 7a ntse) M. s .s s ! 3. I j i' - Examiner Standards 10 of 27 Rev. 7, Sup. 1, June 1994
florsk 7'. 09fn$Df! ES-303 3.b Form ES-303-1 v APPLICANT DOCKET NUMBER: 55-4,/M2f PAGE 3 0F 6 C. 5ENIOR REACTOR OPERATOR INTEGRATED PLANT OPERATIONS (SIMULATOR TEST) GRADING
SUMMARY
CIMPETENCIE5/ SCENARIOS CopDIENT RATING FACTORS (TIGNT 3.0 2.0 1.0 TOTAL ,08 SERVED _ PAGE NO.
- 1. ALARMS / ANNUNCIATORS y'a f' Sco 2
av 3 A. PRIORITIZE 0.30 0.60 0.30 / / B. INTERPRET 0.35 0.70 0.35 / / C. VERIFY 0.35 0.70 0.35 (50) / /
- 2. DIAGN0515 1 2 3 A. RECOGN!!E 0.25 0.50 0.25 ./ ./
- s. ACCURACY 0.25 0.50 0.25 # #
C. I,lAGNosE 0.25 0.50 0.25 _# __,,. #
- 0. CREW RESPONSE 0.25 0.50 0.25 (14) #,#
- 3. SYSTEM RESPON?,E 1 2 3 A. INTERPRET 0.35 0.70 0.35 # #
B. ATTENTIVE 0.20 0.40 0.20 / 7 C. PLANT EFFECTS 0.45 0.90 0.45 (f.0) _ , , , , , 4 / 4. [ A. PROCEDURES REFERENCE 0.25 0.75 0.50 1 2 3
[#
(- s. CORRECT usE 0.50 1.50 1.00 h@ ' '
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C. CitEW INPLEMENTATION 0.25 0.50 0.25 # #~
- 5. CONTROL 80ARD OPERATIONS 1 2 3 A.
B. LOCATE 0.25 h 0.50 0.25 / MAN!PULATE 0.25 0.50 0.25 .,,,,,, C. RESPONSE 0.25 0.50 0.25 ,_ #- _
- 0. MANUAL CONTROL 0.25 0.50 0.25 (LO) ., ,. _, ., !
- 6. CapetuulCATIONS 1 2 3 A. CLARITY 0.45 0.90 0.45 # #
- s. CREW INFORMED 0.35 0.70 0.35 7 v ,,,
C. RECEIVE INFORMAfloh 0.20 0.40 0.20 (7.0) / ' /
- 7. DIRECTING OPERATIONS 1 2 3 A. TIMELY ACTION 0.20 0.60 h 0.20 #
B. C. SAFE DIRECTION 5 DVER$1GNT 0.40 0.20 1.20 0.A0 0.40 h0.20 # 8
- 0. CREW FEEDSACK 0.20 0.40 C.20 (M) #
- 8. TECHNICAL SPECIFICATIONS 1 2 ,3 '
A. RECOGNIZE 0.40 1. 0.80 0.40 _,,,_ B. LOCATE 0.20 0.40 0.20 C. CONPLIANCE 0.40 h 0.80 0.40 (I'9 ,,,,, Examiner Standards 12 of 27 Rev. 7, Sup. 1, June 1994
i i ~ APPLICANT DOCKET NO: 55 61425 PAGE 4 of 8 i s v OPERATING EXAMINATION
SUMMARY
The candidate did not perform satisfactorily in the simulator
; portion of the operating examination. The candidate failed to i refer to and use procedures correctly when directing ra)id RPV i i depressurization when it had not been determined that tie reactor <
would remain shutdown under all conditions without boron. He also. -
- failed to provide timely. well thought out directions that' ;
demonstrated appropriate concern for the safety of the plant, . staff, and public. . His incorrect use of procedures and t inappropriate direction were significant errors that unnecessarily degraded the condition of the plant. He was given ratings of *1* in referring to procedures, using procedures, and providing direction. He was evaluated as unsatisfactory in the competency < of compliance with and use of procedures. ! l. 1
?
I ') 1 i f i 4 1 i
1 4
- APPLICANT DOCKET NO: 55-61425 PAGE 5 of 8 FORM ES 302-2 Cross R,eference Comments A.1 The candidate was asked if refueling could continue if SRM D' I failed downscale with core refueling in progress, with bundle 49-12 to be moved from the fuel pool to the core. 33 bundles already loaded into the core, and all other SRMs operational. The candidate incorrectly stated that refueling could not continue because SRM D was located in the same quadrant as bundle 49-12.
> The correct answer is that refueling can continue because the SRM i in the quadrant that bundle 49-12 is located in and one SRM in an i adjacent quadrant are still operable. I The candidate's incorrect response was conservative and his weak understanding was not sufficient by itself to warrant an ' unsatisfactory grade in this area. a 8-4 I 9 }- i 1 l 1
~. - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ .__ _ _ _ _
f' . , APPLICANT DOCKET NO: 55 61425 PAGE 6 of 8 j FORM ES 302 2 Cross Reference Comments C.4.A The candidate failed to refer correctly to important procedures in important instances. The. candidate was acting in the position of the Senior Reactor Operator (SRO) during the major transient of the second scenario. The scenario involved fuel failure. a steam line break in secondary containment with seven control rods failing to insert. The candidate was in E0-100-104. " Secondary Containment Control" step SC/R-6 that directs when area radiation levels exceed maximum safe levels in two or more areas to rapidly depressurize the reactor. The candidate gave the order to open six ADS valves to depressurize the reactor without first referring to procedure E0-100-112. " Rapid Depressurization". Procedure E0- , 100-112. step RD-5 required action to stop and prevent all RPV ' injection prior to opening the ADS valves. The balance of plant operator (B0P) opened the ADS valves as i directed then indicated that he would need to secure the low pressure emergency core cooling systems (ECCS). The candidate . then gave the order to override all low pressure ECCS !' (approximately two minutes after the ADS valves were opened). By the time the order was given reactor pressure had already decreased to approximately 350 psig. The 80P completed the ' actions to override all low pressure ECCS systems. but one of the O low pressure coolant injection systems injected cold water into the reactor vessel before the pumps were secured. The candidate did not refer to E0-100-112 until after the low pressure coolant injection systems had been overridden. The candidate's failure to refer to E0-100-112 prior to directing action to ra) idly depressurize the RPV resulted in failure to stop and prevent RPV injection prior to dearessurization. As a result an injection of cold water occurred w1en it was not assured that the reactor would remain shutdown under all conditions without boron. The reactivity addition from the cold water injection could have caused a reactor power excursion and substantial core damage. The candidate failed to refer to the procedure in an important instance. K/A 295015 G.12 (3.7/4.4) E0-100-104 & E0-100-112 10 CFR 55.45(a)(13)
~
\
4 APPLICANT DOCKET N0: 55 61425 PAGE 7 of 8 FORM ES-302-2
- Cross Reference Comments C.4.B As described in C.4.A, the candidate while acting in the position ;
; of SR0 during the major transient of the second scenario failed to 1 use procedures correctly resulting-in significant errors that
- degraded the plant unnecessarily. When rapid depressurization is ,
required and it has not been determined that the reactor will ! remain shutdown under all conditions without boron, step RD-5 of E0-100-112 directs the operator to wait until all RPV injection is
,. stopped and prevented in accordance with ste) LO/L-19 of E0-100-113. " Level /)ower Control " before opening t1e ADS valves. Step LQ/L-19 of E0-100-113 directs the operator to stop and prevent injection except from SLC, CRD RCIC, and HPCI. The candidate's : ; direction to open the ADS. valves before the low pressure emergency coolant systems and condensate had been overridden was not in accordance with the direction in E0-100-112. .
The candidate's failure to correctly implement E0-100-112 resulted
. in an injection of cold water into the RPV when it was not assured '
4 that the reactor would remain shutdown under all conditions-
- ' without boron. The reactivity addition from the cold water :
injection could have caused a reactor power excursion and substantial core damage. The candidate made a significant error , in the use of procedures that degraded the plant unnecessarily. A K/A 295015 G.12 (3.7/4.4) E0-100-112 & E0-100-113 10 CFR 55.45(a)(13) l 1 I 6 i O i F
g APPLICANT DOCKET NO: 55 61425 PAGE 8 of 8 FORM ES-302 2 Cross, Reference Comments C.7.B While the candidate was acting in the position of the SRO during the major transient of the.second scenario, he failed to provide timely, well thought out directions that demonstrated appro)riate concern for the safety of the plant. During the scenario t1e candidate's two major concerns were inserting the rods that had failed to scram and monitoring increasing radiation levels. The candidate was in E0-100-104. " Secondary Containment Control." and 4 had assumed responsibility for monitoring secondary radiation levels. The candidate failed to monitor conditions closely and as a result, two areas had exceeded maximum safe radiation levels for a) proximately five minutes before it was recognized by the crew. W1en it was recognized that two areas had exceeded max safe levels, the candidate failed to provide well thought out direction to rapidly depressurize as discussed in C.4.A. The candidate's failure to provide timely direction to rapidly depressurize the RPV when radiation levels in two areas of , secondary containment were above max safe allowed radiation levels < in the secondar unnecessarily. y containment to continue to increaseAllowing radiation l control rod drive areas could have resulted in higher personnel .(q j exposures if operators had to enter the area to attempt to insert the control rods that were stil.1 withdrawn. The candidate's failure to provide well thought out direction for rapid depressurization when it had not been assured that the reactor would remain shutdown under all conditions without boron resulted in an injection of cold water into the RPV. The reactivity addition from the cold water injection could have caused a reactor power excursion and substantial core damage. K/A 295033 A2.01 (3.8/3.9) K/A 295033 G.12 (3.8/4.4) E0-100-104 & E0-100-112 10 CFR 55.45(a)(13) i I v
! Operator License Examination Report Form ES-303-1 t
/ ES-303 l
l U.S. NUCLEAR REGULATORY COMMISSION 1 OPERATOR LICENSE EXAMINATION REPORT DOCKET NUMBER 55-(,2o44 . l APPLICANT'S NAME I R EXAMINATION TYPE (INITIAL OR RETAKE) FACILITY NAME REACTOR OPERATOR 1 HOT COLD FACILITY X SENIOR REACTOR OPERATOR (SRO) INSTANT SR0 UPGRADE y BWR DESCRIPTION SR0 LIMITED TO FUEL HANDLING PWR -- j. WRITTEN EXAMINATION
SUMMARY
WRITTEN BY % d,k TOTAL EXAMINATION POINTS 4l M W GRADED BY E ,1,L / Ng TOTAL APPLICANT P0li(TS 13 W Y l*7 p DATE ADMINISTERED t[l2.ilc3 APPLICANT GRADE tb.t.~2$r.r"NI /# l OPERATING TEST
SUMMARY
ADMINISTERED BY T~m W hy DATE ADMINISTERED to I zz. - w / f6 A. ADMINISTRATIVE 10PICh 5 B. CONTROL ROOM SYSTEMS AND FACILITY WALK-THROUGH 0 C. INTEGRATED PLANT OPERATIONS (SIMULATOR TEST) EXAMINER RECOMMENDATIONS PASS FAIL WAIVE SIGNATURE DATE CHECK BLOCKS WRITTEN EXAMINATION K [ [/A/Mw shs/n OPERATING TEST )C f2Ma//w /ch/tc. FINAL RECOMMENDATION V $9[ //d4[94 LICENSE RECOMMENDATION ISSUE LICENSE SIGN < SE Of CH 3 ATE DENY LICENSE C , I %l Ilt)oh% C') /" 6 Rev. 7, January 1993 V Examiner Standards 9 of 27
ES-303 2 Form ES-303-1 APPLICANT DOCKET NUMBER: 55- (o2.o W PAGE Z OF S A. ADMINISTRATIVE TOPICS ' EVALUATION COMMENT PAGE (S OR U) NUMBER
- 1. CONDUCT OF OPERATIONS 6
- 2. EQUIPMENT CONTROL 6
- 3. RADIATION CONTROL 5
- 4. EMERGENCY PLAN 6 i
B.1 CONTROL ROOM SYSTEMS SAFETY JPM GRADE 1 FUNCTION (S OR U) QUESTION GRADE (S.OR U) 1 (' SYSTEM COMMENT l GRADE PAGE l SYSTEM /JPM TITLE (S OR U) NUMBER l
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k Examiner Standards 10 of 27 Rev. 7, Sup. 1, June 1994 i
l i ES-303 3.b Form ES-303-1 APPLICANTDOCKETNUMBER: 55- 6 M 4 PAGE 30F 8 C. SENIOR REACTOR OPERATOR INTEGRATED PLANT OPERATIONS (SIMULATOR TEST) GRADING
SUMMARY
COMPETENCIES / SCENARIOS C0f04ENT RATING FACTORS WE!CHT 3.0 2.0 1.0 TOTAL 08 SERVE 0 PAGE Wo.
- 1. ALARMS / ANNUNCIATORS 1 2 3 A. PRIORITIZE 0.30 @ 0.60 0.30 2 2 d B. INTERPRET 0.35 @ 0.70 0.35 / / Z ,,_.
C. VERIFY 0.35 @ 0.70 0.35 (f.0) / / 4 _
- 2. 0!AGNOSIS 1 2 3 A. RECOGNIZE 0.25 @ 0.50 0.25 g M 2 _,,,,.
B. ACCURACY 0.25 @ 0.50 0.25 g Z g ,_, C. O!AGNOSE 0.25 @ 0.50 0.25 g / / D. CREW RESPONSE 0.25 0.75 @ 0.25 d.7f) /
- 3. SYSTEM RESPONSE 1 2 3 A. INTERPRET 0.35 h 0.70 0.35 g g / _,,_
B. ATTENTIVE 0.20 0.60 @ 0.20 g 2 2 _ktf 7 C. PLANT EFFECTS 0.45 1.35 0.90 @ (l.9 ) 2 d d $$
- 4. PROCEDURES 1 2 3
( A. REFERENCE 0.25 @ 0.50 0.25 I 2 / B. CORRECT USE 0.50 Q 1.00 0.50 2 2 d C. CREW IMPLEMENTATION 3.25 0.75 @ 0.25 (2.765 g ,_ _ ,
- 5. CONTROL BOARD OPERATIONS 1 2 3 A. LOCATE 0.25 @ 0.50 0.25 2 2 _
B. MANIPULATE 0.25 @ 0.50 0.25 / I _ C. RESPONSE 0.25 @ 0.50 0.25 ,__, d 2 _,_ D. MANUAL CONTROL 0.25 @ 0.50 0.25 G.o ) / 2 .,_ l 6. COMMUNICATIONS 1 2 3 I A. CLARITY 0.45 O 0.90 0.45 ./ / / ,_,,, B. CREW INFORMED 0.35 1.05 @ 0.35 / / 2 _,, C. RECEIVE INFORMAfl0N 0.20 0.60 @ 0.20 (l.4 g g d ,_,,,
- 7. L.!RECTING OPFRATIONS 1 2 3 A. TIMELY ACT10N 0.20 0.60 @ 0.20 g g 2 ,,,,_,,
B. SAFE DIRECTIONS 0.40 1.20 @ 0.40 / C. DVER$1GHT 0.20 0.60 @ 0.20 / _ ,__, ,,,,_, 4 j 0. CREW FEEDBACC 0.20 @ 0.40 0.2J (Z2) / ,,,,,,, _ ,,,,_
- 8. TECHNICAL SPECIFICATIONS 1 2 3 A. RECOGNIZE 0.40 6 0.80 0.40 ,,/_, / g ,,,,,,,
B. LOCATE 0.20 0.60 0.20 / [_
@0.80 C. COMPLIANCE 0.40 h 0.40 (2.7) g 2 F
i ( Examiner Standards 12 of 27 Rev. 7, Sup. 1, June 1994
~ ;
l l i 2 APPLICANT DOCKET NOe. 62044 PAGE 4 of 8 0 \ 4.O FORM ES-303-2 Cross Reference Comments B.1.2 During a job performance measure (JPM) to override an inadvertent HPCI ; i initiation, the applicant failed to place the HPCI auxiliary oil pump switch to ! stan as required by step 3.9.5.a of OP-152-001, "High Pressure Coolant l Injection (HPCI) System." The applicant correctly stopped the HPCI injection by placing the flow controller on minimum flow. Then he was asked if HPCI : was in a stable condition. At this time he placed the auxiliary oil pump switch . to start. He was then asked if there was procedural direction for overriding an inadvenent HPCI initiation. He referred to section 3.2 of OP-152-001 for automatic / manual startup of the HPCI system and indicated that a caution provided direction to secure HPCI if misoperation in automatic was confirmed. , He failed to refer to section 3.9 of OP-152-001 for overriding HPCI injection. Failure to refer to the correct section of the procedure could have prevented the applicant from placing the auxiliary oil pump switch in the start position. ; Failure to place the auxiliary oil pump switch in the start position could have prevented the auxiliary oil pump from starting as required on low oil pressure if HPCI dropped below approximately 2150 rpm without an initiation signal present. The applicant's failure to immediately place the. auxiliary oil pump m switch in the start position (a non-critical step) and failure to identify the correct section of the procedure for overriding a HPCI injection were not sufficient to be considered unsatisfactory performance on the JPM. K/A 206000 A2.17 (3.9/4.3) ! OP-152-001, "High Pressure Coolant Injection (HPCI) System," pg 47. l B.I.6 The applicant was directed to manually synchronize diesel generator (DG) A with Unit 2 4.16 KV bus 2A and to pick up 4000 KW of load. The applicant successfully synchronized the DG with the bus and picked up approximately 1000 KW of load. While increasing load to 2000 KW, he stated that he could l not increase load any further because, in accordance with a caution in section - 3.3.4 of OP-024-001, " Diesel Generators," the DG output could ne' exceed the present transformer T-10 bus load. He indicated that DG output was approaching 200 amps which was the present load on transformer T-10 busses. The applicant incorrectly' interpreted amperage as indication of bus load and i would have unnecessarily secured the DG. The actual T-10 bus load was ) approximately 5000 KW which was sufficiently above the anticipated DG j output of 4000 KW. The applicant failed to pick up 4000 KW of load as j directed; therefore, he was evaluated as unsatisfactory on this JPM. ! K/A 264000 A4.05 (3.6/3.7) 10 CFR 55.45(a)(8) l' OP-024-001, " Diesel Generators," pgs 18,19, and 23 - 26. 1 4
.~ - . - - . . - . - . . -
+
t 7, APPLICANT DOCKET NO.': 62044 PAGE 5 of 8 i ,
\-- FORM ES-303-2 i Cross Reference ' Comments .
. B.2.1 The applicant was asked what effect allowing the scram air header to remain 4 vented following a failure to scram would have on other actions to insert contro!' rods. The applicant correctly indicated that the scram' discharge y volume (SDV) vent and drain' valves could not be opened to allow the scram 3
- discharge volume to drain. He incorrectly stated that control rods could be driven manually. The correct answer was that control rods could not be- i driven manually due to loss of air to the CRD flow control valves. The applicant's incorrect answer to part of one question was not sufficient to .
warrant an unsatisfactory grade on the questions for the CRD system. K/A 201001 K6.03 (3.0/2.9) ; SY017 K-2, " Control Rod Drive Hydraulic System," pg 14, LOs 3f and 11f. l B.2.2 The applicant was asked why it is necessay to install temporary RPV level ] indication following a control room evacuation. He indicated that the range of i I , the wide range level indication available on the remote sifutdown panel (RSDP)
- ( was not broad enough to monitor RPV water level once the plant was in cold
.is shutdown. He stated that RPV water 1: vel would be raised above +60 inches,
.'the high end of the wide range level indication The correct answer is that wide range level indication is increasingly less accurate as RPV pressure .
decreases. At approximately 200 psig, the wide range level instrument will indicate +60" (upscale) when actual RPV level is 0". The applicant's lack of. . understanding of the effect of RPV pressure on the level instrument could have. I led to nonconservative actions for RPV level control as the RPV l depressurized. The applicant's lack of understanding was not sufficient to I warrant an unsatisfactory grade on the questions for the RSDP. l l K/A 295016 A2.02 (3.2/3.6) l ON-100-009, " Control Room Evacuation," pg 13. ! i W 1 o N' j l i
P t ~ APPLICANT DOCKET NO.; 62044 . PAGE 6 of 8. ; FORM ES-303-2 i Cross keference Comments , 4 C.3.b During the first scenario, the applicant was acting as the Unit Supervisor (US) when a break occurred in an RPV instrument reference leg. The reference leg i break caused a loss of feedwater and a reactor scram. It also caused elevated l i temperature and pressure in the primary containment. RPV level was being controlled with RCIC, when a RCIC controller failure occurred resulting in a steady RPV level decrease. The applicant directed the reactor operator (RO)- I to lower pressure to 600 psig so that the condensate system could be used for injection. Given these conditions, the applicant should have anticipated that RPV level could decrease below -129" and the effect of the automatic actions that would occur. He failed to monitor RPV level closely enough to ensure that condensate would be able to inject prior to reaching -129" and gave no direction to the crew to anticipate transferring pressure control from the 4 turbine bypass valves to the safety relief valves (SRVs) if the MSIVs closed due to RPV level below -129". When RPV level reached -129", the applicant directed the balance of plant h V operator (BOP) to override the low pressure ECCS systems. Less than one minute later, RPV level dropped below the top of active fuel (TAF) and the applicant directed the operators to lineup low pressure ECCS and reduce pressure with SRVs. IIe failed to anticipate that ADS would automatically initiate 102 seconds after RPV level dropped below -129" and that the low pressure ECCS systems would be needed to restore adequate core cooling. The applicant also failed to anticipate that some automatic actions might not occur due to the instmment failures (due to the break in the RPV instrument leg). The operators manually initiated the actions that did not occur even though they were not given any direction to anticipate that the actions would not occur. K/A 295031 K2.12 (4.5/4.5) K/A 295031 A1.06 (4.4/4.4) 0 O o
. . - . - _ _ . ~ - .- . - - . - - - . - - - - - - . - -
l l i PAGE 7 of 8
. APPLICANT DOCKET NOa 62044 FORM ES-303-2 l Cross Reference Comments !
1 C.3.b During the first scenario, the applicant failed to closely monitor RPV pressure ; and RPV instrument run temperature to ensure that the RPV saturation < temperature curve was not exceeded. He checked the RPV saturation curve l several times prior to the initiation of ADS. However, he did not check the ! I
< RPV saturation curve for nine minutes from the time that ADS initiated until the end of the scenario. It was important to check the RPV saturation curve as
, the RPV depressurized because RPV saturation temperature decreases as RPV ; pressure decreases. ! l~ During followup questioning, the applicant was asked to determine if RPV saturation temperature had been reached at the end of the scenario. He stated that, with instrument run temperature at 247*F and RPV pressure at i approximately 0 psig, RPV saturation temperature had been exceeded and that RPV level could not be determined. His assessment appeared to be ! conservative, because during the scenario the RPV level instruments did not l . show any indication of RPV saturation. However, the applicant's failure to l monitor the RPV saturation curve following RPV depressurization could have - resulted in failure to identify that RPV water level could not be determined. l 1 K/A 295028 A2.03 (3.7/3.9) None of the applicant's failures to anticipate predictable changes in plant response resulted in actual degradation of plant conditions; therefore, the applicant was given a '2' in remaining attentive to control room indications. C.3.c During the second scenario, the applicant was acting as the BOP, when he was directed to open six ADS valves to rapidly depressurize the RPV due to two secondary containment area radiation levels above the max safe level. At the time of the direction, seven control rods were not inserted and reactor pressure was approximately 1000 psig. After the applicant opened six ADS valves as directed, he informed the US that he would need to shutdown the low pressure emergency coolant system (ECCS) pumps. He was then directed to override low pressure ECCS. At the time of the direction RPV pressure was approximately 350 psig. He correctly overrode the low pressure ECCS systems by initiating the systems and then securing the pumps. However, because RPV pressure was below 436 psig, the injection valves opened and RHR A injected some water into the RPV before the pumps were secured. r i
[ APPLICANT DOCKET NO;: 62044 PAGE 8 of 8 N FORM ES-303-2 , Cross Reference Comments l C.3.c Step RD-5 of EO-100-112, " Rapid Depressurization," does not allow the ADS valves to be opened until RPV injection has been stopped and prevented when ! it has not been determined that the reactor will remain shutdown under all conditions without boron. The applicant failed to recognize that the low , pressure ECCS systems had to be overridden before the ADS valves were opened in order to prevent an injection of cold water into the RPV. Even , though he recognized the need to shutdown the low pressure ECCS pumps ! after the ADS valves were open and properly overrode the systems when directed, it was too late to prevent a cold water injection. The reactivity addition from the cold water injection could have caused a reactor power. . excursion and substantial core damage. The applicant's action to open the ADS valves before low pressure ECCS system injection svas prevented was a serious mistake that unnecessarily degraded the condition of the plant and merited a rating of '1' in demonstrating a thorough understanding of how the i f plant, systems, and components operate and interact. K/A 295015 K1.04 (3.8/3.8) 10 CFR 55.4.9(a)(8) p EO-100-112. " Rapid Depressurization." EO 100-113, " Level / Power Control." i j During the first scenario, the applicant was acting as the US when a HPCI l pipe routing area temperature element failed upscale. The crew correctly ] determined that the high temperature indication was due to an instrument failure rather than an actual high temperature condition. However, the applicant failed to recognize that the indicated high temperature would cause HPCI to isolate after a 15 minute timer timed out. As a result of the - applicant's failure to recognize that the instrument failure would render HPCI i inoperable, he did not identify that a reactor shutdown was required due to the l inoperability of one RHR system and HPCI during the scenario. (He correctly : f identified that a reactor shutdown was required during followup questioning.) The applicant's failure to recognize that a HPCI isolation would occur was a l minor error, but contributed to the rating of '1' in demonstrating a thorough understanding of how systems and components operate. , K/A 206000 K4.02 (3.9/4.0) 10 CFR 55.45(a)(12) , v , I ( b 4
. l Au hment 1 l AN 4 PP&L-SUSQUEHANNA STCP-QA-612 4
- Rev 1 g g *. TRAINING CENTER Page 1 er 17 3 mr PP&L SIMULATOR SCENARIO ic . i Scenario
Title:
Plant Startup, FW Flow Detector Failure, Place RFPT in Service, Loss of1Y115, Loss of CRD, ATWS Scenario Duration: One hour Scenario Number: Three Revision /Date: 1,10/9/96 m:: .::e n. M *1 " > :s 'l+W sss~a%c; th :: % l$$k%$?Niin%W $b?8?SW'. . Prepared By:
/#!7 96 instructor Date y@ $
e ' Date Nuclear Ope;4tions Training Supervisor Approved By: Modd /o-/CM Date Supervising Manager / Shift Supervisor W. lk
d Attachment 1 Scen rio Throo STCP-OA-612 Rev.1,10/9/96
- Rev.1 Page 2 of 25 Page 15 of I
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Scenitio Thres Attachment 1 Rev.1,10/9/96 STCP-Q A-612 Page 3 of 25 Rev.1
- Page 3 of 17 SCENARIO
SUMMARY
The plant is operating at 38% power with 49 Mlbm/hr core flow. The "A" RFPT is in standby with the other two RFPTs in service in three element control. Power ascension with recirculation flow is in progress. The "A" APRM is failed upscale and is bypassed and the *B" EHC pump is out of service. Event one, The PCOU will raise recirculation flow to 55 Mlbm/hr. The PCOU will be primarily involved in this reactivity manipulation. Event two. The supply breaker to 1D115 trips open causing a loss of 1Y115. The crew will diagnose the problem and implement ON 117-001, Loss of instrur'nent AC and transfer the 1Y115 loads with the transfer switches. When the transfer switch for SPOTMOS is transferred, its altemate feed will trip too. The US will determine the Technical Specifications actions for this loss. Both PCOs and the US will be involved in this lastrument failure. Event three. The "B" feedwater flow instrument fails low causing total indicated feed flow to drop below 20% causing a recirculation pump runback to the number 1 limiter. RPV level will increase until it offsets the flow mismatch (about 44 inches). The crew will diagnose the problem and implement ON-145-001, Reactor Water Level Control System Malfunction and ON-164-002, Loss of Recirculation Flow. The PCOU will place feedwater in single-element control. Both PCOs and the US will be actively involved in this instrument failure. O Event four. The PCOX places the "A" RFPT in seivice and the "B" RFPT in standby to restore normal feedwater flow indication until the *B" flow transmitter problem is resolved. Once the "A" RFPT is in service, indicated feedwater flow will be greater than 20% allowing the runback to be reset and core flow to be restored to 55 Mlbm/hr. The PCOX and US will be primarily involved in this normal activity. Event five. An accumulator fault occurs on rod 18-35. The PCOU will respond per the annunciator response procedure and direct an NPO to investigate. The PCOU will then direct the NPO to recharge the accumulator. The US will determine the Technical Specifications actions required. The PCOU and the US will be primarily involved in this component failure. Event six. CRD pump A trips causing a loss of CRD. The crew will attempt to start the standby pump, but its breaker will fail to close. Within a few minutes, another accumulator trouble alarm will be received requiring the mode switch be placed to SHUTDOWN. Both PCOs and the US will be involved with this component failure. Event seven. The reactor fails to scram and the ARI valves will fail to vent the scram air header. The crew will enter EO-100-113, Level / Power Control. The crew will initiate SLC but the "A" SLC pump shaft shears cnd the "B" pump shaft seizes, so RCIC must be used to inject boron. The crew wil! lower RPV level to <- 60 inches and manually control RPV pressure with the SRVs. When level is stable, a small coolant break will occur in the drywell. Drywell pressure will increase requiring entry into EO-100-103, Primary Containment Control. The crew willinsert rods using EO-100-113 sheet 2 by pulling RPS fuses and venting the scram air header. When the crew vents the scram air header or pulls the RPS fuses, the rods Gwill fully insert. Both PCOs and the US will be involved in the two major transients. The PCOX will be nvolved in the component failure of SLC failing to initiate.
Attachment 1 , Scenarb Thr.",e STCP-QA-612 Rev.1.10/9/9s Rev.1 . r- age , of 25 Page 15 of 17 e v s l 6 l wuwmumwrmmrs:mamm. mm=mug BTHIS PAGE IS;tNTENTIONAL,m.LY4LEEEBt!ANK'd. y z .< 1g+nw
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Scenario Thr:e Att: chm::nt 1 Rev.1,10/9/96 GTCP-QA-612
. Page 5 of 25 Rev.1 Page 5 of 17
-l SCENARIO KNOWLEDGES AND ABILITIES General:
294001 A102 4.2/4.2 294001 A104 3.1/3.2 j 294001 A105 3.4/3.8 294001 A109 3.3/4.2 294001 A110 3.6/4.2 294001 A111 3.3/4.3 l 294001 A112 3.5/4.2 294001 A113 4.5/4.3 294001 A115 3.2/3.4 Event 1: Raise core flow to 55 Mlbm/hr 202002 A105 3.6/3.6 202002 A407 3.3/3.2 202002 A207 3.3/3.3 202002 G013 3.6/3.4 Event 2: Loss of instrument bus 1Y115 262001 G012 3.1/3.0 262001 G015 3.5/3.8 262001 G005 2.7/3.5 262001 A401 3.1/3.0 262001 G012 2.9/3.1 Event 3: Feedwater flow element B fails low l 263000 G015 3.4/3.8 216000 G011 3.2/4.2 216000 A206 2.9/3.1 Event 4: Place RFPT A in service and RFPT B in standby 259001 G013 3.6/3.4 259001 A402 3.9/3.7 Event 5: Loss of CRD l 295022 AA201 3.5/3.6 295022 G010 3.7/3.5 295022 G011 3.9/4.0 Event 6: ATWS 295037 EA101 4.6/4.6 295037 EA103 4.1/4.1 295037 EA110 3.7/3.9 295037 EA201 4.2/4.3 295037 EA202 4.1/4.2 295037 EA205 4.2/4.3 295037 G011 4.4/4.7 295037 G012 3.9/4.6 V
Attachment 1
- Scenario Three STCP.QA-612 l Rev.1.10/9/96 Rev.1 Page 6 of 25 Page 15 of 17 O
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Scenario Three Attachm:nt 1 STCP-QA-612 j R;v.1,10/9/96
. _Page 7 of 25 Rev.1 i Page 7 of 17 l REFERENCES
- 1. AR-103-H06, AccumulatorTrouble
- 2. EO-100-102, RPV Control, revision 7
- 3. EO-100-103, Primary Containment Control, revision 7
- 4. EO-100-113, Level Power Control, revision 7
- 5. EP-PS-100, Ernergency Diiector-Control Room, revision 12
- 6. ES-150-001, Boron injection Using RCIC, revision 9
- 7. ES-158 001, Deenergizing Scram Pilot Solenoids, revision 3
- 8. GO-100-002, Plant Startup, Heatup, And Power Operation, revision 25
- 9. 01-055-001, Scram Accumulator Log, revision 2 10.ON-117-001, Loss of Instrument Bus, revision 15 11.ON-145-001, RPV Level Control System Malfunction, revision 7 12.ON-164-002, Loss Of Recirculation Flow, revision 14 13.ON-155-007, Loss Of CRD System Flow, revision 12 14.OP-145 001, RFP And RFP Lube Oil System, revision 22 l
- 15. Technical Specifications 3.3.7.5-1, 3.1.3.5
- 16. Technical Specifications Interpretation 1-96-007 l I
t t l l Attschment 1 Scen do Th:22 Rev.1,10/9/96 . STCP-QA-612 l - Rev.1 l Pape 8 of 25
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]. Scenario Thme Attachment 1 I a Rev.1,10/9/96 STCP-QA-612 ;
. Page 9 of 25 Rev.1
!- Page 11 of 17 - l l SCENARIO SPECIAL INSTRUCTIONS I ! 1. Reset the simulator to IC-131. Ensure rod step at A1-524. Ensure B and C RFPTs are in ] service in 3-element and A RFPT is in standby. - i i ! i 2. Execute preference file, YPP.SC.3, establishing the following conditions: i ! Malfunctions: 5:5 :
. NM178007A 125 APRM A failed upscale ;
~
4 . PM05:1P208A SLC pump shaft shear.
. RP1580078 RPS B failure to scram
. PM10:1P132B CRD pump B breaker fail as-is
. PM03:1P208B SLC pump motor overcurrent Remote Functions: 3 :
. PM101P1138 OUT EHC pump B breaker racked out
. R?155018100 ARI air supply isolation valves bypassed e RD155030 0 ARI vent valve isolated j
j Overrides: 0:0 l
. ZDlHSE111502 NORMAL HSE-111502 Stays in Normal Position i Triocers: 0 i
j _P ushbuttons Assionments l 1. MRF DC102149 OPEN 1D614-30 Trips Open Deenergizing 1Y115 l 2. MRF DB157001 ALT HSE-111505 Transfer Switch To Attemate
- 3. IMF TR02:FTC321N002B 0 Feedwater flow instrument B fails low
' ~
- 4. IMF RD1550191835 Accumulator Trouble Rod 18-35
- 5. IMF PM03:1P132A CRD Pump A Motor Overcurrent Fault i 6. IMF RD1550191043 Accumulator Trouble Rod 10-43 i
- 7. IMF RR164010 5 5:00 Bottom Head Drain Leak,5% Over 5 Mins.
. 5. Turnover information: Prepare a turnover sheet indicating continue plant stariup at step 6.91 of GO-100@2. Recirculation flow is 48 Mlbm/hr. Power is 38%. Rod step A1-524. Indicate ! that APRM A is failed upscale and is bypassed and the "B" EHC pump is out of service for motor replacement. i
- 6. Place a status control tag on EHC pump B.
O
l J Attochmont 1 Scenario Threa STCP-QA-612 l Rev.1.10/9/96 Rev.1 Page 10 of 25 Page 15 of 17 1 I
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/\ \ SCENARIO EVENT DESCRIPTION FORM Initial Conditions: Reactor power is 38% and core flow is 49 Mlbm/hr. APRM A is upscale and bypassed and EHC pump B is out of service.
RX MANIP TIME DESCRIPTION R Raise core flow to 55 Mlbm/hr i Loss of 1Y115 and attemate supply to HSE-111502 (SPOTMOS power) i Feedwater flow element B fails low M Place &, ": A lr. ee. ;ce1tred Rf PT S M Maadby p qq, l C Accumulator trouble rod 18-35 C Loss of CRD M/C ATWS, SLC pump failure M Small break LOCA . O I O
Attichment 1 STCP-Q A-612 l Sc:Jnario Thr:2 l Rev.1,10/9/96 Rev.1 Page 14 of 17 l Page 12 of 25 SCENARIO EVENT FORM (EVALUATION) Event No: 1 Bdef
Description:
Raise recirculation flow to 55 Mibm/hr. P l. POSITION TIME STUDENT ACTIVITIES PCOU Raises core flow to 55 Mlbm/hr per GO-100102, Plant Startup and , reactivity manipulation book _j Plots position on power to flow map 4 e US Supervises reactor powerincrease NOTES: m M @
Scenario Three Attachment 1 R;v.1,10/9/96 STCP-QA-612
. Page 15 of 25 Rev.1
-- Page 15 of 17
- INSTRUCTOR ACTIVITIES, ROLE Pl AY, i
AND INSTRUCTOR'S PERSONAL NOTES Event No: 2 Brief
Description:
Supply breaker to 1D115 trips causing a loss of 1Y115 and subsequently the alternate supply to SPOTMOS transfer switch trips causing a loss of SPOTMOS div.1 INSTRUCTOR ACTIVITY:
)
When the PCOU has raised core flow to 55 Mlbm/hr or as directed by the lead evaluator
- l Depress P-1, modifying remote function DC102149 OPEN.
This opens the supply breaker to the 1D115 inverter. When directed to transfer the HSE-111505 at 1C661: J Depress P-2, modifying remote function DB157001 ALT . This transfers HSE-111505 to attemate. ROLE PLAY: As NPO sent to inverter 1D115, wait two minutes then report that all of the lights on the inverter are out. As NPO sent to 1D614030, wait two minutes then report the breaker is tripped.' If directed to reclose it, report it immediately retrips. As the NPO sent to 1Y216-26, report the breaker is tripped, if directed to reclo'se it, report it immediately retrips. As electrical maintenance contacted about the 1D115 inverter and SPOTMOS alternate supply breaker tripping, wait five minutes then report it will take a while to evaluate the cause of the failures. J V
Attichm:nt 1 ! Scenario Thr33 STCP-QA-612 Rev,1,10/9/90 , Rev.1 ; Page 14 of 17 , Page 16 of 25 l SCENARIO EVENT FORM (EVALUATION) 1 Event No: 3 Brief
Description:
FW flow instrument B falls downscale. P N/P POSITION TIME STUDENT ACTIVITIES PCOU / 'Mowledge RPV high level annunciator aose problem as a failure of FW flow B instrument j Respond per ON-145-001, FWLC Malfunction
- (
Transfer FWLC to single element. Restore RPV level to 35 inches Responds per ON-164-002 Loss Of Reactor Recirculation Flow l 1 Plots position on power to flow map PCOX Dispatch NPO to check transmitter I Report no apparent problems with transmitter US Direct actions per ON 145-001, FWLC Malfunction and ON-164-002, , Loss Of Reactor Recirculation Flow Contact l&C for support on problem f l 1 i NOTES: f 1
J
- i. ,
l Scen rio Three AttzchmInt 1 ' Rev.1; 10/9/96 STCP-QA-612 i
. Pa0e 17 of 25 Rev.1 Page 15 of 17 INSTRUCTOR ACTIVITIES, ROLE PLAY, l AND INSTRUCTOR'S PERSONAL NOTES ,
i t j + 3 Event No: 3 Brief
Description:
FW flow instmment B fails downscale. INSTRUCTOR ACTMTY: i ARer the crew has completed restoring the loads of 1Y115 and addressing the Technical SpecrRcations for the inoperable SPOTMOS division or when directed by the lead evaluator: j Depress P-3, inserting malfunction TR02:FTC321N0028 0 , l 1 This will fail the FW flow B transmitter downsct.le. ROLE PL.AY:
- As the NPO sent to check the flow transmitter, wait two minutes then report there are no apparent problems j As l&C, after five minutes report the transmitter is failed and it will take about four hours j d replace the transmitter.
As plant management, when contacted about the delay in the startup, ask the US for his recommendation on continuing the startup. If the startup doesn_'t continue, prompt the US to swap the A and B RFPs to allow the runback to be reset and the startup to continue. ? i . l y I i
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Attichment 1 STCP-QA-612 Scen:rio Thr::e Rev.1,10/9/06 Rev.1 Page 14 cf 17 . Page 18 of 25 SCENARIO EVENT FORM (EVALUATION) Event No: 4 Brief
Description:
Place RFPT A in service and place RFPT B in standby. P N/P POSITION TIME STUDENT ACTIVITIES ' Places RFPT A in service per OP-145-001, RFP And RFP Lube Oil PCOX System increases speed until pump begins to faed. Closes recirc flow valve. Nuits controller and places in automatic. Balances flow between pumps Places RFPT B in standby per OP-145-001. Opens recirc flow valve to establish minimum flow protection Lowers speed to establish discharge pressure ~100 psig below reactor pressure Restores FWLC to three element PCOU Resets recirculation pump runback l US Directs placing RFPT A in service and RFPT B in standby. l
' Directs transferring back to three element control {
Resets recirculation pump runback l l l I I I I NOTES: Plant management may need to direct swapping the A and B RFPTs to allow the runback to be reset and the startup to continue _
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. l 1
Scenario Thr:o Attachment 1 Rev.1,10/9/96 STCP-QA-612 Page 19 of 25 Rev.1 3 -- Page 15 of 17 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 4 Brief
Description:
Place RFPT A in service and place RFPT B in standby. , INSTRUCTOR ACTMTY: No instructor activity required for thiS event. ROLE PLAY: . As necessary i I l l l l i 4 9 s
Attichmen' 1 Scenario Three STCP-QA-612 Rev.1,10/9/96 Rev.1 Page 20 cf 25 Page 14 of 17 - J SCENARIO EVENT FORM (EVALUATION) Event No: 5 Brief
Description:
Accumulator fault on rod 18-35 l P N/P l POSITION TIME STUDENT ACTIVITIES l PCOU 10wledges annunciator AR-103-H03, Accumulator Trouble
. entify affeded rods as 18-35 Dispatch NPO to investigate l Reports Nitrogen leak to the US i insert control rods per CRMR US Contad mechanical maintenance i
Evaluate Technical Specifications 3.1.3.5, determine the rod must be dedared inoperable if not fixed within 8 hours Contad reador engineering about inserting rod 18-35 Dired insertion of rods per CRMR from reador engineering _ l I I
. I I
l I NOTES: O
< i Scen rio Thre3 Attachm:nt 1 ,
Rev.1,10/9/96 STCP-QA-612 ' Rev.1
- . Page 21 of 25 Page 15 of 17
' - INSTRUCTOR ACTMTIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 5 - Bdef Desedption: Accumulator fault on rod 1s-35 l INSTRUCTOR ACTMTY: ,
- When the PCOX has swapped RFPs or as directed by the lead evaluatoc
! Depress P.4, inserting malfunction RD1550i91835 .
)
5 This causes an accumulator fault on rod 18-35
- ROLE PL.AY:
As NPO sent to HCU 18-35, report pressure is 850 psig, dropping. Also report you can hear
- an air leak that sounds like it is coming from the bottom of the accumulator As mechanical maintenance sent to HCU, report a weld is leaking Nitrogen and the HCU will have to be isolated to repair it. The repairs will probably take at least 12 hours.
As reactor engineering, acknowledge the request for guidance on inserting the rod. Tell the US you will be up in about ten minutes with a CRMR for inserting rod 18-35 and its companion rods. . 4 ; i I 1 4 1 l
l Attrchment 1 STCP-QA-612 l Scenario Tht:0 Rev.1 t Rev.1.10/9/96 , , . Page 14 of 17 Page 22 of 25 l SCENARIO EVENT FORM (EVALUATION) i _l Event No: 6 ; Brief
Description:
Loss of CRD
. I P N/P POSITION TIME STUDENT ACTIVITIES i
PCOU Monitor full core display for accumulator faults ! Repor ac::umulator faults as they occur Scram the reactor whea "1e second fault is received (one on a withdrawn control rod) Report failure of the resuorto scram due to failure of RPS B to actuate l PCOX Recognize trip of the operating CRD pump l Attempt to restore CRD by manually closing the tiow control valve and startin0 the standby pump ' Report failure of the standby pump to start. l Dispatch NPO to check out the pump Dispatch NPO to check out the pump breaker Manually initiate ARI 1 l Repor+ failpre of scram air header to depressurtze. US Direct actions per ON-155-007, Loss of CRD System Flow . l Direct manual scram of reactor 1 l 1 l I l l NOTES: ] l 1 1 1 I l l l
- ~ _ . .
e, Scen:rio Thr:0 /dt chm:nt 1 R:v.1,10/9/96 STCP-QA-6i2 Page 23 of 25 Rev.1 Page 15 of 17 INSTRUCTOR ACTIVITIES, ROLE PLAY, 7 AND INSTRUCTOR'S PERSONAL NOTES Event No: 6 Brief
Description:
Loss of CRD INSTRUCTOR ACTMTY: When the crew has completed actions for the accumulator fault and are prepadng to insert rod 18-35 or when directed by the lead evaluator Depress P-5, inserting malfunction IMF PM03:1P132A This willidp the operating.CRD pump on overcurrent. Afterabout Eve minutes: Depress P-6, inserting malfunction IMF RD1550191043 This willcause rod 10-43 accumulator fault. QOLE PLAY: As the NPO directed to check out the "A" CRD pump, wait two minutes then report the motor is very hot to the touch. As the NPO directed to check out the "A" CRD pump breaker, wait' two minutes then report the 50/51 device is tripped. As electrical maintenance, wait five minutes then report there appears to be a fault in the "A" CRD pump motor. O
AttichmW11 STCP-QA-612 Scenario Thre3 Rev.1 Rev.1.10/9/96 Page 14 of 17 i Page 24 of 25 SCENARIO EVENT FORM (EVALUATION) Event No: 7 Brief
Description:
ATWL ith failure of SLC to initiate P N!P POSITION TIME STUDENT ACTIVITIES PCOU Report failure of control rods to insert Arm and depress the manual scram pushbuttons Throttle feedwater flow and lower level to 40 to -161 inches. Report all rods inserted when scram fuses are pulled Report high drywell pressure PCOX Control RPV pressure between 800 and 1087 psig with SRVs until power is below the capacity of the bypass valves Manually initiate ARI Recognize failure of scram air tv. .er to depressurize. Attempt to initiate SLC _ OverTide ADS Inject SLC with RCIC ] Prevent HPCI and RCIC from injecting. l Initiate suppression chamber sprays l Initiate suppression pool cooling l
. I US Enter and direct actions per EO-100-113, Level / Power Control l Enter and direct actions per EO-100103, Primary Containment l l
Control Direct ES-150-001, Boron injection Using RCIC l Direct ES-158-001, DeenerDizing Scram Pilot Solenoids Direct ES-134-001, Restoring Drywell Cooling With A LOCA Signal Present NOTES: May classify the event as a Site Area Emergency per EP-PS-100. Emergency Director-Control Room, EAL 1 Loss Of Functions Needed To Bring The Reactor Subcritical And Loss Of Ability To Bring The Reactor To Cold _ . Shutdown.
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%icst SIMULATOR SCENARIO Scenario
Title:
Flow Comparator Failure With Failure Of APRM Upscale Trip, RCIC Pump Operability, Loss Of Feedwater Heating, Loss Of 18246, Unisolable RCIC Steam Line Break Scenario Duration: One hour ; i Scenario Number: Two
) Revision /Date: 1,10/8/96 i
l EMME!$$$1sh@!F%1 @@M&pDpi!@W %@M&P?4?QM$isfMf!@psfiger3%f glRgp Prepared By: /0/6/96 Instructor Date Reviewed By: N 3d huclear Opejg n's Training Supervisor Date Approved By: 41). /0 -tr-96
] supenising Manager /snm supenisor Date
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m Attachment 1 Scenario Two STCP-QA-612 Rev.1,10/8/96 Rev.1 Pa0e 2 of 21 Page 15 of 17 O i 1 l 1 swuwe.an a w, an + anen+mwwnwnmesma O HIS*PAGE
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Scenario Two Attachment 1 Rev.1,10/8/96 STCP-QA-612
. Page 3 of 21 Rev.1 Page 3 of 17
-l SCENARIO
SUMMARY
l , initially, the reactor is operating at 100% power. The "A" APRM is failed upscale and is bypassed. The "B" l CRD pump is tagged out of service. Event One. The PCOX will perform S0-150-002, Quarterly RCIC Flow Verification. When RCIC is operating, a small steam leak will occur in the pipe tunnel, slowly raising temperatures in the area. The PCOX will be primarily involved in this normal evolution. Event Two. The "A" flow comparator for the RBMs and APRMs fails low. The PCOU will identify the failed , comparator and recognize the failure of the APRM flow biased scram function and dispatch an NPO to l the relay room to investigate. Using the annunciator response procedure and ON-164-001, Recire Drive j Flow Instrument Failure, the PCOU will bypass the failed instrument with the joystick at the 651 panel, i and have the NPO place the operate selector switch to ZERO at the relay room cabinet. The US will l initiate maintenance activities on the failed flow unit and APRM/RPS failures and direct a manual scram be inserted on RPS A to meet the Technical Specifications impact. The PCOU and the US will be primarily involved in these two .astrument failures. Event Three. A loss of feedwater haating occurs when the extraction steam isolation valve to the 58 feedwater heater fails closed. The crew will implement ON-156-001, Unexplained Power increase and ON-144-001, Loss of Feedwater Heating. Reactor engineering will be notified to evaluate thermal limits nd the crewwill begin investigating the cause of the failure. Both PCOs and the US will be actively Snvolved in this component failure. Event Four. A power reduction to 80% poweris performed by the PCOU using recirculation flow. The new operating position on the power to flow map will be plotted. The PCOU will be primarily involved with this reactivity manipulation.
- Event Five. A loss of 18246 480 vac MCC occurs when the breaker for the inboard RCIC isolation valve cnd the supply breaker trip simultaneously. The crew will reenergize the bus and implement recovery
, actions for losses of reactor building ventilation and recirculation pump motor cooling. The US will address the Technical Specifications for the inboard RCIC isolation valve being inoperable. Both PCOs and the US will be actively involved in this component failure. Event Six. A steam line break in the common RCIC and HPCI pipe routing area occurs requiring entry into EO-100-104, Secondary Containment Control. The crew will attempt to isolate RCIC but the outboard valve will not close. The crew will manually scram the reactor as temperatures continue to rise towards maximum safe values. The crew will implement EO-100-102, RPV Control, and manually scram the reactor. Seven control rods will fail to insert requiring entry into EO-100-113, Level / Power Control. The CRD north areas and remote shutdown panel area will rise above 10 R/hr, requiring the crew to enter EO-100-112, Rapid Depressurization. The crew will rapidly depressurize the reactor. Both PCOs and the US will be actively involved in this major transient and two component failures. V
Att chm:nt 1 Scenario Two STCP-OA-612 Rev.1,10/8/96 Rev.1 Page 4 of 21 Page 15 of 1 P'
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. Page 5 of 21' ,
Rev.1 Page 5 of 17
'l SCENARIO KNOWLEDGES AND ABILITIES l General:
294001 A102 4.2/4.2 294001 A104 3.1/3.2 294001 A105 3.4/3.8 294001 A109 3.3/4.2 - 294001 A110 3.6/4.2 294001 A111 3.3/4.3 j 294001 A112 3.5/4.2 294001 A113 4.5/4.3 294001 A115 3.2/3.4 , Event 1: RCIC pump operability - 217000 A403 3.4/3.3 217000 A404 3.6/3.6 217000 A408 3.7/3.6 217000 A409 3.7/3.6 217000 G013 3.8/3.5 i i Event 2: Flow comparator A failure 215005 A205 3.5/3.6 215005 A207 3.2/3.4 215005 A405 3.4/3.4 215005 G009 3.6/3.4 215005 G012 3.7/3.6 215005 G015 4.1/4.3 pEvent 3: Loss of feedwater heating 295014 AA203 4.0/4.3 295014 AA107 4.0/4.1
~ 295014 G011 4.2/4.4 295014 G010 4.0/3.9 Event 4: Power reduction 202002 A408 3.3/3.3 202002 G013 3.6/3.4 Event 5: Loss of 18246 262001 A405 3.3/3.3 262001 G011 3.1/3.9 262001 G012 3.3/3.3 262001 G015 3.7/3.9 -
Event 6: Unisolable RCIC steam line leak 295033 EA101 3.9/4.0 295033 EA103 3.8/3.8 295033 EA105 3.9/4.0 295033 EA201 3.8/3.9 295033 EA203 3.7/4.2 295033 G011 4.2/4.3 295033 G012 3.8/4.4 d
4 Attachment 1 Scenario Tm STCP-QA-612 Rev.1,10/8/96 Rev.1 e
. Page 6 of 21 Page 15 of 17 i
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Scenario Two Attachm:nt 1 Rev.1,10/8/96 STCP-QA-612
. Page 7 of 21 Rev.1 Page 7 of 17
- 1. EO-100-102, RPV Control, revision 7
- 2. EO-100-104, Secondary Containment Control, revision 7
- 3. EO-100-112, Rapid Depressurization, revision 7 -
- 4. EO-100-113, Level / Power Control, revision 7
- 5. EP-PS-100, Emergency Director-Control Room, revision 12
; 6. ON-100 004, Reactor Power Greater Than 100% (3441 MWTH), revision 1
- 7. ON-147 001, Loss of Feedwater Heating (Extraction Steam), revision 9
- 8. ON-156-001, Unexplained Reactivity Change, revision 10
- 9. ON-164-001, Recirc Drive Flow Instrument Failure, revision 6
- 10. SO-150402, Quarterly RCIC Flow Verification, revision 16
- 11. AR-016-D04, ESS 480V LC 18240 Trouble
; 12. AR-103-C05, APRM/RBM Flow Reference Offnormal
' 13. AR-102-F01/F02 Recire NB Low Cig Flow
- 14. Technical Specifications 3.3.1,3.6.3,3.7.3 s
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- Attichment i scenario Two STCP.QA-612 Rev.1.10/8/96 . Rev.1 Pa06 8 Of 21 Page 15 of 1 4 4 4 4 E a 4 1 4 O
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Scen:rio Two Attrchment 1 Rev.1.10/8/96 STCP-QA-612 Page 9 of 21 Rev.1
-- Page 11 of 17 -l SCENARIO SPECIAL INSTRUCTIONS
- 1. Reset the simulator to IC-18 and place the bypass joystick for div.1 APRMs to "A" and place ESW and loop A suppression pool cooling in service 1
- 2. Execute preference file, YPP.SC.2, establishing the following conditions:
Malfunctions: 12:12
. RD1550064643 22 Rod 46-43 Stuck At Position 22
- RD155006460712 Rod 46-07 Stuck At Position 12
. RD1550061443 32 Rod 4643 Stuck At f usition 32
. RD1550062223 20 Rod 46-43 Stuck At Position 20
. RD1550062659 28 Rod 46-07 Stuck At Position 14
. RD1550061407 44 Rod 46-43 Stuck At Position 44
. RD1550065043 42 Rod 46-43 Stuck At Position 42
. NM178007A 125 APRM A Failed Upscale
. MV09:HV149F008 90 RCIC Outboard Steam Isolation Fails To Close I
. RLO2:C721K12C RPS NMS Scram Trip Relay Fails Energized
. RLO2:C721K12E RPS NMS Scram Trip Relay Fails Energized
. RLO2:C721K12G RPS NMS Scram Trip Relay Fails Energized Remote Functions: 1
. MRF PM131P1328 OUT CRD Pump B Breaker Out Pushbutions Assianments
- 1. IMF NM178012A 0 Flow Comparator A Fails Downscale
- 2. MRF NM1780006 ZERO Flow Comparator A Mode Switch To ZERO G. IMF RC150005 0.510:00 RCIC Steam Line Break in Pipe Area,0.5%,0ver 10 Minutes
- 4. IMF MV05:HV102428 5B Heater Extraction Valve Spurious Closure
- 5. MRF DB105118 OPEN 18246 Supply Breaker Trips Open
- 6. MRF DB106383 OPEN 18246-22 RCIC isolation Valve Breaker Open
- 7. IMF RR17900315:00 Fuel Failure,1% Over 10 Minutes
- 8. MMF RC150005 80 0:30 RCIC Steam Line Break in Pipe Area,80% Over 30 Secs
- 9. MMF RR179003 70 Fuel failure severity increases to 70%
10.lMF TR02:RIT137501510:00 CRD North Area Rad,15 R/Hr Over 6 Minutes 11.lMF TR02:RIT137531811:00 RSD Panel Area Rad,18 R/Hr Over 5 Minutes
- 5. Prepare a tumover sheet indicating the "A" APRM is failed upscale and is bypassed, I&C is troubleshooting it now and that the 'B' CRD pump motor is being replaced, expected to be retumed in approximately 30 hours. Suppression pool cooling is in service to ' support the p RCIC flow verification per SO-150-002 (with a cold quick start and taking ISI data) that will be Q done as soon as possible as it is about to exceed its grace period.
- 6. Place a protective blocking tag on CRD pump B.
Atttchm:nt 1 Scenario Two STCP-QA-612 RIv.1,10/8/96 Rev.1 Page 10 of 21 Page 15 of 17 I 1 i 4
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RX MANIP TIME DESCRIPTION N Commence RCIC SO 150-002, Quarterly RCIC Flow Verification
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l/l Flow comparator A failure /Fallure of RPS *A* APRMs to generate half-scram C/R Loss of feedwater heating /20% power reduction with recirculation flow C Loss of 18246/ loss of power to inboard RCIC isolation valve MT/C RCIC steam leak / outboard valve fails to isolate C Manual scram with failure of seven rods to insert l l f 1
Att:chment 1 Scenario Two STCP-QA-612 Rev.1,10/8/96 Rev.1 P, age 12 of 21 Page 14 o SCENARIO EVENT FORM (EVALUATION) Event No:1 Brief
Description:
RCIC quarteriy pump operability taking ISI data P M'D TIME STUDENT ACTIVITIES ' ' ~ POSITION Perform RCIC pump oper.diitty per S0-150-002, Quarterly RCIC PCOX Flow Verification Station NPO at 1D254 -. . Notify US that RCIC is inoperable while breakers are open Direct throttling of F022 to 40% Direct opening 10254 breakers 22 and 51 ._ j Start GETARS Opens HPCI F011 and manually initiates RCIC - 5 sec. later
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Resets inite signal Directs closing breakers at 10254 Adjusts RCIC flow parameters by throttling F022 ' Coordinate local activities with NPO, HP, and maintenance
.,nfirms Reactor Building Hi Rad alarm due to RCIC surveillance PCOU US Monitors surveillance activity Evaluates Technical Specifications for RCIC inoperable .
NOTES: w 5 i
Scen:rio Two Attachment 1 Rev.1,10/8/96 STCP-QA-612 . Page 13 of 21 Rev.1 Page 15 of 17 INSTRUCTOR ACTMTIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No:1 Brief
Description:
RCIC quarterfy pump operability INSTRUCTOR ACTMTY: When RCICis operating: Depress P-3, inserting malfunction RC150005 0.510:00 This creates a small steam leak in the pipe routing area. It will not be detected but will start raising pipe area temperatures to support the unisolable leak later. Position RCIC F022 to 40% with GCF as directed (then remove the GCF after it is at 40%) on RC3 i Open breakers as directed on DC-1. A Reset GETARS on RP13 then unreset it. ROLE PLAY: As NPO directed to throttle F022 to 40%, wait one minute then report it is at 40%. As NPO at 1D254 directed to open breakers 22 and 51, wait one minute then report them open. When directed to close them later, wait one minute and then report them closed. As maintenance when directed to obtain ISI data, delay reporting back that ISl' data is
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Att chm:nt 1 Scenario Two STCP-Q A-612 Rev.1,10/8/96 ' Rev.1 ' Page 14 of 21 Page 14 of 17 SCENARIO EVENT FORM (EVALUATION) Event No: 2 Brief
Description:
Flow comparator A fails downscale and RPS A falls to trip P N/P TIME STUDENT ACTIVITIES l POSITION Acknowicdges AR-103-C05, APRM/RBM Flow Ref Off-Normal and PC6U recognizes the half-scram condition. Diagnoses failure of flow compara- A. Bypasses flow comparator A per ON-164-001, Recirc Drive Flow instrument Failure Directs NPO to place the flow comparator function switch to ZERO. Inserts 1/2 scram using manual scram pushbuttons I l Directs activities per ON-164-001, Recirc Drive Flow instrument l US Failure Contacts I&C to investigate the instrument failure Determines Technical Specifications 3.3.1 l _l Trip div. A RPS in 12 hours (action a)
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Ensure RPS trip function met in one hour (action c) _ PCOX Contact PCC about half-scram condition NOTES: l _ l I o
Scenirlo Two Attachment 1 Rev.1,10/8/96 STCP-Q A-612
- Page 15 of 21 Rev.1 . Page 15 of 17 J -
INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES s Event No: 2 Brief DescripHon: Flow comparator A fails downscale and RPS A fails to trip INSTRUCTOR ACTMTY: ARet the PCOX has, RCIC operating or when directed by the lead evaluator Depress P-1, inserting malfunction NM178012A 0 This fails the A tiow comparator downscale. When directed to place the function switch for the flow comparator to zero, wait two minutes then: : Depress P-2, modifying remote function NM178006 ZERO l This places the function switch for flow comparator A to ZERO.
} ROLE PLAY:
When directed as the NPO to place the flow comparator to ZERO, wait two minutes then report back that the switch is in ZERO. When directed as I&C to investigate the failure of the flow comparator, wait five minutes then report that K12C, E, and G failed to deenergize and it will take about an hour to determine why this occurred.
. . 1 4
l
Scenario Two Att: chm;nt 1 ! Rev.1,10/8/96 STCP-QA-612 Page 16 of 21 Rev.1 . SCENARIO EVENT FORM (EVALUATION) Event No: 3 and 4 Brief
Description:
Loss of feedwater heating and power reduction with recirculation flow POSITION TIME STUDENT ACTIVITIES P N/P PCOU Recognize increasing reactor power above 100% and reduces power to 100% with recirculation flow per ON-156-001, Unexplained Reactivity Change Reduces reactor power by 20% per ON-147-001, Loss of Feedwater Heating (Extraction Steam) Plots position on power to flow map. Initiates GETARS PCOX Diagnoses problem as the 50 heater extraction steam valve closing Monitors steam line and offgas radiation levels Directs NPO to heater panel to determine cause of isolation Notifies PCC of power change isolates heater string l US Directs actions per ON-156-001, Unexplained Reactivity Change and/or ON-147-001, Loss of Feedwater Heating (Extracilon Steam) (may also enter ON-100-004, Reactor Power Greater Than 100% (3441 MWTH) Contactr Electrical Maintenance i l Notifies Duty Manager of problem ' Notifies HP and Chemistry of power change Contact Reactor Engineering Directs heater string isolation NOTES: l The crew may diagnose the loss of feedwater heating and not enter ON-156-001. t
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Attachment 1 Scenario Two R3v.1,10/8/96 STCP-QA-612 Rev.1
-Page 17 of 21 Page 15 of 17 INSTRUCTOR ACTIVITIES, ROLE PLAY, 7 AND INSTRUCTOR'S PERSONAL NOTES Event No: 3 and 4 Brief
Description:
Loss of feedwater heating and power reduction with recirculation flow INSTRUCTOR ACTMTY: After the crew has completed actions for the flow comparator failurenailure of RPS to actuate or when directed by thelead evaluator: Depress P-4, IMF MV05:HV102428 This causes the extraction steam valve to the 5B heater to spudously close. ROLE PLAY: As the NPC sent to the 1C102 panel, wait two minutes then report that you cannot determine l any cause for the extraction steam valve closing. Call up LP1C10101 and report on any annunciators alarming also. As electrical maintenance, wait five minutes and report it will take about four hours to determine why the valve closed. , l o O O
Attichm:nt 1 Scenario Two STCP-Q A-612 Rev,1,10/8/96 Rev.1 . Page 18 of 21 Page 14 of 17 SCENARIO EVENT FORM (EVALUATION) Event No: 6 Brief
Description:
Loss of 1B246 P N/P TIME STUDENT ACTIVITIES , POSITION PCOU Monitor plant conditions 1 Acknowledge AR-102-F1 & F4, Recirc Low Cig Flow annunciator and c monitors recirculation pump windin0 temperatures Determine drywell temperature / pressure rise is caused by a loss of driac!! coolers i Acknowledge annunciators AR-016-D04, ESS 480V LC 18240 W PCOX Trouble Diagnose problem as a 18246 problem. Dispatch NPO to investigate
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Report problem is 18246 supply breaker and breaker to RCIC F007 are tripped. Direct NPO to reclose supply breaker Report rising drrwc!! temperature Dirr.TNPO to reclose breaker Dire 5 NPO to commence reloading bus, beginning with drywell coolers Report problem with the RCIC inboard isolation valve breaker Dirret NPO to 1C275/276 to restore Reactor Building ventilation Restore cooling water to recirculation pumps US Contact electrical maintenance Determine Technical bgcifications actions for inboard RCIC isoletion breaker problem 3.7.3 and 3.6.3 Close RCic outboard valve within four hours and 14 day LC ' 1 RCIC ,_ Direct recovery from bus loss Notify the duty manager NOTES: - , w l
Attichm:nt 1 Scenario Two Rev.1,10/8/96 STCP-QA-612 Rev.1
-- Page 19 of 21 Page 15 of 17 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 5 Brief
Description:
Loss of1B246 INSTRUCTOR ACTMTY: When the crew has completed actions for the loss of feedwater heating or when directed by the lead evaluator: Depress P-5, modifying remote function DB105118 OPEN AND Depress P 6, modifying remote function DB106383 OPEN This opens the supply breaker to 18246 and breaker 22 (RCIC valve) on 18246. When directed to'close the 18246 supply, then: modify the remote function DB105118 to CLOSE the breaker. (J Once recovery actions for reenergizing 18246 are in progress or when directed to by the cvaluator-Depress P-7, inserting malfunction RR17900315:00 . This causes a single pin fuel failure to occur over Fuel Failure 5 Minutes ROLE PLAY: , As the NPO dispatched to 18240, wait one minute, then report the supply breaker to the 1E'246 bus is tripped open. When directed to investigate the 1B246 problem, report breaker 22, the supply to RCIC F007 is tripped open. As electrical maintenance, wait three minutes, then report there are no obvious problems. Recommend attempting to reclose the supply breaker and recommend leaving the 22 breaker as is. When asked about the RCIC F007 breaker, report is will take a while to determine the cause of the problem. . d
Attrchm:nf 1 Scenario Two S' .' P-QA-612 Rev.1,10/8/96 Rev.1 Page 20 of 21 Page 14 of 17 SCENARIO EVENT FORM (EVALUATION) Event No: 6 Bdef
Description:
Unisolable RCic steam line break in pipe area P N/P TIME STUDENT ACTIVITIES POSITION PCOU Recognize elevated radiation levels in reactor building Manually scrams the reactor. Prevents injection of condensate during rapid depressurization Reports seven rods fail to insert. Attempts to manually insert control rods Acknowledge steam leak detection annunciators for RCIC and HPCI PCOX Confirm temperatures are rising in RCIC and HPCI piping area and timers have started Start ESW and ESS room coolers itspidly depressurtze the reador by opening the ADS SRVs. Prevents injection of RHR and CS during rapid depressurization Maximizes CRD Injects SLC for level control Enter ano direct actions of EO-100-104, Secondary Containment US Control Directs starting ESW and room coolers , Directs manual scram of reactor on approaching maximum safe temperature Enters and directs EO-100-113, Level / Power Control Enter and direct actions of EO-100-112, Rapid Depressurization Directs preventing injection of low pressure systems Directs opening ADS SRVs NOTES:
'May classify the event as an Site Area Emergency per EP-PS-100 Emergency Directoi-Control R i Steam Line Break Occurs Outside Containment Without isolation.
~l
* *^ .,
Scenario Two Attichmsnt 1 Rev.1,10/8/96 STCP-QA-612
+ Page 21 of 21
- Rev.1 Page 15 of 17 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 6 Brief DescriplJon: Unisolable RCIC steam line break in pipe area INSTRUCTOR ACTIVITY:
When the crew has completed correcting the 18246 problem or when directed by the lead evaluator: Depress P-8, modifying malfunction RC150005 80 3:00 This increases the seventy of the RCIC steam line break to 80%. When the crew makes the decision to perform scram imminent actions: Depress P-9, modifying malfunction RR179003 70 This increases the severity of the fuel failure to 25% over 3 minutes. C\ Qfter the crew attempts to manually isolate RCIC: o-w4 se.w Depress P-10, inserting malfunction TR02:RIT137501510:00 When directedby the lead evaluator-Depress P-11, inserting malfunction TR02:RIT137531811:00 These will camp CRD area north and Remote Shutdown Panel area radiation to' greater than 10 R/hr requiring rapid depressurization. ROLE PLAY: As the NPO directed to investigate the steam leak, wait two minutes then report there is a loud roar in the pipe area and it cannot be entered. As HP with the NPO, report radiation levels are increasing in the pipe area, preventing entry. TERMINATION CUE: p The crew has rapidly depressurized the reactor and level and pressure are stabilized. V
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e y s Attachm:nt 1 Scenario Thr;e STCP-QA-612 Rev.1,10/9/96 Rev.1 Pace 25 of 25 Page 15 of 17 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES I Event No: 7 ' Brief
Description:
ATWS with failure of SLC to initiate INSTRUCTOR ACTIVITY: When the crew stabilizes RPV level below -60 inches or when directed by the lead evaluator Depress P-7, inserting malfunction RR164010 5 5:00 This causes a 5% break of the bottom head drain to occur over 5 minutes. When drywellpressure is between 1.72 psig and 4 psig AND the crew has discussed the impact of the drywellpressure on the plant or as directed by the lead evaluator: 1 i Vent the scram air header. OR Pull the RPS fuses.
. Whichever action the crew initiated first.
ROLE PLAY: As the NPO directed to perform ES-150-001, Alternate Boron injection, wait 20 minutes then report the hoses are connected and ready to open the valve. As the AUS directed to perform ES-158-001, Pulling RPS Fuses, wait until drywell pressure is greater than 1.72 psig, but before it is 4 psig and the crew has discussed the irspact of the drywell pressure on the plant, then report you are ready to pull the fuses. As the NPO directed to vent the scram air header, wait until drywell pressure is greater than 1.72 psig, but before it is 4 psig and the crew has discussed the impact of the drywell pressure on the plant, then report you are ready to vent the air header. TERMINATION CUE: RPV level and pressure are stable, all control rods are inserted and the crew is addressing containment parameters. V
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fFR-17-1997 .20815' $NRC REG 1 KOP PA. r . uc ane .
' uNrrED STATES' l NUCLsAR RsGut.ATrHtY COMMISSION : . v C.-
neoCNi i O suoorenu#NuyiNA tems m
..... ss .
December 2, 1996 Frank J. Calabrese, Jr. I l
- ses;s. Kennedy Drive !
i i ~McAdoo,PA 18237 . l
Dear Mr. Calabrese:
! This is to inform you that, on the beels of a Orsding of the written examination , taken on October 21,1996, and the operating test taken on October 22-24,1996, ' ! In connecten with your appilcation for a senior reactor operator liconse for the Susquehanna Steam Electric Generating Station,, Unit Nos. I and 2, whloh indicates i that you did not pass that examination, it la proposed that your application be i denied. Enolosed is e copy of the examination results, Indicating those arose in which you exhibited deficiencies. A copy of the master answer loey is also provid- , ! d- l ; i j if you accept the proposed denial and decline to request alther ari informal NRC staff review or a hearing within 20 days as dieouooed below, thie' proposed denial : will become a final denial. You may then reapply for a license in accordance with , 10 CFR 55.35, subject to the following conditions: ! l I i
- s. Because you did not pase either the written examinetion or the j i operating test, administered on October 21-24,1094, you will bo )
4 required to retake the written examination and the operating test. J i i . ! . b. You may reapply for a license two months from the date of this letter. !. A reexamination will be scheduled, upon request by 'your facility man-agement, shortly after your application is received. L If you do not accept the proposed denial, you may, within 20 days of the date of a this letter, take one of the following actions: !
- You may request an informal NRC staff review of tho grading of your i
examination. Your written request must be sent to the Director, Divi-sion of Reactor Controls and Human Factors, Office o' f Nuclear ~ Roootor Rogulation, U. S. Nuoleer Regulatory Commiselon, Washington, D.C. 20555. Your request must identih the portions of your examination that you believe were graded incorrectly or too
- esverely. In addition, you must provide the basis, iricluding I CERTWiED MAIL- ;
RETURN RECEIPT REQUESTED : i i
- eummam > ,c ..m - , - -
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APR-17-1997 10:15 94tC REG 1 KOP PA. 610 337 53ee i . ca t ; l F. Calabrese, Jr. . supporting documentation, such as procedures, instr'uctions, computer 3 printouts, and chart traces, in as much detail as pos6ible, to support your contention that certain of your responses were' graded incorrectly or too severely. The NRC will review your contentio'ns, reconsider your grading, and inform you of the results, if the proposed denial is sustained, you will have the opportunity to request a hearing pursuant to 10 CFR 2.103(b)(2) at that time.
- You may request a hearing pursuant to 10 CFR 2.103(b)(2). Submit i your request, in writing, to the Secretary of the Conimission, U. S. l Nuclear Requlatory Commission, Washington, D. C. !20555, with a copy to the Assistant General Counsel for Hearings,; Office of the !
l General Counsel, et the some address. ! l t You may not reapply for a license, pursuant to 10 CPR 55.35, until your license j has been finally denied. Failure on your part to exercise one of thlese options within j 20 days constitutes a waiver of your opportunity for informal revlew and your right i to demand a hearing and, for the purpose of reapplioation under 10 CFR 65.35, ! j renders this letter a notice of final denial of your application, effective as of the , date of this letter. l 2 If you have any questions, please contact me at 610-337-5211. f i ncerely,
- p ';. m
- ni/
Glenn W. Meyer, Chief Operator Ucensing and
- Human Performance Branch Division of Reactor Safety Docket No. 55 01425 1 i
Enclosures:
As Stated - 3
; cc W/o enclosures: '
G. J. Kuerynski, Plant Manager W. H. Lowthert, Manager - Nuclear Training i
)
TOTAL F.03 w--,
'yd.. . Id/ % i
; ::: 9 :sz:y 7
December 19,1996
?
Frank J. Calabrese Jr. a- 698 S. Kennedy Drive ; McAdoo,PA 18237-1731
/
Director Division of Reactor Controls
- and Human Factors Office of Nuclear Reactor Regulation i U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Director:
l This letter is to request an informal NRC staff review of the grading of my written SRO examination which was administered on October 21-24,1996, as 1 do not accept the proposed denial. I believe the written examination and the simulator examination have been graded incorrectly or too severely. Please see the attached justification for the written examination and the simulator examination topics, for which I request reconsideration. Enclosed is my simulator examination basis and the basis for Questions 13,22,64, and . 66. For your review purposes, I have also enclosed the simulator examination comments from my proposed denial letter of December 2,1996. Thank you for your kind consideration in this matter which is of utmost importance. If you have any further questions, comments, or concerns, please contact me at the above address or via phone at (717) 929-1577. Sincerely,
-0. 0-Frank J. Calabrese Jr.
Ifc _) . i.
l (n * ' J g i f. 4 1 1 i e
\
l 1 l i SIMULATOR EXAMINATION BASIS 4 i L i i
- a l
i a l Pagel of 2
\
. - . ..- _. . -. - -.- . -.- . . - _ . - . . -~ ..
j- , l
; Comment C,4,A -
Durir.g this time in the scenario, I had directed the PCOX to open 4 (6) ADS valves and override Lo Press ECCS at =900# The PCOX l then began to open (6) ADS valves and override Lo Press ECCS as directed. I then reordered override of all Lo Press ECCS again i at 350# when I noticed the "A" RHR Pp running. (Note: 350# is well above the pressure at which RHR would inlect to the vessel.) I saw D2 injection, no level increase, and no power increase as a result of these actions. I Comment C,4.C - . (Same Rebuttal as Comment C.4.A) ! 1 l 1
- Comment C.7.B -
During the scenario in question, I never assumed primary l e responsibility for monitoring secondary containment rad levels. ) l When initial rad levels were observed, I directed the PCOU to i l- place rad format 1/S and maintain primary responsibility for j l monitoring. I advised I would try to "back him up" with the CRT ' at my console. I was advised by the PCOU when we had exceeded -
- max safe rad in (2) areas. I then confirmed his indications and 3 proceeded to order capid depressurization of the plant as directed in EO-100-104.
Conclusion - Therefore, I believe the grading of my simulator performance is not l only too severe, but somewhat incorrect. The scenario involved
- several major failures of plant equipment which were dealt with in 4
accordance with plant procedures, and no incidents of public safety concem were raised. I respectfully request that the simulator test be reconsidered. i i i i i i i
'~
Page 2 of 2 r . . ., . . -
~
January 16,1997 (pj . . Memo To: Stuart Richards, Chief, OLB, NRR From: Glenn Meyer, Chief, OL&HPB, Region I p(
SUBJECT:
SUSOUEHANNA EXAM APPEALS ; in letters dated December 19,1996, two SRO applicants at Susquehanna appealed their ) license denials. Mr. Calabrese appealed his written and operating failures, and Mr. ; Robinson appealed his written failure. Based on Region l's review, we have concluded that l both written exam failures should be overturned to passes. However, we do not believe a l sufficient basis has been provided to change the operating failure. We will await your l concurrence prior to issuing Mr. Robinson a license. The specifics are addressed below. Written Examination: SRO 13: Delete - Both applicants contend that the correct answer is dependent on the j purpose of the tag and the personnel involved. Based on the supporting l documentation and discussions with f acility personnel, this contention is j true. Therefore, answers 'b', 'c', or 'd' could be considered correct. The question should be deleted from the examination, because it'has three correct answers, j i SRO 16: Delete - Mr. Robinson contends that the correct answer is answer 'c'. However, we believe the question should be deleted based on three factors. G
- Procedural guidance on access above the 767 elevation during peripheral bundle movement is confusing. Although we do not agree with Mr. Robinson's contention that access is permitted, one procedure addressing this states that "no access above the plane of 767 elevations" is permitted under the specified conditions.
Accordingly, 767 could be cor sidered a correct response to the quesOn stem of "What is the maximum elevstion that the operator can go to in the containment?"
- The core location designated in the question stem is not a valid location; locations 23-02 and 23-04 (the locations on either side of 23-03) are peripheral bundles, but 23-03 is not valid.
- Additionally, the learning objective referenced for this question required the applicant to discuss the responsibilities of shift supervision in regard to personnel radiation protection in accordance with health physics procedures. Health physics is responsible for ensuring that access controls to radiologically controlled areas are c:;tablished and implemented. There is no need for an SRO applicant to know the responsibilities of health physics from memory.
Additionally, it is not reasonable to expect a applicant to know the g difference in the requirements based on bundle location from memory. V SRO 22: No change - Both applicants contend that the question is inappropriate for a t
1
. clossd reference exan.ination. We do not agree. The question is based on a facility learning objective which states " state the reactor recirculation pump restart limitations." The question was developed by modifying a question
' , from the facility initial licensed operator exam bank. The question fits the learning objective and was developed and approved by the facility. No _ change should be made to the examination grading. SRO 64: No change - Both applicants contend that a second answer should be accepted. However, this' position was presented by the facility in post-exam com onts and rejected. The applicants do not provide any new information l-on the question. Exam Report 96-12 stated:
" Comment not accepted. The breaker alignments per EO-100-030 and EO-200-030 and the sequencing specified in EO-000-031 allow the staff to anticipate and monitor automatic equipment starts in a controlled manner.
However, the specific sequence in the restoration procedure (EO-OOO-031) does not prevent the automatic stari of equipment without operator action as specified in answer 'd', the additional correct response proposed by the facility. Automatic starts of equipment will occur if initiation signals are present, but energizing the busses in the specified sequence and waiting between each bus ensures that an undervoltage condition doesn't occur when the equipment starts. No change was made to the answer key." SRO 66: No change - Both applicants contend that a second answer should be accepted. However, this position was presented by the facility in post-exam comments and rejected. The applicants do not provide any new information on the question. Exam Report 96-12 stated:
" Comment not accepted. The question asks for the status of the SRVs and l SRV instrumentation based on the information provided. During a station blackout, tailpipe temperature indication would be available and the acoustic monitors would be unavailable due to loss of power. Based on the information provided, the current status of the SRVs (leaking, closed or open) cannot be confirmed. Answer 'b' is not incorrect with respect to the status of SRVs, but it is incorrect with respect to the status of instrumentation because the tailpipe temperature indications do not fail high during a station blackout. Answer 'c'is not incorrect with respect to SRV status, but is incorrect with respect to the status of SRV instrumentation because the SRV leaking annunciator is not due to loss of power. Answer
'd' is not incorrect'with respect to the status of SRVs, but is incorrect with respect to the status of SRV instrumentation because the acoustic monitors are deenergized not cleared. Answer 'a' is not incorrect with respect to the status of SRVs and is correct with respect to the status of the SRV instrumentation. Even though SRV status cannot be confirmed and none of the ansviers are incorrect with respect to SRV status, the question is still valid with respect to the status of SRV instrumentation. The applicants
, would be able to determine that answers 'b', 'c', and 'd' are incorrect based (A'") on the status of SRV instrumentation. Answer 'a' would be the only correct answer. No change was made to the answer key."
REVISED GRADING: Based on the above positions, Mr. Robinson has 72 of 90 for 80.0%, and Mr. Calabrese has 72 of 90 for 80.0%. We have previously addressed exam validity with George Usova due to deleting questions, and we continue to believe that the exam is valid as no Ks or As were unaddressad (questions deleted above are admin). ' OPERATING TEST: We do not support any change in grading based on the following. C.4.a & The applicant contends that he directed the PCOX (BOP) to open 6 ADS C.4.b: valves and override low pressure ECCS when reactor pressure was 900 psig. The examiners' notes and recollections of the scenario do not support the applicant's contention. The applicant also contends that he saw no injection. However, the PCOX reported that there was some injection into the RPV from RHR A. The applicant's contentions that there was no observable level or power increase are consistent with the examiners' observations. Regardless of what reactor pressure was when the applicant ordered the PCOX to overrido low pressure ECCS, he did not wait until all RPV injection l ) was stopped and prevented prior to directing the PCOX to open the ADS (f valves as required by step RD-5 of EO-100-112, " Rapid Depressurization." He never ordered the PCOU (RO) tu stop and prevent injection from condensate and he did not give the order to override low pressure ECCS until af ter he had ordered the PCOX to open the ADS valves. It is necessary to prevent injection of low pressure ECCS before the RPV is depressurized because the method used to prevent injection will cause an injection if RPV prest.ure is below the shutoff head of :he pump when it is performed. The applicant also did not refer to EO 100-112 until after he had directed actions spoeified in the procedure. Even though there was no observable level or power increase during this ! scenario, the applicant's failure to refer to and correctly implemont EO-100-112 degraded the plant unnecessarily. In this scenario, seven rods were not inserted. The improper injection of cold water could have caused a reactor power excursion and substantial core damage. The applicant failed to refer i to the procedure in an important instance and made a significant error that i degraded the plant unnecessarily. No change should be made to the examination grading, as we believe his significant errors of failing to refer to the applicable EOP and improperly implementing this major evolution represent a sufficient basis for f ailure. 1 C.7 b: The applicant contends that he did not assume primary responsibility for
,m, monitoring secondary containment radiation levels. His contention is mpot; l } while we continue to believe that the candidate "had assumed responsibility
'V for monitoring secondary radiation levels," as the SRO, he was responsible
i . I I for monitoring these parametersseither personally or via his crew. He should , have ensured that radiation levels were monitored closely by whomever. ' Neither the applicant nor the rest of the crew recognized that two areas had exceeded maximum safe radiation levels (the criterion to begin a rapid , depressurization) for approximately five minutes. The applicant failed to . ' provide timely direction to rapidly depressurize the RPV and failed to provide well thought out direction'to stop and prevent injection as discussed in C.4.a i and b. ~..sefore, no change should be made to the examination grading. Note: The applicant's grade in this rating factor did not contribute to his failure of the examination. l I l l I i l 7^ . . k 4
j i t I i
- t i' January _30, 1997 l I
L MEMORANDUM TO:: Thomas A. Peebles, DRS/Rll Paul M. Steiner, DRS/Ril . L Elden A. Plettner, DRS/ Rill f . t FROM: Stuart A. Richards, Chief- )_ Operator Licensing Branen Original signed by: i h Division of Reactor Controls and Human Factors *
' Office of Nuclear Reactor Regulation
SUBJECT:
SUSOUEHANNA APPEAL PANEL ! This memorandum is to inform you of your selection as member of the appeal panel
- , to_ review the licensing examination failures and application denials of Messrs. Gordon E. Robinson and Frank J. Calabrese Jr. The chairman of the panel will be Thomas Peebles. The panel will convene as soon as possible. The panel is j tasked to review the items in contention on the written examination and operating test. !
!- The panel should acquire or develop any and all documentation to support their findings l and is expected to communicate with the examiner of record to discuss the preliminary l results of the panel review. ! Your recommendation containing the appropriate documentation to sustain or overturn the ! I.. denial of the license should be issued by February 12,1997. l i 4 DISTRIBUTION: ) HOLB R/F q ~ JLennartz, Rill { n n. . , .e m - me m m. n.= e . c.,, .n nu.a.i : . c.,, we on n ne.acw. r No copy j OFFICE. HOLB/DRC_H - lC HOLB/DRCH dL. NAME- JMunto 6F Ng*' j '- DATE 018./97 gpvc_-- 01/3./97 l OFFICIAL RECORD COPY l rV
, UNITED STATES
*Ctig'o NUCLEAR REGULATCRY COMMISSION
,4 p' REGloN 11 101 MARIETTA STREET, N.W., sulTE 2900 o o j ATLANTA, GEORGIA 3032:00190
- February 14, 1997
,,,g MEMORANDUM TO: Stuart A. Richards j FROM: Thomas A. Peebles, DRS, Rllf' Elden Plettner, DRS, Rlli y e gf Paul Steiner, DRS Ril CJ fC._
SUBJECT:
APPEAL PANEL INFORMAL REVIEW FOR SUSQUEHANNA Per your request, we have conducted a review of Mr. Calabrese's and Mr. Robinson's concerns as stated in their requests for an informal review. The appeal panel communicated via e-mail and telephone. We have attached the analysis and conclusions for your review. The panel has reviewed: the material submitted by the candidates; the examination of record; ) and discussed the examination with the exam;ner of record and a representative of the j facility. ! l a) The written examination was regraded with both candidates passing. Mr. Calabrese's score was raised to 87% with 76 correct out of 87 total questions.
,3 Mr. Robinson's score was raised to 87% with 76 correct out of 87 total i questions (x_
b) The operating test was regraded and the panel believes that Mr. Calabrese's 1 performance during the simulator scenario did not constitute an adequate basis l for approval of his application for a Senior Operator License. Additionally, the panel suggests that the entire written examination be reviewed to assure it's validity, conuidering the number of comments a:cepted and the number of qtestions deleted. If you have any questions, please contact me at (404) 331-5541.
Attachment:
As stated l 1 c)
., e
i i i p, SUSQUEHANNA APPEAL PANEL
\vl ,
i Written Exam Review j Mr. Calabrese and Mr. Robinson both appealed SRO questions 13,22,64, and 66. Mr. Robinson also appealed question 16. The two candidates submitted identical substantiating ! information otherwise. The appeal panel used this information, along with information l supplied by the examiners, the Region I review, and additional information provided by the ; . - bcility staff, to draw their conclusions. , j
. Question 13: Panel recommendation: Delete due to multiple correct answers. i As administered l A va!ve is tagged with a pink lag during an outage. Repositioning / operation of the valve can be approved by which one of the following individuals or combinations of individuals?
- a. Only the work group
- b. Only Shift Supervision
- c. Shift Supervision and the Operations Outage Supervisor.
- d. The work group and Shift Supervision. l Answer Key choice: d One candidate requested that answer b be considered the correct answer and the other l requested that answer c be considered correct.
The panel recommends: Answers b. c or d could be correct. The controlling procedure, Procedure NDAP-0A-0302 Rev. 6 " System Status and Equipment Control." in section 4.6 under the duties of the Work Group / Worker i the following statement exists: i Section 4.6.3 states "When authorized by ODerations Shift or . Outaae Group Supervision and the individual or work aroup reau1rina the Status Control Tao, a worker may manipulate components. This can include the operation of status control tagged (pink tag) components when required for venting and ' draining systems." The procedure allows that 'd' is correct. Additionally, facility personnel state that the Operations Outage Supervisor can be and often is the work oroup during outage conditions, making "c" correct. As 'd' and 'c' are correct, it can be argued that 'b' is also correct, in that Shift Supervision can encompass Outage Group supervision. Question 16: Panel recomn1endation: Delete as the question is invalid. i f
- f. .
2 . As administered i The bundle from location 23-03 is being transferred from the core during core off- l load. A leak has occurred requiring an operator to enter containment to investigate. ! What is the maximum elevation that the op?rator can go to in the containment?
- a. 738'
- b. 752'
- c. 767' !
- d. 779' Answer Key choice: b The candidate requested that answer c be considered the correct answer.
The panel recommends: There is no requirement for an SRO applicant to be expected to know the difference in procedural requirements based on fuel bundle location. If the intert was to have l the question address procedural requirements for a peripheral bundle during this j situation, a nonplausible core location causes the question to be invalid. There is no p fuel bundle location 23-03 in the core. Therefore, delete the question as it is invalid. I \}' Question 22: Panel recommendation: Delete, invalid question based on inadequate
\
procedure. ' As administered j i Given the following conditions:
- A reactor cooldown is in progress.
- Recirculation pump 1A was secured at 0815 due to concems with seal leakage.
- At 0930, Recirculation pump 1B was inadvertently tripped.
- At 0945 the 1B pump is restarted.
- The 1B pump is tripped again at 0950.
What is the earliest time the 1B pump is allowed to be started? a.1000 b.1005 c.1030 d.1035 Answer Key choice: d O , , Q) t The candidates requested that answer c be considered the correct answer.
3 .
- l (V The panel recommends:
The question is based on a note in OP-164-001, Rev 26. Procedure OP-164-001 Rev. 26 in section 3.3.25:
" NOTE: With motor windings at ambient temperature (<-104 F), motor may be i started and brought to speed two times in succession. With motor !
i windings at rated temperature &104 F), motor may.be started and l brought to speed once After all permissible starts have been l made, windings must return to rated or ambient temperature before j further starting attempts may be made. Motor windings can be : assumed to Mve returned to rated temperature after 45 minutes -l shutdown or after 15 minutes running at rated speed. " (Italics and bold added to word 'or') i The note is not logical, and the guidance it provides to an operator is not usable. The facility agrees the note is incorrect and will change the procedure. The licensee lesson plan used , the same NOTE. According to the NOTE. the temperatures are defined as; ambient (<=104 O and rated (>104 F).. Therefore, as the motor winding temperature is always either <=104 ! F or >104 F, the pump can be started any time, no matter how many starts have been made. l This is technically incorrect. l Question 64: Panel recommendation: Delete as 'there are multiple correct answers. As administered Station Power Restoration, EO-000-031, provides a specific sequence for reenergizing busses from an off-site source to AVOID:
- a. diesel generators tripping on overspeed when loads are transferred to off-site power.
- b. underfrequency condition on off-site sources due to manually reenergizing non-emergency busses. '
- c. undervoltage condition caused when a ECCS initiation signalis present.
- d. starting equipment automatically without operator action.
Answer Key choice: c The candidates requested that answer d be considered the correct answer. 4 The panel recommends: 1 The stem of the question asks why a post-blackout procedure, "provides a specific sequence
- for reenergizing busses from an off site source to AVOID:". After review of the procedure, it could be contended that it is designed to avoid the undesirable situations provided in '
answers "a", "c" and "d*. Also, ans,wers "c" and "d" are arguably the same. E I
gm , 4 . ,
,/ Question 66: Panel recommendation: Grace remains as original.
As administered. Given the following:
- A station blackout has occurred.
- MAIN STEAM SRV LEAKING is alarming.
- MAIN STEAM DIV 1 SRV OPENis clear.
- MAIN STEAM DIV 2 SRV OPENis clear.
Based on this information, what is the status of SRVs and equipment to monitor SRVs?
- a. An SRVis leaking. The acoustic monitors fait during a station blackout.
- b. All SRVs are closed. Tailpipe temperature indications fail high during a station blackout
- c. Status of the SRVsis unknown because the annunciators are indications of loss of power to instrumentation.
- d. An SRVhas opened, then reclosed, caused the acoustic monitors to clear.
Answer Key choice: a ( i () The candidates requested that answer b be considered the correct answer. The panel recommends: The candidates contend that the status of the SRVs is unknown based on the information provided in the stem, and therefore, no answer exists. The first part of this statement is true, the second is not. It is true that the status of the SRVs is not determinable from the information provided in the stem. However, that is not the intent of the question. The question is designed to probe the candidatos knowledge of power supplies during a blackout. Only distractor "a"is plausible, regardless of the status of the SRVs. Written Examination
Conclusions:
The panel overtums the failures on both of the written examinations. Mr. Calabrese's score was raised to 87% with 76 correct out of 87 total questions. Mr. Robinson's score was raised to 87% with 76 correct out of 87 total questions. Operating Test Review The region graded Mr. Calabrese a failure of the operating test per the Procedures competency. The appeal panel reviewed information from the examiners, the Region I second review, facility training input, and the candidate's appeal package. It should be noted that the facility failed to retain the chart recordings after the scenarios and thus we are f ') unable to independently affirm or deny the candidates claim that Low Pressure injection did ' \ j not occur. This could have added critical information in determining the safety impact of the
5 (3 , l candidaic's performance. Additionally, the candidate and the examiners stated that no level increase occurred, but with the opening of the ADS valves, there should have been a noticeable increase, due to the step change in F.PV pressure. The panel recommends the candidate be graded as follows: l C.4.(a) The panel believes that the candidate did not meet procedural guidance or licensee expectations during the Steam Line Break with ATWS scenario. The candidate was faced with a Rapid Depressurization (RD) situation and did not refer to E0-100-112, " Rapid l Depressurization"., Rev. 7, prior to directing the RD. Licensee management requires SROs to i reference the RD procedure every time it is called upon. The panel rates the Procedures l
" Reference" competency a '1'.
C.4.(b) The candidate in the SRO position, did not use procedures correctly, including implementing the procedural steps in correct sequence resulting in an unnecessary degradation of the plant. The secondary containment control procedure required the SRO to RD based on two l areas being above max safe radiation levels. The RD procedure requires: RD-1 EXIT RPV PRESSURE CONTROL RC/P OR LQlP. RD-2 PREVENT UNCONTROLLED COND INJECTION EXCEPT AS REQ'O !(~ ]/ RD-3
- TO ASSURE ADEQUATE CORE COOLING.
DETERMINE WHETHER RX WILL REMAIN S/D UNDER ALL CONDITIONS W/O BORON >>> NO. RD-5 WHEN ALL RPV INJECTION IS STOPPED AND PREVENTED LAW EO-100-113 LQ/L-19 OR LAW EO-100-114 RF-13. YFLLOW
- COOLDOWN > 100 r./HR MAY BE REQ'D.
RD-6 IS SUPP POOL LVL > S' >>> VES RD-7 OPEN ALL ADS VLV'S The candidate directed step RD-7 be performed before step RD-5. The BOP performed step RD-7 as directed, and then questioned the SRO about whether step RD-5 should be performed. The plant was depressurizing, and at 350 psig the SRO directed step RD-5 be performed. Shut-off head of the RHR pump is approximately 280 psig. The examiners stated that injection flow occurred prior to step RD-5 being completed. The injection of water could have caused a power spike and subsequent core damage due to the ATWS condition. The initial examination is an individual evaluation and not a crew evaluation. The panellacks confidence that the candidate would have prevented injection flow at allif the BOP had not prompted him. This could have resulted in a further degradation of the plant and presented a threat to the health and safety of the public. The panel grades the Procedure " Correct Use" competency a T. Operating Test
Conclusions:
[ '} The panel sustains Mr. Calabrese's failure in the operating examination based on a total i j score of 1.5 in the C.4 " Procedures" competency.
I
/ March 3,1997
']
D Failure on your part to request a hearing within 20 days constitutes a wavier of your right I i to demand a hearing and, for the purpose of reapplication under 10 CFR 55.35, renders this letter a notice of final denial of your application, effective as of the date of this letter.
]
For your information, I am enclosing a copy of the staff's resolution of each of your i comments. If you have any questions, please contact Stuart A. Richards, Chief, Operator j Licensing Branch, at (301) 415-1031. ! Sincerely, Bruce A. Boger, Director Division of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation Docket No.: 55 61425
Enclosure:
As stated g cc w/ encl: G. J. Kuczynski, Plant Manager gD W. H. Lowthert, Manager - Nuclear Training DISTRIBUTION: l Central Files ! HOLBRF GMeyer, Rl ; VCurley, RI TPeebles, Rll PSteiner, Ril EPlettner, Rill T3 vuosewe a espy of fNo eseumont,indeste he the bes: 'C' s Copy without ettechment/ enclosure *E' = Copy wMh attachment / enclosure *N* a No copy 0FFICE HOLB/DRCH 6 DDIR/DRCH AJ DIR/DRCH ,' l NAME SRichards # LSpessardd9 BBoger /.5 DATE 02/7L,/97 04/ ?f /97 OJ/ 3 /97
/ OFFICIAL RECORD COPY O
d
m- enn +S4a'e- -6,A- u,, 'y
'_ _ m:u .
.p % UNITED STATE 3 i1 j NUCLEAR REGULATORY COMMISSION WASHINGTON D.C. 30006 4001 I
r . i
***** March 3, 1997 ,l i
J L, Mr. Frank J.' Calabrese Jr. 698 S. Kennedy Drive McAdoo,PA' 18237-1731 l
Dear Mr. Calabrese:
1-in response to your letter of December 19,1996, we have reviewed the grading of the 4 SRO written and operating examination administered to you on October 21 - 23,1996, and have reconsidered the proposed denial issued to you on December 2,1996. In light of the additional information you provided, we have determined that you pas:cd -! the written examination, however we find that you did not pass the operating. test. i Consequently, the proposed denial of your license application is sustained. if you accept the proposed denial and decline to request a hearing within 20 days as discussed below, p 2the proposed denial will become a final denial. You may then reapply for a license in accordance with 10 CFR 55.35, subject to the following conditions:
- a. Because you passed the written examination on October 21,1996, you may request a wavier.of that portion. This wavier will be granted by the NRC and will be valid up to one year from your examination date.
- b. Because you did not pass the operating test administered to you on !
- October 22 - 23,1996, you will be required to retake an operating test.
- c. You may reapply for a license 2 months from the date of this letter.
i F If you do not accept the proposed denial, you may, within 20 days of the date of this '
- letter, request a hearing in accordance with 10 CFR 2.103(b)(2). Submit your request, in writing, to the Secretary of the Commission, U. S. Nuclear Regulatoiy Commission, Washington, D.C. 20555, with a copy to the Assistent General Coursel for Hearings and ,
[ Enforcement, Office of the General Counsel, at the tsame address. l 2 l (
-\
- (;
a_. . . . , . , . - - - - . - , - - - . - - .
I i NRC REVIEW FOR FRANK J. CALABRESE JR. - SRO CANCalDATE in response to a letter from Mr. F:enk J. Calabrese dated December 19,1996, the NRC has reconsidered the proposed denial issued to Mr. Calabrese on December 2,1996, and has reviewed the grading of the written examination and operating test administered on
- October 21 - 23,1996.
l
- i
- CANDIDATE'S CONTENTIONS - WRITTEN EXAMINATION Mr. Calabrosa contend
- s that questions 13,22,64, and 66 were graded incorrectly or too severely. His letter of December 19,1996, provided detailed information that he i concludes supports his contentions. The grading of question 16 was also reconsidered !
based on the appeal of another candidate. l NRC ANALYSIS Question No.13 A valve is tagged with a pink tag during an outage. Repositioning / operation of the valve l can be approved by which one of the following individuals or combinations of individuals?
- s. Only the work group
- b. Only Shift Supervision
- c. Shift Supervision and the Operations Outage Supervisor
- d. The work group and Shift Supervision L
j Answer Key Choice: d The candidate contends that the combination of individuals who can approve repositioning of pink tagged valves is dependent upon the purpose of the pink tag and the personnel involved. The candidate argues that under certain circumstances Operations is the work group associated with the tag and therefore only Shift Supervision (answer 'b') must give permission to reposition / operate the valve. The controlling procedure, Procedure NDAP-QA-0302, Rev. 6, " System Status and . Equipment Control," in section 4.6 under the duties of the Work Group / Worker states:
"When authorized by Operations Shift or Outage Group Supervision and the ) individual or work group requiring the Status Control Tag, a worker may manipulate components... This can include the operation of status control tagged (pink tag) components whern required for venting and draining systems."
d
- , . , - - -- a, a e --r. -
w -- - ,
fS l Although the candidate is correct that under certain circumstances answer 'b' could be ! considered correct, the question as stated provides no information indicating that such circumstances exist. Absent auch information, answer 'd' is the only correct answer for the question. The NRC concludes that the grading of this question should not be changed. Question No.16 i The bundle from location 23-03 is being transferred from the core during off-loed. A leak has occurred requiring an operator to enter containment to investigate. j What is the maximum elevation that the operator can go to in the containment? a.738'
- b. 752'
- c. 767' a d. 779' 1
l Answer Key Choice: b A different candidate appealed this question and contended that answer 'c'should be ^ considered the correct answer based on the applicable procedural requirements. The NRC concluded that the question should be deleted because the procedural guidance on access C above the 767' olevation is confusing, the fuel bundle location in the question does not ( exist, and there is no need for an SRO candidate to be able to apply the knowledge being tested by this question from memory. I 1 Question No. 22 Given the following conditions: A reactor cooldown is in progress. Recirculation pump 1 A was secured at 0815 due to concerns with seal leakage. At 0930, Recirculation pump 1B was inadvertently tripped. At 0945 the 1B pump is restarted. The 18 pump is tripped again at 0950. What is the earliest time the 18 pump is allowed to be started?
. a. 1000
- b. 1005 a c. 1030
- d. 1035 1
^~*""
O 2 y ,-,
O The candidate contends that the question should be deleted because it is inappropriate to expect the candidate to know the subject material from memory. The NRC disagrees because the question was bened on a facility learning objective which requires the candedste to state the reactor recirculation pump restart limitations. The NRC concludes that there should be no change to the grading of this question. Question No. 64 Station Power Restoration, EO-000-031, provides a specific sequence for i reenergizing busses from an off-site source to AVOID: i
- a. diesel generators tripping on overspeed when loads i are transferred to off-site power. l
- b. underfrequency condition on off site sources due to manually ,
i reenergizing non-emergency busses. l
- c. undervoltage condition caused when a ECCS initiation signa; is ;
present. .
- d. starting equipment automatically without operator action. l 1
l Answer Key Choice: c l The candidate argues that choice 'd'should also be accepted as e corect answer due to ) supporting statements in the applicable procedure. The NRC agrees that choice 'd'should I be accepted as an additional correct answer.
\
Question No. 66 $ Given the following: A station blackout has occurred. MAIN STEAM SRV LEAKING is alarming. MAIN STEAM DIV 1 SRV OPEN is clear. MAIN STEAM DIV 2 SRV OPEN is clear. I Based on this information, what is the status of SRVs and equipment to monitor SRVs? I
- a. An SRV is leaking. The acoustic monitors faP during a station blackout.
- b. AH SRVs are closed. Tailpipe temperature indications fail high during a station blackout.
- c. Status of the SRVs is unknown because the annunciators are ,
indications of loss of power to instrumentation.
- d. An SRV has opened, then reclosed, causing the acoustic monitors to
'+ clear. Answer Key Choice: a 3
?
i The candidate contends that none of the answers are correct, in part because insufficient 4 information is provided to determine the status of the SRVs. The question asks for the status of the SRVs and SRV instrumentation based on the information provided. The NRC review concludes that during a station blackout, tailpipe l t temperature indication would be available and the acoustic monitors would be unavailable l due to loss of power. Based on the information provided, the status of the SRVs (leaking, I closed or open) cannot be confirmed. However, answer 'b' is incorrect with respect to the - status of. instrumentation because the tailpipe temperature indications do not fail high during a station blackout. Answer 'c' is incorrect with respect to the status of SRV
; instrumentation because the SRV leaking annunciator is not due to loss of power. Answer )
4
'd' is incorrect with respect to the status of SRV instrumentatien because the acoustic !
. monitors are doenergized, not cleared. Answer 'a'is not incorre':t with respect to the i I l status of SRVs and is correct with respect to the status of the SHVs instrumentation. Even though SRV status cannot be confirmed and none of the answers are incorrect with j
< respect to SRV status, the question is still valid with respect to the status of SRV !
4 instrumentation. The applicants should be able to determine that answers 'b', 'c', and 'd' l 1 are incorrect based on the status of SRV instrumentation. Answer 'a' is therefore the only , a correct answer. l NRC CONCLUSION - WRITTEN EXAMINATION l The NRC concluded tnat question 16 should be deleted and a second correct answer . accepted for question 64. The candidate therefore has 73 correct answers on a 91 question test for an overall score of 80.2%. i ! CANDIDATE'S CONTENTIONS - OPERATING TEST ! The candidate contends that rating factors C.4.A, C.4.B, and C.7.B were graded
- incorrectly or too severely. At issue is the candidate's actions during the major transient
- portion of the second scenario. The scenario involved a steam line break in the secondary containment, coupled with the failure of seven control rods to insert into the reactor core.
As the Senior Reactor Operator during the scenario, the candidate was required to perform i a rapid depressurization in accordance with EO-100-112, " Rapid Depressurization." The
- NRC examination comments state that the candidate did not refer to the procedure prior to
! directing the rapid depressurization and then directed that the ADS valves be opened prior to RPV injection being stopped, contrary to the procedure. Consequently, the candidate was graded a '1' in rating factors C.4.A, Procedures-Reference, and C.4.B, Procedures- . Correct Use. Overall the candidate was graded a 1.5 in the Procedures competency and thereby failed the operating portion of the examination. The candidate contends that he correctly ordered the actions required by the procedure in the proper sequence. The candidate states that the board operator (PCOX) incorrectly l l carried out his direction and that he then reordered the proper action. He further contends i that no injection of the RPV occurred. j i .( l [ 4 . i i i
- - - - , , r--.e m ., ,, . en. e , ,s.- .c, ,.w,w *,1-, , . . . -.. m. .-. , ,,,.., ---zw.- , m. rn amv.- re- mm--o,--+a
J r ( NRC ANALYSIS The NRC examiner's notes and recollection support the original grading of the candidate. Whether there was injection or not, the candidate was observed by the examiner to incorrectly order steps to implement the rapid depressurization, without reference to the
- applicable procedure. The errors observed were safety significant and support the grading
- of '1' in the two rating factors. Absent additional information, the NRC concludes that .
revision of the grading of this competency is not warranted. The candidate's contentions regarding the grading of rating factor C.7.8 were not considered because the candidate received an overall passing grade for the associated competency. NRC CONCLUSION - OPERATING TEST No revision of the original grading is warranted. _ i i 9 O v 5
1 O\s / DOCKETED USNRC Frank J. Calabrese Jr. 698 South Kennedy Drive McAdoo, Pa. 18237-1731-17 MR 18 P4 :10 (717) 929-1577 i l 0FFICE OF SECRETARY ' 00CKETING & SERVICE BRANCH
, Secretary of The Commission U.S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Secretary,
l
- This letter is to request a hearing in accordance with l 10 CFR 2.103 (b) (2) per your letter of March 3, 1997 as I do not accept the proposed denial. ;
I am satisfied with the outcome of the Written Exam, i but am not satisfied with the grading of the Operating (Simulator) Exam. I believe my Simulator Exam has been graded incorrectly or too severely as I have stated in my previous request of an informal NRC Staff review dated December 19, 1996. l ' g- sI 'Thank you for your kind consideration in this matter which is of utmost importance; Please provide me the g N-- information on when and where the hearing will be held at your earliest convenience. If you have any other comments , questions, or concerns, contact me at the above address or by phone at (717) 929-1577 Sincerely, T(6JLO F.J. Calabrese Jr.
, y/<why cc: Assistant General Counsel For Hearings and Enforcement Office of General Counsel US Nuclear Regulatory Commission Washington, D. C. 20555 1
/ \
N
a nob . UNITED STATES
/ *; NUCLEAR REGULATORY COMMISSION g WASMNGTON, D.C. 20$55 0001
%,.....) March 25, 1997 SECRETARY MEMORANDUM TO: B. Paul Cotter, Jr.
Chief Administrative Judge Atomic S fety and Licensing Board Panel J L FROM: Jo C. Hoyle, ecretary 4
SUBJECT:
REQUEST FOR HEARING SUBMITTED BY FRANK J. CALABRESE, JR. i Attached is a request for hearing dated March 14,1997, submitted by Frank J. Calabrese, Jr. (Docket No. 55-61425). The hearing request is in response to a letter from the NRC staff dated March 3,1997, sustaining a denial of Mr. Calabrese's senior reactor operator's license application. (]
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Mr. Calabrese's request for hearing and additional documentation (including his letter to the NRC staff dated December 19,1996) are being referred to you for appropriate action in accordance with 10 C.F.R. Sec. 2.1261. j l Attachments: As stated cc: Commission Legal Assistants OGC l l CAA l OPA EDO , l < NRR Frank J. Calabrese, Jr. 4 h
4 l l J DOCKETED USNRC Frank J. Calabrese Jr. f'~'s 698 South Kennedy Drive ; ( ) McAdoo, Pa. 18237-1731 37 PMR 18 P4 :10 (717) 929-1577 0FFICE DF SECRETARY 00CKETING & SERVICE BRANCH l Secretary of The Commission U.S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Secretary,
This letter is to request a hearing in accordance with 10 CFR 2.103 (b) (2) per your letter of March 3, 1997 as I do not accept the proposed denial. I am satisfied with the outcome of the Written Exam, but am not satisfied with the grading of the operating (Simulator) Exam. I believe my Simulator Exam has been graded incorrectly or too severely as I have stated in my previous request of an informal NRC Staff review dated December 19, 1996. Thank you for your kind consideration in this matter which is of utmost importance. Please provide me the
/~~N information on when and where the hearing will be held at
( j' your earliest convenience. If you have any other comments , x questions, or concerns, contact me at the above address or by phone at (717) 929-1577 Sincerely, ) TA . 1 F.J. Calabrese Jr. 2 /,vh1 cc: Assistant General Counsel For Hearings and Enforcement Office of General Counsel US Nuclear Regulatory Commission Washington, D. C. 10555 ? m \ v
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.y +4 UNITED STATES
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' E' j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20086 4001 e (
***** March 3, 1997 l
Mr. Frank J. Calabrese Jr. 698 S. Kennedy Drive McAdoo,PA 18237-1731 , D' ear Mr. Calabrese: ,1 In response to your letter of December 19,1996, we have reviewed the grading of the 3 . SRO written ena operating examination administered to you on October 21 - 23,1996, j . and have reconsidered the proposed denialissued to you on December 2,1996. In light of the additional information you provided, we have determined that you pas::cd the written examination, however we find that you did not pass the operating test. 4 Consequently, the proposed denial of your license application is sustained. If you accept the proposed denial and decline to request a hearing within 20 days as discussed below, the proposed denial will become a final denial. You may then reapply for a license in ,
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accordance with 10 CFR 55.35, subject to the following conditions:
- s. Because you passed the written examination on October 21,1996,'you may
, request a wavier of that portion. This wavier will be granted by the NRC and will be valid up to one year from your examination date. [
- b. Because you did not pass the operating test administered to you on l
October 22 - 23,1996, you will be required to retake an operating test.
- c. You may reapply for a license 2 months from the date of this letter.
If you do not accept the proposed denial, you may, within 20 days of the date of this letter, request a hearing in accordance with 10 CFR 2.103(b)(2). Submit your request, in I writing, to the Secretary of the Commission, U. S. Nuclear Regulatory Commission, !. Washington, D.C. 20555, with a copy to the Assistant General Counsel for Hearings and
- Enforcement, Office of the General Counsel, at the same address.
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March 3,1997 Failure on your part to request a hearing within 20 days constitutes a wavier of your right to demand a hearing and, for the purpose of reapplication under 10 CFR 55.35, renders this letter a notice of final denial of your application, effective as of the date of this letter. For your information, I am enclosing a copy of the staff's resolution of each of your , comments. If you have any questions, please contact Stuart A. Richards, Chief, Operator Licensing Branch, at (301) 415-1031. ' Sincerely, i 1 1 Bruce A. Boger, Director Division of Reactor Controls ] l and Human Factors Office of Nuclear Reactor Regulation Docket No.: 55-61425
Enclosure:
As stated cc w/ encl: G. J. Kuczynski, Plant Manager
) W. H. Lowthert, Manager - Nuclear Training DISTRIBUTION:
Central Files HOLB RF GMeyer, RI VCurley, RI TPeebles, Ril PSteiner, Rll EPlettner, Rlli e.ue m nu.n.e r . c , .*h .n. chm.no.new. w . u. e.,y v3 w. . , .e me. 4 we w m. % c . c.,y 0FFICE HOLB/DRCH l0 DDIR/DRCH AJ LSpessard6 DIR/DRCH BBoger
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( NAME SRichards # _ . DATE 02/%/97 OJ/ 2, /97 Oj/ J /97
/ OFFICIALRECORDCCfY p
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I i , NRC REVIEW FOR FRANK J. CALABRESE JR. - SRO CANDIDATE
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. 1 In response to a letter from Mr. Frank J. Calabrese dated December 19,1996, the NRC has reconsidered the proposed denial issued to Mr. Calabrese on December 2,1996, and has reviewed the grading of the written examination and operating test administered on October 21 - 23,1996.
CANDIDATE'S CONTENTIONS - WRITTEN EXAMINATION Mr. Calabrese contends that questions 13,22,64, and 66 were graded incorrectly or too severely. His letter of December 19,1996, provided detailed information that he concludes supports his contentions. The grading of question 16 was also reconsidered based on the appeal of another candidate. 4 NRC ANALYSIS I Question No.13 A valve is tagged with a pink tag during an outage. Repositioning / operation of the valve can be approved by which one of the following individuals or combinations of individuals?
- a. Only the work group
- b. Only Shift Supervision
- c. Shift Supervision and the Operations Outage Supervisor
- d. The work group and Shift Supervision Answer Key Choice: d .
The candidate contends that the combination of individuals who can approve repositioning of pink tagged valves is dependent upon the purpose of the pink tag and the personnel involved. The candidate argues that under certain circumstances Operations is the work group associated with the tag and therefore only Shift Supervision (answer 'b') must give permission to reposition /operata the valve. The controlling procedure, Procedure NDAP-QA-0302, Rev. 6, " System Status and Equipment Control," in section 4.6 under the duties of the Work Group, Worker states:
"When authorized by Operations Shift or Outage Group Supervision and the individual or work group requiring the Status Control Tag, a worker may manipulate components... This can include the operation of status control tagged (pink tag) components when required for venting and draining systems."
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i 8 Although the candidate is correct that under certain circumstances answer 'b' could be concedered correct, the question as stated provides no information indicating that such circumstances exist. Absent such information, answer 'd' is the only correct answer for the question. The NRC concludes that the grading of this question should not be changed. Question No.16 The bundle from location 23-03 is being transferred from the core during off- .! load. A leak has occurred requiring an operator to enter containment to
! investigate.
What is the maximum elevation that the operator can go to in the containment?
- a. 738'
) b. 752'
- c. 767' \
- d. 779' Answer Key Choice: b A different candidate appealed this question and contended that answer 'c'should be considered the correct answer based on the applicable procedural requirements. The NRC concluded that the question should be deleted because the procedural guidance on access above the 767' elevation is confusing, the fuel bundle location in the question does not exist, and there is no need for an SRO candidate to be able to apply the knowledge being tested by this question from memory.
Ouestion No. 22 i Given the following conditions:
- A reactor cooldown is in progress.
- Recirculation pump 1 A was secured at 0815 due to concerns with seal leakage.
- At 0930, Recirculation pump 1B was inadvertently tripped.
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- At 0945 the 18 pump is restarted.
- The 1B pump is tripped again at 0950.
I What is the earliest time the 1B pump is allowed to be started?
- a. 1000
- b. 1005
- c. 1030
- d. 1035 Answer Key Choice: d 2 j
l ! The candedate contends that the question should be deleted because it is inappropriate to
- expect the candedate to know the subject material from memory. The NRC disagrees because the question was bened on a facility learning objective which requires the candidate to state the reactor recirculation pump restart limitations. The NRC concludes that there should be no change to the grad:ng of this question.
i Question No. 64 Station Power Restoration, EO-000-031, provides a specific sequence for 4 reenergizing busaes from an off-site source to AVOID:
- a. diesel generators tripping on overspeed when loads i are transferred to off site power.
- b. underfrequency condition on off-site sources due to manually reenergizing non-emergency busses.
- c. undervoltage condition caused when a ECCS initiation signai is '
present.
- d. starting equipment automatically without operator action.
! . Answer Key Choice: c The condidate argues that choice 'd'should also be accepted as a correct answer due to supporting statements in the applicable procedure. The NRC agrees that choice 'd'should ~ be accepted as an additional correct answer. ; Ouestion No. 66 l I Given the following: A station blackout has occurred.
- MAIN STEAM SRV LEAKING is alarming.
- MAIN STEAM DIV 1 SRV OPEN is clear.
MAIN STEAM DIV 2 SRV OPEN is clear.
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Based on this information, what is the status of SRVs and equipment to monitor SRVs?
- a. An SRV is leaking. The acoustic monitors fait during a station blackout.
' All SRVs are closed. Tailpipe temperature indications fail high during b. a station blackout.
- c. Status of the SRVs is unknown because the annunciators are indications of loss of power to instrumentation.
.d. An SRV has opened, then reclosed, causing the acoustic monitors to clear. l Answer Key Choice: a 4 .
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l The candidate contends that none of the answers are correct, in part because insufficient information is provided to determine the status of the SRVs. The quest.on asks for the status of the SRVs and SRV instrumentation based on the information provided. The NRC review concludes that during a station blackout, tailpipe temperature indication would be available and the acoustic monitors would be unavailable due to loss of power. Based on the information provided, the status of the SRVs (leaking, closed or open) cannot be confirmed. However, answer 'b' is incorrect with respect to the status of instrumentation because the tailpipe temperature indications do not fail high during a station blackout. Answer 'c'is incorrect with respect to the status of SRV l instrutnentation because the SRV leaking annunciator is not due to loss of power. Answer
'd' is incorrect with respect to the status of SRV instrumentation because the acoustic monitors are deonergized, not cleared. Answer 'a'is not incorrect with respect to the status of SRVs and is correct with respect to the status of the SRVs instrumentation.
Even though SRV. status cannot be confirmed and none of the answers are incorrect with respect to SRV status, the question is still valid with respect to the status of SRV instrumentation. The applicants should be able to determine that answers 'b', 'c', and 'd' are incorrect based on the status of SRV instrumentation. Answer 'a' is therefore the only correct answer. NRC CONCLUSION - WRITTEN EXAMINATION The NRC concluded that question 16 should be deleted and a second correct answer accepted for question 64. The candidate therefore has 73 correct answers on a 91 question test for an overall score of 80.2%. CANDIDATE'S CONTENTIONS OPERATING TEST The candidate contends that rating factors C.4.A, C.4.8, and C.7.B were graded incorrectly or too severely. At issue is the candidate's actions during the major transient portion of the second scenario. The scenario involved a steam line break in the secondary containment, coupled with the failure of seven control rods to insert into the reactor core.
- As the Senior Reector Operator during the scenario, the candidate was required to perform a rapid depressurization in accordance with EO-100112, " Rapid Depressurization." The NRC examination comments state that the candidate did not refer to the procedure prior to directing the rapid depressurization and then directed that the ADS valves be opened prior to RPV injection being stopped, contrary to the procedure. Consequently, the candidate was graded a '1' in rating factors C.4.A, Procedures Reference, and C.4.B, Procedures-Correct Use. Overall the candidate was graded a 1.5 in the Procedures competency and thereby failed the operating portion of the examination.
The candidate contends that he correctly ordered the actions required by the procedure in the proper sequence. The candidate states that the board operator (PCOX) incorrectly carried out his direction and that he then reordered the proper action. He further contends O that no injection of the RPV occurred. 4 4
l NRC ANALYSIS - The NRC examiner's notes and recollect.m support the original grading of the candidate. Whether there was injection or not, the candidate was observed by the examiner to incorrectly order steps to implement the rapid depressurization, without reference to the applicable procedure. The errors observed were safety significant and support the grading of '1' in the two rating factors. Absent additional information, the NRC concludes that , revision of the grading of this competency is not warranted. l l The candidate's contentions regarding the grading of rating factor C.7.B were not considered because the candidate received an overall passing grade for the associated competency. NRC CONCLUSION - OPERATING TEST i No revision of the original grading is warranted. t t f 5
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) I December 19,1996
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Frank J. Calabrese Jr, 698 S. Kennedy Drive McAdoo,PA 18237-1731 ~ Director Division of Reactor Controls
- and Human Factors Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission l
' Washington, DC 20555 l T l l 4
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Dear Director:
This letteris to request an informal NRC staff review of the grading of my written SRO l examination which was administered on October 2124,1996, as I do not accept the l
- proposed denial.
I believe the written examination and the simulator examination have been graded ; incorredly or too severely. Please see the attached justification for the written
- exarninatsen and the ' simulator examination topics, for which I request reconsideration.
Enclosed is my sin.2htor examination basis and the basis for Questions 13,22,64, and
- 66. For your review purposes, I have also enclosed the simulator exaministion
- t comments from my proposed denialletter of December 2 1996.
4
' Thank you for your kind consideration in this matter which is of utmost importance, if ;
you have any further questions, comments, or concems, please contact me at the above address or via phone at (717) 929-1577. Sincerely. - l-
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. Frank J. Calabrese Jr.
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l I O . SIMULATOR EXAMINATION BASIS O O Page 1 of 2 s
! l [^ Comment C.4.A - During this time in the scenario, I had directed the PCOX to open (6) ADS valves and override Lo Press ECCS at =900#. The PCOX then began to open (6) ADS valves and override Lo Press ECCS ! as directed. I then reordered override of all Lo Press ECCS again l at 350# when i noticed the "A" RHR Pp running. (Note: 350# is l: well above the pressure at which RHR would iniect to the vessel.) I saw no injection, no level increase, and no power increase as a result of these actions. Comment C.4.B - (Same Rebuttal as Comment C.4.A) l Comment C.7.B - During the scenario in question, I never assumed primary j i responsibility for monitoring secondary containment rad levels. When initial rad levels were observed, I directed the PCOU to place rad format 1/S and maintain primary responsibility for monitoring. I edvised I would try to "back him up" with the CRT at my console. I was advised by the PCOU when we had exceeded max safe rad in (2) areas. I then confirmed his indications and proceeded to order rapid depressunzation of the plant as directed in EO-100-104. Conclusion - Therefore, I believe the grading of my simulator performance is not j only too severe, but somewhat incorrect. The scenario involved l O, several major failures of plant equipment which were dealt with in accordance with plant procedures, and no incidents of public safety concem were raised. I respectfully request that the simulator test be reconsidered. i l 0 l V Page 2 of 2 l l l
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APPLICANT DOCKET N0: 55-61425 PAGE 6 of 8 p FORM ES 302 2 Cross Reference Comments C.4.A The candidate failed to refer correctly to important procedures in important instances. The candidate was acting in the position of the Senior Reactor Operator (SRO) during the major transient of the second scenario. The scenario involved fuel failure, a steam line break in secondary containment with seven control rods failing to insert. The candidate was in E0-100-104. " Secondary > Containment Control". step SC/R-6 that directs when area radiation levels exceed maximum safe levels in two or more areas to rapidly depressurize the reactor. The candidate gave the order to open six ADS valves to depressurize the reactor without first referring to procedure E0-100-112. " Rapid Depressurization" Procedure E0-100-112. step RD-5 required action to stop and prevent all RPV injection prior to opening the ADS valves. The balance of plant operator (80P) opened the ADS valves as I directed then indicated that he would need to secure the low pressure emergency core cooling systems (ECCS). The candidate then gave the order to override all low pressure ECCS (approximately two minutes after the ADS valves were opened). By ! the time the order was given reactor pressure had already g decreased to approximately 350 psig. The BOP completed the i actions to override all low pressure ECCS systems, but one of the (b low pressure coolant injection systems injected cold water into the reactor vessel before the pumps were secured. The candidate did not refer to E0-100-112 until after the low pressure coolant injection systems had been overridden. The candidate's failure to refer to E0-100-112 prior to directing action to rapidly depressurize the RPV resulted in failure to stop and prevent RPV injection prior to depressurization. As a result an injection of cold water occurred when it was not assured that the reactor would remain shutdown under all conditions without boron. The reactivity addition from the cold water injection could have caused a reactor power excursion and substantial core damage. The candidate failed to refer to the procedure in an important instance. K/A 295015 G.12 (3.7/4.4) E0-100-104 & E0-100-112 10 CFR 55.45(a)(13) V
APPLICANT DOCKET N0: ~ 55 61425 PAGE 7 of 8 FORM ES-302 2 Cross Reference Comments C.4.8 As described in C.4.A. the candidate while acting in the position
.of SR0 during the major transient of the second scenario failed to use procedures correctly resulting in significant errors that degraded the plant unnecessarily. When rapid depressurization is-required and it has not been determined that the reactor will remain shutdown under all conditions without boron, step RD-5 of E0-100-112 directs the operator to wait until all RPV injection is stopped and prevented in accordance with step LO/L-19 of E0-100- t 113. " Level / Power Control." before opening the ADS valves. Step LO/L-19 of E0-100-113 directs the operator to stop and prevent injection except from SLC. CRD. RCIC, and'HPCI. The candidate's direction to open the ADS valves before the low pressure emergency coolant systems and condensate had been overridden was not in accordance with the direction in E0-100-112.
The candidate's failure to correctly imalement E0-100-112 resulted in an injection of cold water into the RPV when it was not' assured , that the reactor would remain shutdown under all conditions without boron. The reactivity addition from the cold water injection could have caused a reactor power excursion and substantial core damage. The candidate made a significant error in the use of procedures that degraded the plant unnecessarily. : ] K/A 295015 G.12 (3.7/4.4) E0-100-112 & E0-100-113 10 CFR 55.45(a)(13) 4 O
i APPLICANT DOCKET N0: 55 61425 PAGE 8 of 8 e) O FORM ES 302 2 Cross Reference Comments C.7.B While the candidate was acting in the position of the SR0 during the may)r transient of the second scenario, he failed to provide timely, ell thought 'out directions that demonstrated appro)riate concern for the safety of the plant. During the scenario t1e . candidate's two major concerns were inserting the rods that had failed to scram and monitoring increasing radiation levels. The candidate was, in E0-100-104. " Secondary Containment Control." and had assumed responsibility for monitoring secondary radiation levels. The candidate failed to monitor conditions closely and as a result. two areas had exceeded maximum safe radiation levels for approximately five minutes before it was recognized by the crew. When it was recognized that two areas had exceeded max safe levels. the candidate failed to provide well thought out direction to rapidly depressurize as discussed in C.4.A. l I The candidate's failure to provide timely direction to rapidly depressurize the RPV when radiation levels in two areas of secondary containment were above max safe allowed radiation levels j in the secondary containment to continue to increase unnecessarily. Allowing radiation levels to increase in the control rod drive areas could have resulted in higher personnel (Un) exposures if operators had to enter the area to attempt to insert the control rods ttat were still withdrawn. The candidate's failure to provide well thought out direction for rapid depressurization when it had not been assured that the ; reactor would remain shutdown under all conditions without boron resulted in an injection of cold water into the RPV. The reactivity addition from the cold water injection could have caused a reactor power excursion and substantial core damage. K/A 295033 A2.01 (3.8/3.9) K/A 295033 G.12 (3.8/4.4) E0-100-104 & E0-100-112 10 CFR 55.45(a)(13) b h
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),RO EXAMINATION QUESTION NO.13 A valve is tagged with a pink tag during an outage. Repositioning / operation of the valve e sn be approved by which one of the following individuals or combinations of individuals?
- a. Only the work group
- b. Only Shift Suparvision
- c. Shift Supervision and the Operations Outage Supervisor
- d. The work group and Shift Supervision Answer Key Choice g Candidate's Choice b i
Basis The individual or combination of individuals who can approve the repositioning / operation i of a valve that is pink tagged during an outage is dependent upon the purpose of the pink tag and personnelinvolved. For example, during outages, pink tags are utilized to identify and control the position of [ valves which form the piping structuralintegrity boundary for operable systems. The only group that would require the repositioning / operation of a pink tagged boundary valve is Operations. Also, in accordance with NDAP-QA-0302, Section 4.4.4, only Shift Supervision is responsible for tracking LCOs and TROs and maintaining the Daily LCO and TRO Logs. Since the repositioning / operation of a pink tagged boundary valve may affect equipment or system operability, the valve operation must be approved by Shift < Supervision only, in this situation, no work group is involved and outage group i supervision is specifically excluded from tracking and maintaining the LCO and TRO Logs. Therefore, in accordance with NDAP-QA-0302, Section 6.3.14, the only group that can permit the repositioning / operation of components is Operations Shift Supervision. I contend Choice "b" is the correct choice based upon the above example. 1' Supportina Documentation NDAP-QA-0302; Rev. 6; Pages 1,12,13,17, and 26 l l l l 1
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rm PROCEDURE COVER SHEET NUCLEAR DEPARTMENT PROCEDURE A N s. h K SYSTEM STATUS AND EQUIPMENT CONTROL NDAP-QA-0302 Revision 6 J Page 1 of 89
$ l%ZD EFFECTIVE DATE: g_7_ g PERIODIC REVIEW FREQUENCY: g gg PERIODIC REVIEW DUE DATE: g/ h l l
REVISED PERIODIC REVIEW DUE DATE: ; PROCEDURE TYPE: QA Program (X) YES ( ) NO Plant Procedure (X) YES ( ) NO 4 REVIEW METHOD: 1 () Altemate ( ) Expedited ( [ PORC j) ERC Prepared by , m4 'Date 3/L6lf4 Reviewed by & h )A L / (/p 'Sdpervisor ' Date 3f26 /fI-l Recommended u d. Ok -- l- Date ' J/A7f7[o Functional Unit Manager / 96- 031 Date J /z v/% PORC Committee Meeting No. jUk Date ERC Committee Meeting No. Approved t,, V J ,27- fg fp Date FORM NDAP-QA-D002-1, flev'.1, Page 1 of 1 i
O PCAF #! # M IS NDAP-QA-0302 ! V PAGE- 9 -0F If Revisio~n 6 Page 12 of 89 ' 4.2 Day Shift Supervisor 4.2.1 Reviewing the Daily LCO,4 e TaoLog Sheets. 1 4.2.2 Approving extensions of Status Control Tag removal dates. 4.2.3 Designating an individual to perform monthly audits of the Status Control forms. 4.3 Shift Supervisor / Outage Group Supervisor 4.3.1 Ensuring that system status and equipment controlis maintained ! in accordance with this procedure. , 1 4.3.2 Maintaining unit separation. i i 4.4 Shift Supervision / Outage Group Supervision: . t 4.4.1 Ensuring the status of each system is properly determined, ; D maintained and controlled. 4.4.2 Releasing systems, equipment and components for work after ; proper classification as to their safety status significance and
- impact on Operational status.
4.4.3 Authorizing and controlling changes in the position of plant equipment. aal Tt**s 4.4.4 Shift Supervision, only: Tracking LCO's'and maintaining the Daily logs, LCO,w Tao 4.4.5 Ensuring systems are properly retumed to operable status, only ! l after required Operability / Operability Testing has been completed.
- 4.4.6 Approving issuance of Status Control Tags and ensuring Status i
Control Tags are being issued for equipment protection and/or
- . status control.
4.4.7 Reviewing instructions for Status Control Tags to ensure direction given does not deviate from established : station procedures, policies, or Technical Specifications. 4.4.8 Authorizing Status Control Tag removal, and ensuring the Status fG Control forms and Status Control Tag index are properly completed.
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< p - NDAP-QA-0302 !
( . Revision 6 Page 13 of 89 ~ l 4.4.9 When L nit Coordinator is not available, Shift Supervision may
*N/A" their review space on form NDAP-QA-0502 7. ;
4.5 Operations (") 4.5.1 Ensuring that a removal / corrective mechanism for the Status , Control Tag is in place,'and that appropriate documentation occurs n accordance with this procedure.
.4.5.2 Ensurirg a completed copy of the Status Control form, NDAP-QA-0302-4 is sent to the appropriate System Engineer, i when rulated to svstem performance.
J 4.5.3 Operators are responsible for monitoring and maintaining the ; status uf plant systems and control of equipment / components. l
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4 (") 4.5.4 Plant oserators are responsible for ensuring all LLRT tcgs, Red Tags, titriped Tags, Status Control Tags are removed, properly disposed of upon clearance, and associated control forms are updatei i (") 4.5.5 Plant Control Operator shall account for associated control forms for 4.5.4 above and ensure form completion. 4.6 Work Group / Worker 4.6.1 Applyirig and removing Status Control Tags and LLRT Tags, when approved by Operations; and completing associated forms. 4.6.2 Returning Stutus Control Tags and LLRT Tags to Operations when removed, except those which are contaminated. 4.6.3 When authonzed by Operations Shift or Outage Group Supervision and the individual or work group requiring the Status Control Tag (when applicable), a worker may manipulate components. Workers shall monitor the effected system for any changes as a result of the component manipulation. This can include the operation of Status Control Tagged (Pink Tag) compenents when required for venting and draining systems. 4.7 Maintenance , (") 4.7.1 Assun.ing ownership of Status Control Tags in cases where
.O correcuve/ preventative maintenance is required.
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1 l i l l NDAP-QA-0302 9 Revision 6 Page 17 of 89 5.12 Status Control Tags (PT) Standard printed Status Control Tags (Attachment E) which are attached to . operating devices to denote the device is temporarily in a controlled status and i may only be operated, or have its position changed with the permission of l individual or work group who required the tag and Albat Shift Supervision, or Operations Outage GroJp Supervision. These tags are Neon Pink in color = PT for Pink Tag. l 5:12.1 Status C 2ntrol Tags may also be under the controls of a permit. When Status Control Tags are applied by a permit, NDAP-Q A-0322, Permit and Tag, will direct their application, operation, and removal. 5.13 Status WA i 1 i ~ WA's entered in the Syt tem Status File to provide status control documentation. These WA's are used fc r tracking equipment status and are not to be used to l ( n I release work on plant systems. These WA's will use a 'z' prefix. Since no work plan will be associated with them, the Operations WA Review shall be entered as k./ N/A on the System Stat;s File. 1
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Revision 6 j Page 26 of 89 l 6.3.11 Status Control Tag index (form NDAP-QA-0302-1 Attachment B) is filed in the Work Control Center for refuel inspection outages I otherwise Unit #1 and Unit #2 control rooms maintain separate j g index books for non refuel outage conditions. l 6.3.12' The' Status Control form is filed with Operations in System Status File.
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6.3.13 Status Control Tags sha'il not be applied to' an operating device for !
. longer than six months and shall.be removed following their i expiration date. Upon submission of written justification an l extension in the Status Control Tag expiration date of up to the next Refueling and inspection Outage may be approved by Manager-Nuclear Operations or Day Shift Supervisor. If an extension is granted the documentation shall be attached to the Status Control form. e NOTE: If a Status Control Tag (s) has expired and the Shift Supervisor or Outage Group Supervisor determines that removal of Status
. Control Tag (s) may adversely effect plant operation the Status , Control Tag (s) may remain applied pending approval by the l Manager-Nuclear Operation or Day Shift Supervisor. 6.3.14 Repositioning / operating components controlled by Status Control Tags may be performed with the permission of the individual or ! work group who required the tag and either Operations Shift j Supervision or Operations Outage Group Supervision. ! 6.3.15 A Status Control Tagged component may be removed from the system when removal of the component is required to perform maintenance on the tagged component as follows: NOTE: This step shall not apply to components status tagged under the Permit and Tag Process. Only . those tags applied by this procedure may be removed from the system. l l' a. A note shall be added to the Status Control Form.
! b. The work group shall document the activity and Status i Control Form Number in their work package.
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l l I SRO EXAMINATION QUESTION NO. 22 v Given the following conditions:
- A reactor cooldown is in progress.
- Recirculation pump 1 A was secured at 0815 due to concems with seal leakage.
- At 0930 Recirculation pump 1B was inadvertently tripped.
- At 0945 the 1B pump is restarted. j
- The 1B pump is tripped again at 0950.
What is the earliest time the 18 pump is allowed to be started?
- a. 1000
- b. 1005
- c. 1030
- d. 1035 Answer Key Choice d ,
Candidate's Choice g ; 4 V Basis l I The K&A objective being tested by this question is; l KA: 202001G010 Ability to explain and apply all system limits , and precautions. l 1 I Asking for the recirculation pump motor restart times in a closed book examination does not meet the stated K&A objective. it is expected that the operator be cognizant of large motor restart precautions, but not memorize the exact time for each motor. The restart of I a recirculation pump is a significant operation which would always be performed in accordance with OP-164-001, Reactor Recirculation System. OP-164-001, Section ' 3.3.25, includes a " NOTE" with the restart time for the recirculation pumps. I contend this question is inappropriate for the subject SRO examination. The question j should be deleted from the examination. J Supportino Documentation OP-1M-001; Rev. 26; Page 12; Section 3.3.25
'rvA Catalog; Rev. 0; Page 3.1-24 ;
m K/A Catalog; Rev.1; Page 2-4 : l
i t n OP-164-001
/
Revision 26 C) Page 12 of 35 ; NOTE: Time interval between pump start and opening of i discharge valve should be minimized to preclude ~ f . possible pump overheating. On pump start vessel level will decrease. Vessel level control should be monitored until pump has completed starting sequence. CAUTION STARTING RECIRCULATION PUMP WHILE AT POWER WILL RESULT IN INSERTION OF. : POSITIVE REACTIVITY. !
- 3. 3 ~. 23 PLOT all power changes on Power / Flow Map, Form NDAP-QA-0338-10.
3.3.24 If GETARS available, INSTRUCT STA to start GETARS to collect data for pump start or if STA is not available, DEPRESS GETARS INIT pushbutton on PC0 desk. 3.3.25 START Reactor Recirc Pump 1P401A(B) by depressing MG i SET A(B) DRV MTR BKR HS-14001A(B) START push button m- - one (1) second (to allow start sequence relay to seal in). NOTE: With motor windings at ambient temperature (s 104*F), motor may be started and i brought to speed two times in succession. With motor windings at rated temperature (> 104*F), motor may be started and brought to speed once. After all permissible starts have been made, windings must return to rated or ambient t temperature before further starting attempts may be made. Motor windings can be assumed to have returned to rated temperature after 45 minutes shutdown or i after 15 minutes running at rated speed. 3.3.26 OBSERVE: ,
- a. MG SET A(B) DRIVE MTR BKR CLOSES.
- b. GEN 1A(IB) SPEED indication INCREASES. ,
- c. After 11 seconds, GENERATOR A(B) FIELD BREAKER I closed indicator light ILLUMINATES. :
(m~) .
)
T W SYSTEM: 202001 Recirculation System- i
. Tasks as -noted prev fously- .
IMPORTANCE -f R0 SRO-
, s.
SYSTEM GENERIC K/As .{ tl. Knowledge _ of operator responsibilities during all modes of . l
' plant operation.
- 3. 9 _3.9 l
- 2. ;tKnowledge.of. system status criteria which require the .
. notification of plant personnel. = 3.0 3.8- 1 <
- 3. tKnowle'dge of which events related to system operation / status should be reported to outside agencies. 2.9*. 4.3* j s
- 4. Knowledge of system purpose and/or function. 3.8 .-3. 6 .
{ J 5, tKnowledge of limiting conditions for operations and safety j limits. 3.4. 4.2* 1
- 6. tKnowledge of bases in technical specifications for limiting conditions for operations and safety limits. 3.0* 4.l*- ;
'7. : Knowledge of purpose and function of major system components 3.8 'i and controls. 3. 8
(
- 8. . Knowledge of the annunciator alarms and indications, and use of the response instructions. 3.6 2. 2
' 9. Ability to locate and operate components, including local
. controls. 3. 8 3.5
- 10. Ability to explain and apply all system limits and prec a'ut ions. 3. 5 3.7 l
" 11. tAbility to recognize indications for system operating parameters.which are entry-level con'ditions for technical specifications. 3. 4 4.2* ,
- 12. Ability to verify system alarm setpoints and operate controls identified in the ' alarm response manual. 3.6 3.3
. 13' . Ability:to perform specific system and integrated plant 3.4 procedures during all. modes of operation. ' 3. 6
- 14. t _ Ability to perform witho'ut reference to procedures those actions that require immediate operation of. system components -
or: controls. 3.9*- 3.7*
,j s ,
q
-q
;K/A catalog:..BWR- 3.1-24 w - , - e
2.1 Conduct of Operations (continued) l 1 Ii 1 2.1.27 Knowledge of system purpose and or function. ] (CFR 41.7) IMPORTANCE RO 2.8 SRO 2.9 l 2.1.28 Knowledge of the purpose and function of nudor system components and controls. (CFR 41.7) i IMPORTANCE RO 3.2 SRO 3.3 2.1.29 Knowledge of how to conduct and verify valve lineups. , (CFR 41.10, 45.1, 45.12) l IMPORTANCE RO 3.4 SRO 3.3 j l 2.1.30 Ability to locate and operate components, including local controls. (CFR 41.7, 45.7) IMPORTANCE RO 3.9 SRO 3.4 I 2.1.31 Ability to locate control room switches, controls and indications and to detennine I that they are correctly reflecting the desired plant lineup. (CFR 45.12) IMPORTANCE RO 4.2 ' SRO 3.9 2.1.32 Ability to explain and apply system limits and precautions. (CFR 41.10, 43.2, 45.12) IMPORTANCE RO 3.4 SRO 3.8 i 2.1.33 Ability to recognize indications for system operating parameters which art entry-level conditions for enhalcal specifications. (CFR 43.2, 43.3, 45.3) l IMPORTANCE RO 3.4 SR0 4.0 l 2.1.34 Ability to maintain primary and secondary plant chemistry within allowable ; j limits.
- (CFR 41.10, 43.5, 45.12)
IMPORTANCE RO 2.3 SRO 2.9 l 4 4 4 1 I NURao-1123, Rev. I 2-4
- . _ _ _ _ __ ._ - I
i l l i SRO EXAMINATION QUESTION NO. 64 Station Power Restoration, EO-000-031, provides a specific sequence for reenergizing busses from an off-site source to AVOID:
- a. diesel generators tripping on overspeed when loads are transferred to off-site power.
- b. underfrequency condition on off-site sources due to manually reenergizing non-emergency busses.
- c. undervoltage condition caused when a ECCS initiation signal is present.
- d. starting equipment automatically without operator action.
Answer Key Choice c l
. \
Candidate's Choice d Basis Procedure EO-000-031 is performed to recover from a station blackout. The CAUTION in EO-000-031 after Step 2.1.1 ensures breakers have been aligned during the blackout in accordance with EO-100-030 and EO-200-030. These three procedures work together to address the concem as stated in the DISCUSSION Section (4.0) of EO-000-031. tO "The concem is if a low pressure ECCS initiation signal is present, simultaneous start of large ECCS motors will cause an undervoltage condition." Therefore, EO-000L 031 provides a specific sequence for re-energizing busses from an offsite source to avoid the simultaneous, automatic start of large ECCS motors which will result in an undervoltage condition, I contend that Choice "d' is the correct choice because the stated procedures ensure the simultaneous start of large ECCS motors is avoided, which prevents an undervoltage condition. Supportina Documentation EO-000-031; Rev.10; Pages 3,5, and 17
)
v 4
( , L . . PCAF #I -% o3& O 2 E0-000-031
- V PAGE 0 F._.p Revision 10
- Page 3 of 19
.: CONFIRM , 2.0 OPERATOR ACTIONS 2.1 GUIDELINES FOR CHOOSING OPERATOR ACTIONS l 2.1.1. If one or more ESS bus energized via associated Diesel Generator, PERFORM step 2.3. CAUTION BREAKER ALIGNMENTS IN E0-100-030 AND E0-200-030 MUST BE COMPLETED BEFORE PROCEEDING. 2.1. 2. If power available at Startup Transformer T-10, PERFORM step 2.4.
. 2.1.3 If power available at Startup Transformer T-20, PERFORM step 2.5.
1 e l i T w , , .,
PCAF #1 L ONT
, '~') PAGE S OF P E0-000-031
(
'" Revision 10 Page 5 of 19 CONFIPy 2.4 POWER AVAILABLE AT STARTUP TRANSFORMER T-10 2.4.1 PERFORM following to energize BUS 10:
- a. INSERT key and PLACE SU XFMR 10 TO BUS 10 SYNC SEL HS-00014 Keyswitch to ON.
- b. CLOSE SU XFMR 10 TO BUS 10 BKR OA10301 by placing switch to CLOSE.
- c. OBSERVE SU XFMR 10 TO BUS 10 BKR 0A10301 CLOSES. ,
- d. RETURN SU XFMR 10 TO BUS 10 SYNC SEL HS-00014 to 0FF and REMOVE key.
2.4.2 CLOSE SV BUS 10 TO XFMR 1010A10306 l to energize ESS XFMR 101 and bus 0A205. l (~N 2.4.3 CLOSE SV BUS 10 T0 XFMR 1110A10312 l
) to energize ESS XFMR 111 and bus OA206. j CAUTION .
EQUIPMENT MAY AUTO START IF INITIATION SIGNAL. PRESENT. I ADS MAY INITIATE WHEN RHR/CS PUMP (S) START. j 2.4.4 If de-energized, ENERGIZE busses, waiting approximately l 1 minute between each bus, by placing applicable control switches to OPEN position allowing' auto closure by . matching semaphores: '
- a. XFMR 101 TO BUS 1A BKR 1A20101
- b. XFMR 111 TO BUS IC BKR 1A20301
- c. XFMR 111 TO BUS IB BKR 1A20201
- d. XFMR 101 TO BUS 10 BKR 1A20401
- e. XFMR 101 TO BUS 2A BKR 2A20101
- f. XFMR 111 TO BUS 2C BKR 2A20301
,e m 9 XFMR 111 TO BUS 2B BKR 2A20201 I )
s_/ h. XFMR.101 TO 605 20 BKR 2A20401
_. _ _ _.. . _ . . . . . ~ . _ . . . _ _ _ . _ _ _ _ _ _ _ _ . _. _ . . _ l l E0-000-031 Revision 10 J Page 17 of Ig
]
l 4.0 ~ DISCUSSION
.f
\
This procedure.provides instructions for restoring AC power following a' ! station blackout of duration which does np.t exceed 125V DC station battery )
. capacity. Following are calculated 125V LC battery capacities: !
- a 10610 6.8 hr 30610 6.3 nr i j 1D620 6.4 hr 20620 5.9 hr -
1D630 IJ.2.hr 20630 11.3 hr 1D640 12.2 hr 2D640-10.8 hr E0-100'-030 and E0-200-030 ensure the station portable diesel generator, Blue
- Max, is connected to ID610, 10620, 2D610 and 2D620, thus extending their capacity indefinitely.
Station power restoration is accomplished in two (2), ways: (1) power available. from diesel generator (s); (2) power available from offsite. source. When )' restoring with diesel generator (s), restoration consists of loading available diesel generator (s) with those systems and components required to cool primary containment. No other acticas are required. However, when restoring from offsite source, restoration must be controlled to prevent an undervoltage O condition from occurring. EO-100-030 and EO-200-030 perform switching to prevent 13.8 KV aux bus and 4KV supply breakers from closing when SU XFMR 10(20) to BUS 10(20) BKR OA10301(OA10401) is closed. The concern is if a low pressure ECCS initiation signal is present, simultaneous start of large' ECCS 4 motorr will cause an undervoltage condition. I i 4 4 a
- j
- i ( _
)
S
- l
8 [ SRO EXAMINATION QUESTION NO. 66 Given the following: l
. A station blackout has occurred.
. MAIN STEAM SRV LEAKING is alarming.
. MAIN STEAM DIV 1 SRV OPEN is clear.
. MAIN STEAM DIV 2 SRV OPEN is clear.
Based on this information, what is the status of SRVs and equipment to monitor SRVs?
- a. an SRV is leaking. The acoustic monitors fail during a station blackout. J
- b. All SRVs are closed. Tailpipe temperature indications fail high during a station blackout.
- c. Status of the SRVs is unknown because the annunciators are indications of loss of power to instrumentation.
- d. An SRV has opened, then reclosed, causing the acoustic monitors to clear.
Answer Key Choice a Candidate's Choice b i Basis Based upon the given information in the question, the status of the SRVs is unknown. The " MAIN STEAM SRV LEAKING" alarm is initiated when the SRV tailpipe temperature exceeds 250" F. Therefore, this alarm could be the result of any one of the following situations:
- 1. The SRV is open.
- 2. The SRV has cycled and the tailpipe temperature is 250* F.
- 3. The temperature recorder from which the alarm originates has not been reset.
- 4. The SRV is leaking.
- 5. A LOCA has occurred and the containment temperature has increased to the point where the SRV tailpipe temperature sensor is exposed to a temperature above 250* F.
Since it is not possible to determine the exact status (open, closed or leaking) of the SRVs, Choice "a"is incorrect. ( Choices "b" and "c" are incorrect becau::e the temperature recorder is powered from an x inverter (1D240) which has a battery as an citernate power source.
. l l
Choice "d" is incorrect because the acoustic monitors are deenergized during a station
' blackout. They are powered from the ECCS 4.16 KV busses via instrument AC panels. I I contend there is no correct choice for this question, and the question should be deleted
- l t from the examination. !
' Supportino Documentation J AR-110-001; Rev. 5; Page 1g of 34; Window E01 , OP-157-003; Attechment A; Rev. 5; Page 18 4 4 i 1 5-s i 4 I I l l
l
- 4
{ AR-Il0-001 Revision S. . Page 19 of 34 3 E01
.NAIN STEAM SETPOINT: 250*F
! SRV
' t
-LEAKING ( E01 ) ORIGIN: TRS-821-IR614 -
2 i 1.0' ) j PROBABLE CAUSE: Pressure Relief (ADS or Safety) Valys leaking by seat or positioned 1 open. . ' , , I 3 2.0 OPERATOR ACTION.
- i. . ,
2.1 OBSERVE following 'on Panel IC601:
- 2.1.1 SRV OPEN PSV-F013 VI-14181A.
2.1.2 SRV OPEN PSV-F013 VI-141818. 2.2 OBSERVE SRV/ ADS Temperature TR-821-IR614 on Panel IC614 to determine relief valve indicating temperature increase in .
. discharge piping.
, 2.3 OBSERVE relief valve solenoid energized /deenergized status lights E at Panel IC601.
'2.4 If safety relief valve determined to be open, PERFORM ON-183-001 Stuck Open Safety-Relief Valve.
2.5 !~ COMPLY. with' Technical Spe,cification Section 3.4.2. L
- 3.0 AUTOMATIC ACTION:
4 None 4.0
REFERENCE:
~ ' 4.1 E-324 Sh'12 4.2 Mi-821-129(8) 4.3 IOM 305 O'
~
G
., 1
.. .- . j
. l
. l Attachment A j' OP-157-003 s Revision 5 Page 18 of 23 PREFERRED ALTERNATE i POWER SOURCE POWER SOURCE j l
1B236-61 !B216-74 l
) ICBE002 ICBE001 )
1X218 EXTERNAL MAINTENANCE : UPSBYPASSPANEL ID241 l i
- a tt , i 8i DIST PNL l
_L 2 ty21s
-- I s-r u.ao. m 3 --
for Comud Dr' - i UPSID240 PANEL l STAflCSHHCH i< - [ - , (x rr ACBYPA55 o gypyy 6)(CB-0) q 4) CRITM BUCAC l - ourPur (CB 3) RECTIFIU/BATTOY ACINPUT
'N#U '
b -- (CB.1) m eron . 250&DC Contact o) g BATTar (CB.2) BREAKER 1 2 3 BYPASS X E# " po o TEST X X M77a NORMAL X X An Tindicates contacts are closed CONTACTSAREKtKEBEFOREBREAK SINGLE LINE DIAGRAM of UPS 1D240 and EXTERNAL MAINTENANCE BYPASS PANEL 1D241 CONFIGURATION 5-- Page'l of 1 . 4
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nen w s w ss. - a n n Q nc.a d , s w.~ . -- r.-.. ( PROCEDURE COVER SHEET ( - { NUCLEAR DEPARTNENT PROCEDURE OPERATIONS POLICIES OP-AD-001 8 .) AND WORK PRACTICES Revision 9 7 Page 1 of 75 l l l i EFFECTIVE DATE: _ N - PERIODIC REVIEW FREQUENCY: O WA _
- PERIODIC REVIEW DUE DATE: 9/Go /9R i REVISED PERIODIC REVIEW DUE DATE- I PROCEDURE TYPE: QA Program (X ) YES (,__) No Plan't Procedure (Y ) YES (,__) NO REVIEW METHOD:
(_) Alternate (JL) Expedited PCAF'S: ( ) PORC , (_ 1-97-6045
) ERC 1-96-0965 o 1-96-0801 Prepared by. .D D Ua '
ate-llpd9'r Reviewedbi A44. O' Date /[J//i7
, s/ Supervisor Recomended Manacer - Nte' ear 00s Funct1onal Ln1 t Nanager Date See amolicable PCAF's NA Date _
PORC Comittee Meeting No. NA Date - ERC Committee Meeting No. Appraved by Bant Manaaer - Sqsouchanna
# Date See nonlicable PCAF's FORM NOAP-QA-0002-1, Rev.1, Page 1 of 1 O
A __ e _ _ _ _ _ _ _ _ _ _ _ _ _ -e
gen _3e-iw ~.13mo um wvcwmu~n * - ~-~ ' ' ~ - -
,l
. l l - (* ,
OP-AD-001 - i . Revision 9
. . Page 30 of 75 l 6.18 PROCEDURE -
COMPLIANCE . (5) 6.18.1 - Individual operators are responsible for controlling the
, , plant and keeping it within allowable limits et all-
' times. Procedures represent Management's expectations and bounds of authorization to operate plant systems and equipment. Procedures form the basis for which individual operator actions will be evaluated and judged for adequacy. Propedure compliance is our standard to I operate the plant safely and efficiently.' 6.18.2 controlling the plant within allowable limits requires a balance bf adherence to procedures and sound rating - ' JAMment. In order to maintain this balance ' fol owing must be considered: ,
- a. Attributes of Prc
- edures: .
. (1) Procedures are developed and.re' viewed in l advaisce. The developers and reviewers have
( time 'to consider carefully the. effects, of . each' step. - (2) Procedures are changed to reiflect lessons
. learned. Failure to use them may mean '.
failure to learn from previous mistakes. l ' (3) Procedures are step by step methods. Without them the individual must remember every - action, every precaution, and every effect - 4 somewhere else in the plant. - -
- ~
(4) Procedure steps assume no problems with the * {< maiterial' readiness of the equipse,nt/ systems. . . 4
- b. Attributes of Judgeme~n t (1) Judgement can take into account the ' current j ,
, circums,tance; procedures cannot.
' (2) Judgement can find and correct deficiencies in procedures. .
* (3) Judgementrequirestimetothinkaboutthe
- effects of what you're doing. Without time, judgement is hampered.
l' . . 4 0 ' t . .
%-3o-an w(t u m . nu urv ur acu e ow- r---
j OP-AD-001 Rhvision 9 Page 31 of 75 L 6.18.3 General Procedure Compliance ,' , ,
'(8) "
- a. If an existing procedure addresses the evoluNion.to
. be performed and the current circumstances, the , procedur,e shall be used. ,
If the exis' ting procedure is wrong, .it shall' be I '(') - b. l
. corrected prior to use. .. . - -
~
(') -
- c. A significant consideration in selecting the ' proper procedure to control an evoletion is khether all the precautions and prerequisites can be adequately ~
, met. If not the pro'cedum shs11 be changed prior touseor,nolused. ,
' ~
I d .- If ao procedure exists which addresses the- - i evolution and the current circumstances, the i following courses of action should be evaluated: Write.a procedure.to perform the evolution.
- (1)- ..
(2) Find another.means of accomplishing the same
. thing that is covered by a procedure. This
.could mean using another com
,N ' system, or another line-up. ponent, another " (3) To perform the evolution without a prof.adure
. or with exceptions to the procedure the statuscontroloftheevolutiontobe
. performed shall be maintained by one.of the j following authorities:- '
Pe'reit
- 4) i b) Statys Control Tag '
L *
.c 8ypass Tag l (d Check Off (CL) List
! Switching order - - l - PM Work 1.st activity WA (4) Any of the authorities in the preceding step ;
, may be used subject to the following - .
restrictions: ,
- ' (a) Shift Supervision has determined that
, the evolution can be. safely carried -
; out, and has approved 1,t.
l
< l
. . ., l
. 8 0
o*'*'
. r.eR-30-tw? 13:47 .
uwww. RI i.m-c
- e6 e
. . "**". l
(- - -
'0P- AD-001 .
l
', . Revision 9 !
Page 32 of 75 (b) Shift Supervision has detemined that
- no means exist.to do the evolution within the bounds of existing .
[
. . procedures. -
l .
-(c) Shift Supervision has detemined that the evolution is not too complex to j
perfom without procedure. .
, (d) Verbal communication is adequate for
'the task at hand.
- e. Is accordance with 10 CFR 50.54x,.a reasonable action that departs from.a license condition or a
- technical smeification may be taken in an' .
emergency wten this action is immediately needed to
' protect the public health and safety and no action -
' consistent with license conditions and technical
' specifications that can provide adequate or. * '
equivalent proter. tion is immediately ap
, Section 6.22, Technical Specifications, provides parent <.-
speci.fic requirement. . . f. Supervisors'/0perators are responsible for the nutcome of their judgements in perfomance of their job. .
~
\
i
' (1) Deviations from procedure place the entire
- responsibility of analyzing plant response on the Supervisor / Operator who chooses to -
~
. deviate. ,
. . .. l (2) Each Supervisor /0perator is expected to j
intelligently apply procedures to operate the plant.
- 9.' When Emergency'dperating Procedures are entered: .
(1) Reactor water level band shall be haintained
$13 inches to +54 inches unless' directed <
otherwise. (2) Reactor pressure band shall be maintained 800 .
. , psig'to 1087 psig unlessrdirected otherwise.
(3) Activities listed in Attachment 0 may be performed hhan required without reference to , ' f, , procedures. 1 ' l
. . \
. s O
i . _ _ _ _
-, - -rw-, -,n .e , , , - - _ . , , , . , , -, .,. , - , . , , -
uno. wurc t.r ma m>o o' " ' * ' ""
%. w2m is .
i . ( 5 op_AD-ool A
- Revision 9 .
Page 33.of 75 6.i8.4 . Alarm Response
- Procedure Compliance .
?
. a. AR and LA procedures provide information and direction regarding an abnormal plant parameter.
Specifically provide.d are alarm setpoint, probable cause, operator action and automatic action. . Since
. some actions may not be appropiciate'for the ,
- observed conditions, not all steps' are required to '
be performed. '
- b. AR and LA proceduNs are not' meant to take place of
, .. Off Normal
- . procedures,(ON) or Emergency Operating (EO)and '
- . operations procedure. ,
L c. For. unexpected alarms, the AR or LA procedure shall be referred to the first time it;is received each shift: . l , (1) To diredt action wheri time permits.' - ' 4
, ('
(2) To verify actions completed in all cases. i d . .. The procedure shall be available at thg work location for reference. 6 18.5 Event Based Emergency Operating'Precedure Compliance . ! a. ' If a conflict' exists between event based . procedures
, and symptos based E0P,s, symptom based E0Ps take precedence. - - - ' !
b.. - Addresses other than normal conditions which, if not corrected, could result in core damage or an uncontrolled release to the~ environment. . t
- c. E0P's shall be entered when events dictate'; they 1
shall no,t be entered of own accord. ' i d. Immediate' Operator Actions, per Attachment 8, shall be committed to memory. i,
- e. Immediate Operator. Actions are given in the preferred sequence, *but may be performed in any order. ' '.* * .
I 'I .
- f. ' Subsequent Operator Actions are given in a logical .
sequence and should be performed in that order, if - possible. 1
MW .%-1%7 13:4i1 I N . w8/8 e'. 8 e e 8* "8 OP-AD-001 i - - Revision 9-Page 34 of 75 i \
. . g. The procedure shall be present and continuously
{ . referred to as soon as practical after being initiated. 9
~
I h.
- Checked" : s are provided as a 'lacekeept a -
' only. CompetedEDsarenotretainedasrecoN's.i I
- 1. When the esorgency no lo er exists, the -
appropriate system operat ng procedure shocid be - ! entered.- - i 6.15 6 Symptos Based Emergency Operating Procedure (E0P) Compiiance 'I
~
. a. E0Ps implement eth' intent of the SSES -
) . Emergency Procedure Guideline and provide symp"ce-orfested technical direction .for
- emergency operating response. .
I
'b . '
If ' aconflict exists between event based procedures and sympton based E0Ps, symptom *' based j - E0Ps shall taka' precedence. ,
. . . 1
- c. Entry conditions are synstomatic of both. !
emergencies and everts wtich may degrade into j emergencies. *EOPs specify actions app'ropHate for both. Therefore, entr.y into EDPs is not cpnclusive that an energency has occurred. .
- d. Entry conditions shall be committed 'to memory.
- a. -
E0ps s6a11 be entered when an entry condition !~ . exists or when directed.by an off Normal (ON) procedure; E0Ps shall not be entered of own accord. .
- f. 20P flowchart shall be present 'and continuously
, referred to while being executed. ,.
. g. Entire ' step shall be read since steps maiy 1
.contain multi ple conditions and actions.
h. Logic ters 'when' acts as 'a stop sign. and' - subsecuent actions may not.be performed imtil . statec conditions are met. These statements are l , identified wi,th a red stop sign. .
^
. I . 9 "d m-A y.,
A_ _ 3o_19yi_ 23ms ___ uuA iwi.n.v neu ~o ,
~
, a- .
.' a-f.
- OP-AD-001 , i Revision 9 l Page 35 of.75 !
'4 ' l
. i. Logic term "if* prohibits acti ns of that step '
(including substeps) from being ' performed until stated conditions are met. If identified . conditions are not met, execution of flowchart - shoeld proceed promptly in accordance with followingstep(s). 1 . , j.. ' Logic ters 'before* requires action any time .
; ,- pr'or to and as soon as possible if the action level is reached or exceeded. Utilized where an
. event-independent margin is not appropriate or .
, cannot be defined. l L . .. . .
L a ..
~
- k. Command steps specify actions to be performed. -
)
i
- They any be performed in parallel, if .
appropriate. - j l .
- 1. Awareness ste>s include parameter control guidance whici
- overrides' subsequent guidance i
d enever its conditions are met. They are . i applicable to all subsequent steps within a .
)
floup6th. They are applicable whether,the steps . were entered from above,.or transfer resulted in entry below the steps. They remain applicable - to those steps until the flowchart is exited.- 1 I * '
, . m. Steps containing lists, e.g., methods of -
alternate RPV injection, do ad imply priority based on. order of list. Systems shall be . selected based on plant . conditions,. not b.ased on order'of list.
- n. Decision steps usually ask a' question concerning
.a plant parameter or operating condition.
- o. Cautions warn the operator that a condi ion may-
' existthatmayaffectsystem/ equipment - operabilit'ywhile performing subsequent step (s). Caution symbols are placed <n a shadowed .
~
J rectangle. Where space on the E0P flowchart !
- in the flowpath
> receding the step (s)large or tooa to which t apply. Wien cautions are too numerous to place in the flowpath, they are
. referenced by nueber.in a shadowed rectangle attached to the step. The number in the shadowed rectangle corresponds to tha. caution
( - 3
- with' that . number in the list of cautions.
4
. e
.g 4
~^
en- s- 1,<~ c 4 ~~ ~ ~ uanc m,u-c ce ne J. c..aTa,. F .cs ..
. .~ . .
OP-AD-001-I - Revision 9 0 , Page 36 of 75 p '. . Entry / Exit arrows identify transition points between two flowcharts; they are not used when
, . transferring within the same procedure. Entry arrows are enlarged to speed transition between flowpaths. Inside the entry arrow is the step
' . number to which the' transition leads. The exit
, arrows are not enlarged. .Inside the exit arrow '
is the flowchart /procedum number to. transfer-
' to. The specific step on the next flowchart is
' identified under the exit aitow.
- q. ' Concurrent exit and continue arrows indicate a point where two flowpaths are required to be. .
executed concurrently. ..
- r. Alam symbol alerts the operator to review RPV pressure ore control awareness stepswhich 0-100-113) spondinghave the corn (in i
alam symbol. In place of _a concurrent arrow, the alam symbol is used. When requirement for rapid depressurization is reached in a flowpeth, '
/ ' '" the phrase " RAPID DEPRESS REQ'D" appears.
t However, if 6 SRV's are already open, it is 4
' incorrect to enter EO-100-112 RAPID -
' DEPRESSURIZATION.' If less than 6 SRV's are . open, the awareness step directs the operator to
. depressurize in accordance with E0-100-112.
' L. Subentry conditions' identify multiple optional flowpaths available for implementation with no specific priority. ,
' ' t. AVAILABLE - the state or condition of being ready and able to be used (placed into ~ '
operation
- action or) function.toAs accomplish applied to the stated (or implied) a system -
this requires the operability of necessary , support systems (electrical power supplies. . cooling water, lubrication, etc.). '
- u. CAN/CANNOT BE DETERMINED - the current value or
' status of an identified parameter relative to that specified in the procedure can/cannot be ascertained using all available indications (di'ectr and indirect, si,ngly or in combination).
I 4 9
_ _ _ . _ _ _on-
#& was ax u m A v.rc c.c nr.a e E""
i . . . i - ' f DP-AD-001 .
. Revision 9
' 'page 37 of 75 3
- v. CAN/CANNOT BE MAINTAINED AB0VE/8ELOW - the value of the identified parameter (s be kept above/beTow specified) limits.is/is This not able to i
determination includes making an evaluation that
, considers both corrent and* future system-
! ' performance in relation.to the current value and
.', trend of the parameter rosarded, "cannot" does(s). As specifically not imply that the i
actual value of the parameter must first pass j ' the specified limit. e ' w'. CAN/CAle107 BE RESTORED A80VE/8ELOW - the value . of the identified parameter s is/is not able to be returned to above/below.s(pe)cified limits .
~
1 after havihg passed those limits. This determination includes making an evaluation that . can'siders both' current and future system ' performance in. relation to the current value and trend of the parameter s). specific time interval (but dods not permit .Does not prolonged operation'beyond a limit without l ( taking specific action. . j ,
' x. ' Key limits and action levels have been bolded and enlarged ,for ease. of readability. '
- y. Checkoff boxes provide a space to document that <
the step hu been read and addressed. The box '
- appears to the lower left of the step.
- z. '. operator!shall document flowcha'rt status to maintain placekeeping and ensure a satisfactory
- l , ,
turnov,er can be. accomplished, ' ' an. ' E0> shall be exited when it can be determined . - positively either an emergency no longer exists (i.e.,. problem is known and under plant control) for those parameters controlled by this ! flowchart rg the exit / transfer condition specified 'n the procedure is satisfied. After an E0P has been entered, subsequent clearing of 1 . all entry conditions is AS1 by itself " conclusive that an emergency, no,1onger, exist's. 6.18.7 Emergency Support (ES) Procedure Compliance - \
- i .
'a. These procedures bypass plant interlocks or l .I operate equipment by alternate methods. ,
1 L _
. - ~ _ _-. . _ _ _ _ - . . _ _ .-
ys-30-1997 13 50 usrAC WWC W s m48 " "1 "' "*l
. OP-AD-001
- I Revision g Page.38 of 75
. b. They shall be implemented oniy when directed by an E0 or another ES precedure and under -
j direction of Shift Supervisor.
- c. Sections of procedures not required to'be
- performed shall ha.ve N/A entered.
- d. Procedures shall be present and followed' step by step while being executed. .
- e. lissan Factors Engineering hai designated the color.' green' to aid in dentifying. components which must be located to perform ES procedures. -
controls and indications in the control room ! which are familiar to the operator and can be i . readily located are not color coded.. - 6.18.8 ' GeneralOperating(G0)Procedurecompliance 2
. . , a. Go procedures provhde direction' fo'r completin; soda changes or other complex evolutions whica 7
~ generally take place over relatively long periods of time and require one or more of the '
n following: .,
'(1) surveillance.s.to be performed.
t i
.(2) Multiple systems to be placed in and/or out of service.
~
(3) , Logs or other rec ^ ords to be kept..
- b. 'If a conflict exists between General Operating .
'. Jrocedures and symptom based E0ps, symptom based J.0Ps take precedence. -
- c. Due to their complexity and infrequent use, G0s shall be present and continuously referred to while being perfonned. ..
d.. Steps are written in approximate order of
- performance, but Shift Supervision may determine steps to be perfonsed concurrently or out of -
sequence based on current plant conditions.
- s. . Prior to performing steps that change reactor modes, uncompleted steps shall be evaluated for mode change imp'act.
~ iOTdL P.11
/^\
U PROCEDURE COVER SHEET 4 NUCLEAR DEPARTMENT PROCEDURE
'\h E0-100-100 l
c " CAUTIONS Revision 6 f o Page 1 of 9 EFFECTIVE DATE: 6-~/6- 7 PERIODIC REVIEW FREQUENCY: f F S A-PERIODIC REVIEW DUE DATE: Io 30!97 REVISED PERIODIC REVIEW DUE DATE: .
/
( ,)
\
PROCEDURE TYPE: QA Program ( / ) YES ( ) N0 w/ P1 ant Procedure (t/) YES ( ) NO REVIEW METHOD: ([) Alternate ( ) Expedited ( ) PORC ( ) ERC Prepared by NMh ' Date 4 I9l I Reviewed by ~ Date V 20' 6 i supervisor . Recommended x b" ^ -- ^ Date y/ze / f (# Functional Unit Manager Nk Date PORC Committee Meeting No. WN Date ERC Committee Meeting No. fS Approved by ,A 7 1 [upd' Date f"+/-/F FORMNDAP-QA-00$2-1,Rev.1,Page1of1
1 4 EO-100-100 ^ ( Revision 6 ,
-( Page 5 of 9 , j
] 4 4 Caution 4
~
RAPID INCREASE IN INJECTION MAY INDUCE ! LARGE PWR EXCURSION AND SUBSTANTIAL j CORE DAMAGE j j This ~ caution appears in steps that control. injection into a critical core. Two problems exist. First, addition of cold, unborated water l s will result in an increase in positive reactivity. If this effect is ,
- combined with rapid injection that sweeps boron from core region, i resulting power excursion may be sufficiently large to damage the core..
The second problem is rehooding a partially uncovered, non-shutdown l i core. The rate of positive reactivity addition from an unmoderated core is quite large. Power will increase rapidly, eventually producing more steam than open SRV's can relieve. Resulting pressure increase will-j- collapse steam voids, creating additional positive reactivity.. Rapid injection into a non-shutdown core must be avoided. Whenever injection.into a non-shutdown core must be increased,-it must be increased slowly in order to avoid a large power excursion and substantial core damage. Fuel clad failure is delayed for several minutes with a completely uncovered core. This will be extended 1- considerably with partial submergence and' steam flow past fuel bundles. g (
Reference:
SSES-EPG Caution 7) Caution 9 LOSS OF I-A CAUSES LOSS OF NORMAL FW INJECTION Auxiliary bus. load shed causes loss of instrument air. Upon loss of ' instrument air, Feedwater low Load and Low Load Bypass valves fail n closed; and Condensate Recirc and RFP Recirc Flow valves fail open, thereby eliminating normal feedwater injection and complicating level
- recovery. In addition, CRD Flow Control Valves fail closed reducing CRD flow to approximately 20 GPM, by design.
e Purpose of this' caution is to alert the operator of the undesirable effects from failure to promptly recover from loss of instrument air. (
Reference:
SSES-EPG Caution #9) (
.y t
m.,,,.-,_.__..m ,,.,_u.y.- _., m, , , ., , , , . , , , , , ,, -
1 j 7q PROCEDURE COVER SHEET. f ( )
- \J N NUCLEAR DEPARTMENT PROCEDURE
')-
SECONDARY CONTAINMENT CONTROL E0-100-104 Revision 8 E o Page 1 of 28 Ers # 1 EFFECTIVE DATE: M o77-98 2 W4d
~
PERIODIC REVIEW FREQUENCY: PERIODIC REVIEW DUE DATE: 6 30 97 l l REVISED PERIODIC REVIEW DUE DATE: , ) l PROCEDURE TYPE: QAProgram(/)YES ( ) NO Plant Procedure (/) YES ( ) N0 l':. REVIEW METHOD: ( [)' Alternate ( ) Expedited ( ) PORC ( ) ERC Prepared by YY7d b ' Date 4f/9f6 Reviewed by ~ D l ~ Date k' '9f I Supervisor Recommended e -- ' :- =, Date V/2/ '
'l J Functional Unit Manager M Date PORC Committee Meeting No.
Date ERC Committee Meeting No. Approved by , N y [ _2
- Date f #~ A'-f f -
FORMNDAP-QA-0002-1,Rev.1,Pade1of1
. (v) .
k
> 1 E0-100-104 :
- Revision 8 t Page 2 of 28 :
i 1.0- ) GENERAL 2 Purpose e' this guideline is to: l
; * . Protect equipment in. secondary containment I
- . Limit' radioactivity release to. secondary . containment, and either ,
. ._ Maintain secondary containment integrity, or I
- . Limit radioactivity release from secondary containment .
Secondary Containment Control establishes and maintains control over
+ three key secondary containment parameters: area temperatures, area ,
radiation levels and area water levels. Operator actions are performed . 'l concurrently to stabilize and control these parameters.
]
I Normal systems' and methods are used to maintain secondary containment
- parameters at or below maximum normal operating values. If a parameter exceeds its Max Normal operating value, action is taken to isolate primary systems discharging into secondary containment except those systems required to shut down reactor, assure adequate core cooling,
- ,... suppress a fire, or prevent primary containment failure. Actions taken
'* above the Max Normal operating value are dependent on determining if the parameter is-elevated as a result of a primary system discharging into i- Secondary Containment " areas" as defined in this procedure. ; In accordance with OP-AD-001,: Operations Shift Policies and Work Practices, this procedure is exited when it can be positively determined , an emergency no longer exists (i.e., problem is known and under plant l control). ; l l I o l 0 k 5
'et ---_-.-O---_---l--_---
E0-100-104 (mw/
) Revision 8 Page 3 of 28 i
2.0 ERIB1 Entry conditions for this procedure are any of following: ,
. AREA TEMP ABOVE MAX NORMAL
. AREA AT AB0VE MAX NORMAL
. ZONE 3 HVAC EXHAUST RAD > 2.5 MR/HR
. SC SPING RELEASE RATE AB0VE MAX NORMAL l l
4
. UNEXPLAINED AREA RAD LVL AB0VE MAX NORMAL )
l
. ZONE 1 OR 3 < 0.25" WG VACUUM FOR 4 HRS l
~
. AREA WATER LVL AB0VE MAX NORMAL 4
Conditions which require entry into Secondary Containment control are symptomatic of an emergency condition ~ or conditions which, if not corrected, could degrade into an emergency. Entry condition setpoints
' [- were chosen to provide advance warning of potential emergency
( conditions, allowing action to be taken where such action may be , w successful in preventing otherwise more severe consequences. 1 Max Normal operating temperatures have been defined as Technical Specification'high temperature or high differential temperature isolation setpoints of Leakage Detection System; or the maximum temperatures expected in secondary containment areas normally accessed by personnel, whichever is lowest. l Zone 3 exhaust > 2.5 mR/hr may indicate that radioactivity is being j released to the environment when the system should have automatically , isolated. Its value is consistent with Technical Specification LC0 for l this isolation function which is established to mitigate the l
' consequences of accidents. Zone 3 exhaust can be monitored at 10600: l RR-D12-1R605, Refuel Floor Wall Exh Radiation Monitor RR-D12-IR607, Refuel Floor High Exh Radiation Monitor
.RR-D12-1R608, RB Zone 3 Exhaust Railroad Access Shaft Radiation Monitor 4
2 Q
i e E0-100-104
'l Revision 8 O Page 4 of 28 Secondary Containment SPING release rate Max Normal is defined for the Reactor Building and SGTS ventilation stacks as 1/5 of the site release rate LCO value. The Stack Monitor System, SPING, Hi-Hi alarm setpoint is 1/5 of the site release rate LCO value, therefore, entry into this procedure is required when a valid Hi-Hi alarm occurs for the reactor j building or SGTS stack monitor SPING with the following exception. The i Primary Containment vent flowpath is through SGTS, therefore, it is possible for a SPING alarm to be indicative of activity which is originating from Primary Containment. For this reason, entry into this procedure is D91 required if a SGTS Hi-Hi alarm occurs while venting Primary Containment with HVAC in normal mode of operation. In )
accordance with CH-IC-016, Calibration of the Eberline SPING Monitors, l setpoints for Hi-Hi alarm are: ! Particulate Iodine Noble Gas (uCi/cc) (uCi/cci (uCi/cc) Unit 1 RB: +3.24E-08 +5.90E-09 +3.56E-05 , SGTS: +3.06E-07 +5.57E-08 +3.37E-04 C A radiation level above Max Normal may be indicative that water or steam ) ( ' from a 1rimary system (or from a primary to secondary system leak) may be disclarging into the secondary containment. Max Normal operating l radiation levels are equal to ARM high alarm setpoints. The term " unexplained" is used here to prevent unnecessary entry if the cause of the area radiation level is known to not be a threat to secondary containment. An example of this would be work activities on ! the refueling floor when radioactive components are moved past ARMS. Secondary containment " areas" with respect to radiation considerations , are defined in Table 9 where they are listed and separated with ' horizontal lines. These " areas" are physically separated from one another and contain area radiation monitoring instrumentation appropriate for use in detecting a primary system or primary to 4 secondary system leak. Some areas containing ARMS are specifically excluded from Table 9: TIP room; CRD repair room; new fuel area; and recirc fan plenum. Elevated readings in these areas would not normally be due to secondary containment problems. Actions in this procedure to shut down the reactor and rapidly depressurize the RPV would not be appropriate in response to high radiation levels in these areas. Zone 1 or Zone 3 differential pressure < 0.25" WG Vacuum is indicative o of a potential loss of reactor building structural integrity and could (d i result in uncontrolled release of radioactivity to the environment, Consistent with Technical Specification LCO, a four (4) hour qualifier is added to allow off-normal (ON) and system operating (0P) procedures 1 to be utilized in attempting to restore secondary containment HVAC.
ex E0-100-104 (d i Revision 8 Page 17 of 28 i SC/R - Secondary Containment Radiation SC/R-1 .W HEN ANY AREA RAD EXCEEDS MAX NORMAL ISOLATE ALL SYSTEMS DISCHARGING INTO AREA EXCEPT SYSTEMS REQ'D T0:
. SHUT DOWN RX ,
. ASSURE ADEQUATE CORE COOLING -
. SUPPRESS A FIRE
. PREVENT PC FAILURE l l
TABLE 9 SECONDARY CONTAINMENT RADIATION ARM NUMBER MAX NORMAL MAX SAFE ) 1 LO HIGH ARM CHANNEL FIELD FIELD PMS 89 AREA RANGE RANGE DESCRIPTION (MR/HR) (MR/HR) (R/HR) .: B18 FT EL 35+ CASK STOR AREA HI ALARM 10' 10 14+ OR 47* 49 SPENT FUEL P0OL 15+ OR 42+ REFUEL FLOOR AREA 749 FT El 8+ 52 RWCU RECIRC PP ACCESS HI ALARM 10' 10 10* 54 FUEL P0OL PP ROOM 719 FT EL 5* 50 CRD NORTH HI ALARM 10' 10 6* 51 CRD SOUTH 670 FT EL 16+ 53 REMOTE SHUTDOWN ROOM HI ALARM. 10' 10 3+ 48 HPCI PP*TURB ROOM 2+ 57 RCIC PP*TURB ROOM . 25* 55 RHR A*C PP ROOM I 645 FT EL 1* 56 RHR B*D PP ROOM HI ALARM 10' 10 l 4* N/A RB SUMP ROOM i I 2 RANGE: + 0.01 - 10 MR/HR i i 3
- 0.1 - 10 MR/HR )
O i i l
- .- . - . . . - . - - . - - - _ . - - = - - - - - . - - . . - -
E0-100-104 O Revision 8 Page 18 of 28
+
i i Secondary containment " areas" with respect to radiation considerations are defined in Table 9 where they are listed and-separated with horizontal lines. These " areas" are Reactor. '
- . Building elevations that contain area radiation monitoring .
instrumentation appropriate for use in detecting a primary system >
; or primary to secondary system leak. ,
Based on HVAC. design, and isolation features, an airborne radiation problem on one elevation would not normally be expected
- to spread to another elevation. Normal HVAC design maintains air 3 flow from areas of lesser to areas- of greater potential contamination.. If HVAC were isolated, airborne radioactivity from ,
la. primary system break would be contained within high energy rooms ! by a system high temperature /high differential temperature- i isolation, or backdraft isolation dampers. If airborne radiation were to exceed Max Safe in two or more areas, this occurrence would be clearly indicative of a spreading problem which could significantly threaten secondary containment integrity, and. ECCS equipment therein. Therefore, subsequent actions which shutdown the reactor, or scram and rapidly depressurize the RPV (if a primary system is discharging into secondary containment) are i O appropriate when Max Safe radiation is exceeded in two or more areas. Even though RCIC and ECCS rooms are physically separated from each other, they are grouped together as a single area to prevent an unnecessary rapid depressurization. In the event in which fission products are released to the primary coolant, operation of RCIC, . ECCS, or suppression pool cooling will circulate these fission products throughout system piping. Since this piping is located < throughout the 645' elevation, it is expected that the entire 645' elevation would experience radiation exposure from shine. If ; individual RCIC and ECCS areas were defined as separate areas, l
- such an event would unnecessarily require Rapid Depressurization.
Action is taken to isolate systems that are discharging into secondary containment to terminate possible sources of radioactivity release. Minimizing radioactive release to secondary containment alsc, helps accomplish the objective of i~ precluding a radioactive release outside secondary containment under conditions where secondary containment integrity cannot be 4 maintained. Fire suppression systems are nol sources of radioactivity, hence they are nat isolated. Exception.is taken to' isolation of systems required to shut down Og reactor, assure adequate core cooling, or vent primary containment because~ isolation of these systems ultimately results in a greater threat to overall plant safety.
. , , , . ey.w-.. -p% -
l 4 E0-100-104 aA j Revision 8-D Page 19 of 28 a Examples of-systems whose operation-is needed to prevent primary containment failure could be:
. Operation of the suppression chamber vent path irrespective of offsite release.
1 . Operation of the drywell vent path irrespective of offsite release.
. Operation of the suppression chamber sprays irrespective of adequate core cooling.
.. Operation of the drywell sprays irrespective of adequate core cooling.
(
Reference:
SSES-EPG SC/R, SC/R-1) SC/R-2 PERFORM CONCURRENTLY The remaining steps of this flowpath provide instructions to shut down the reactor, or scram and rapidly depressurize the reactor based upon the source of heat addition to secondary containment. ~ i If the source of radiation and/or radioactivity release is other than from a primary system discharging, SC/R-3 provides the appropriate operator actions. If a primary system is discharging O.. w into secondary containment, SC/R-4 through SC/R-6 provide the appropriate actions. Since it may not be possible to ascertain the reason for the radiation increase above Max Normal at the time this step is reached, both flowpaths are performed in parallel.
. i
(
Reference:
SSES EPG Execute steps SC/R-2 and SC/R-3 concurrently)
. SC/R-3 WHEN AREA RAD EXCEEDS MAX SAFE IN 2 OR MORE AREAS SHUT DOWN RX IAW G0-100-004
!- Should secondary containment area radiation levels continue to ! L increase and exceed their Max Safe values in more than one area without a primary system discharging into secondary containment, l it is prudent to begin a normal reactor shutdown in accordance with GO-100-004, Plant Shutdown to Minimum Power. The rapid 4 energy reduction obtained by scramming the reactor will not have l an effect on the energy which is causing the increasing radiation i
' level trend.
- The criteria of'"2 or more areas" identifies the increase in ;
radiation level trend as a wide spread problem which may pose a l threat to continued safe operation of the plant. '
-(
Reference:
SSES-EPG SC/R-3)
li < , ,
-~s. E0-100-104 +
(d l Revision 8 Page 20 of 28 SC/R-4 ' WHEN -A PRIMARY SYSTEM IS DISCHARGING INTO A SC AREA ~ ' ' CONTINUE- j A primary system is defined to be the pipes, valves, and other i equipment which connect directly to the RPV such that'a reduction -l i i . in RPV pressure will cause a decrease in the steam or water being '" discharged through an unisolated break in the system. - Subsequent 3 actions to mitigate the consequences of an elevated area radiation level are dagendent on identifying its source. Since radiation
- indication could be caused by " shine", the decision of whether or not, "a prinary system is discharging into secondary containment"
- . cannot be made based on remote radiation indication, alone.
An " area", for purposes of answering this question, is any secondary containment area defined in Table 9. 1 (
Reference:
SSES-EPG SC/R-2) i SC/R BEFORE ANY AREA RAD REACHES MAX SAFE ks. GO TO RPV CONTROL-The Max Safe operating radiation level is the most limiting area radiation level which will ensure personnel exposure is kept below the emergency exposure limit (25 Rem) while performing E0P actions
- in the secondary containment for a period no longer than 2.5 hours (i.e., 25 Rem /2.5 hr - 10 Rem /hr).
A reactor scram through entry to E0-100-102, RPV Control, promptly reduces to decay heat levels the energy that the RPV may be , discharging to the secondary containment. The instruction to'take this action at any time between the Max Normal and the Max Safe > operating value may help avoid reaching the more severe action of rapidly depressurizing the RPV. , i (
Reference:
SSES-EPG SC/R-2.1) 1 . I 4 a f
b t (
- v_a u w, - ., , .4 -
I
l E0-100-104 (A
'ss)
Revision 8 Page 21 of 28 SC/R-6 WHEN AREA RAD EXCEEDS MAX SAFE . -IN 2 OR MORE AREAS RAPID DEPRESS IS REQ'D ! Should secondary containment area radiation levels continue to increase to their Max Safe values in more than one area with a primary system discharging into secondary containment, the RPV must be rapidly depressurized. Depressurizing the RPV promptly places the primary system in its lowest possible energy state, rejects heat to the suppression pool in preference to outside the containment, and reduces.the driving head and flow of primary systems that are unisolated and discharging into-the secondary containment. The criteria of "2 or more areas" identifies the increase in radiation level trend as a wide spread problem which may pose a direct and immediate threat to secondary containment integrity or continued safe operation of the plant. The alarm statement " Rapid Depress is Req'd" alerts the operator that RPV pressure control overrides containing the alarm light ( symbol must be reviewed. 'w (
Reference:
SSES-EPG SC/R-2.2) J re-- c , -
/' N PROCEDURE COVER SHEET
~
MN NUCLEAR DEPARTMENT PROCEDURE
- RAPID DEPRESSURIZATION E0-100-112
' " Revision 6 Y o Page 1 of 15 EFFECTIVE DATE: 6'/6 -9 PERIODIC REVIEW FREQUENCY: 8 >'#/14.
PERIODIC REVIEW DUE DATE: Io 30 #/ 7 i i REVISED PERIODIC REVIEW DUE DATE: i PROCEDURE TYPE: QAProgram(/)YES ( ) NO ( Plant Procedure (/) YES ( ) NO REVIEW METHOD: ( d Alternate ( ) Expedited ( ) PORC (_ ) ERC j Prepared by WN h' . Date 4!/9!9f~ Reviewed by h 31 Date T 4 -N I Supervisor ~ Recomended M Y- CA Date Y/2/ f4J' Functional Unit Manager NO Date PORC Comittee Meeting No. Mk Date ERC Comittee Meeting No. Approved by A 1'ya Ms-md Date F-#/~/ r r / . FORM NDAP-QA-0002-1, Rev. 1, Page 1 of 1
. \
E0-100-112 Revision 6 Page 2 of 15 O 1.0 EEB&L Reduces RPV pressure as quickly and safely as possible to place reactor in its lowest energy state. This includes assessing plant condit*ons to ' determine optimal vent path. Actions in this procedure are required to-i
. Establish or maintain adequate core cooling
. Stop or minimize discharge of reactor coolant from unisolable primary system breaks Reduce energy in the RPV before reaching conditions for which the pressure suporession system may not accommodate SRV operation or a l loss of coolant accident ;
i
- Minimize radicactivity release from the RPV to the primary containment and areas beyond the primary and secondary containments.
2.0 ENTRY Attachment A provides a list of conditions that require rapid ! O, depressurization. This procedure is only entered after a reactor scram has been initiated. 3.0 PROCEDURE RD-1 EXIT RPV PRESSURE CONTROL RC/P or LQ/P When directed to rapidly depressorize, RPV pressure guidance given in this procedure supersedes such instruction given in E0-100-102 or E0-100-113. Therefore, the RPV pressure control sections of these procedures are not applicable and must be exited. (
Reference:
SSES-EPG override before RC/P-1)
E0-100-112 Revision 6 p Page 3 of 15 RD-2 PREVENT UNCONTROLLED COND INJECTION . EXCEPT AS REQ'D TO ASSURE l ADEQUATE CORE COOLING This step is applicable as long as it does not conflict with restoring or assuring adequate core cooling. If the RPV is depressurized to less than 600 psig without preventing uncontrolled condensate injection then an uncontrolled flood of the RPV will take place as condensate water injects through the feed pumps into the reactor. (
Reference:
SSES-EPG C2-1.1) 1 RD-3 HAS IT BEEN DETERMINED RX WILL REMAIN S/D UNDER ALL CONDITIONS W/0 B0RON If reactor shutdown margin cannot be assured, possible injection of large volumes of cold, unborated water into the RPV during rapid depressurization may result in serious core damage. 1 Therefore, precautionary steps regarding control of RPV injection must be performed if the reactor shutdown margin determination [7 cannot be made. i Positive confirmation that the reactor will remain shutdown under . all conditions is best obtained by observing that all control rods ! are full in. Criteria other than all control rods inserted to the full in position may also be used. These include:
. All control rods but one full in
. Evaluation of shutdown margin The phrase "...under all conditions without boron" requires that the shutdown margin determination disregard any negative reactivity contribution due to injection of sodium pentaborate solution. A borated core may not respond to temperature changes with a negative coefficient of reactivity. Reactor shutdown marain must be based on control rod position alone so that subsequent changes in boron concentration and moderator temperature do not return the core to criticality.
(
Reference:
SSES-EPG C2-1) O
l
-E0-100-112 Revision 6
'N Page 4 of 15 RD-4 IF ADEQUATE CORE COOLING IS ASSURED AND
'HI DW PRESS ECCS INITIATION SIGNAL 1.72 PSIG EXISTS PREVENT INJECTION FROM LPCI AND CS PUMPS NOT RE( D TO ASSURE ADEQUATE CORE COOLING Subsequent actions reduce RPV pressure below the shutoff head of Core Spray and RHR pumps which may have an automatic start signal due to RPV pressure below 436 psig in conjunction with a high drywell pressure condition. If injection from these pumps is not required to assure adequate core cooling, preventing their operation is appropriate since uncontrolled injection only complicates actions to maintain control of RPV water level.
Even when adequate core cooling is assured, it is permissible to keep at least one loop of ECCS pumps running until ADS is. , initiated as long as RPV Flood-up above +54" is prevented. ! i Post-blowdown use of Core Spray and RHR is permitted as dictated by adequate core cooling requirements. (
Reference:
SSES-EPG C2-1.1) C 4 'Y a . r y - .,
- y
E0-100-112 i Revision 6 p Page 5 of 15 RD WHEN ALL RPV INJECTION IS STOPPED AND PREVENTED
, IAW E0-100-113 LQ/L-19
.QB IAW E0-100-114 RF-13
-CONTINUE Since control rod insertion alone may not maintain the reactor shutdown, cautionary measures regarding RPV injection are taken to minimize the possibility of a power excursion. Failure to do so may. result'in rapid injection of a large quantity of cold, unborated water from low pressure systems as RPV pressure decreases to and below the shutoff heads of these pumps. Such an
- occurrence could quickly dilute in-core boron concentration and
- reduce water temperature in the core region. Sufficient positive
~ reactivity might be added in this way to induce a reactor power excursion large enough to severely damage the core. Therefore, if it cannot be determined that the reactor will remain shutdown
- under all conditions without boron, action to depressurize the RPV
- waits until it is confirmed that injection into the RPV is stopped and prevented as required by water level control procedures. -
This step does not direct the action to stop and prevent i injection. RPV injection is a water level control action which is bemg performed in parallel with this procedure in either EO-100-113 or E0-100-114. When the condition of t.his step is satisfied, the operator can proceed to step RD-6. (
Reference:
SSES-EPG C2-1) 4 CAUTION i
- C00LDOWN > 100*F/HR l MAY BE REQ'D
! This caution is applicable throughout this flowpath. Subsequent actions for controlling RPV nressure will result in cooldown rates greater than those allowed Ly Tech Specs but performance of these . c actions takes precedence over abiding by the RPV Cooldown Rate l LCO. , (
Reference:
'SSES-EP6 Caution #6) i O
4
Attachment A l E0-100-112 ((~}j Revision 6 Page 13 of 15 l CONDITIONS RE0VIRING RAPID DEPRESSURIZATION l I. RPV Control (E0-100-102)
- 1. With RPV pressure above 125 psig and a source of injection, RPV water level drops to TAF.
- 2. With'RPV pressure above 125 psig and no source of RPV injection, RPV water level drops to -205 inches.
- 3. With RPV pressure below 125 psig, RPV water level cannot be restored and maintained above TAF.
II. Primary Containment Control (E0-100-103)
- 1. Suppression pool temperature and RPV pressure cannot be maintained below Heat Capacity Temperature Limit. (Below HCTL and <5% PWR)
- 2. Suppression pool water level and RPV pressure cannot be restored and maintained below SRV Tail Pipe Level Limit (SRVTPLL).
O) V
- 3. Suppression pool water level cannot be maintained above the Heat Capacity Level Limit. (Above HCLL and <5%,PWR)
- 4. Suppression pool water level is <12 feet.
- 5. Suppression chamber pressure cannot be maintained below Pressure l Suppression Limit. (Below PSL and <5% PWR)
- 6. Drywell temperature cannot be maintained below 340*F.
i Ill. Secondary Containment Control (E0-100-104)
- 1. Two or more area temperatures exceed maximum safe with a primary system discharging.
- 2. Two or more area radiation levels exceed maximum safe with a primary system discharging.
- 3. Two or more area water levels exceed maximum safe with a primary system discharging.
Page 1 of 2
,_ PROCEDURE COVER SHEET
*N NUCLEAR DEPARTMENT PROCEDURE
- E0-100-113 LEVEL / POWER CONTROL Revision 7 o Page 1 of 68 EFFECTIVE DATE: A7-fI PERIODIC REVIEW FREQUENCY: 2 W4A PERIODIC REVIEW DUE DATE: 4,!30!4 7 l REVISED PERIODIC REVIEW DUE DATE:
PROCEDURE TYPE: QA Program ( / ) YES ( ) NO "s PlantProcedure(/)YES ( ) NO REVIEW METHOD: ( ) Alternate ( ) Expedited ([)PORC ( ) ERC l Prepared by NdN C Date S!M!Si ; Reviewed by - ' l N
~ Supervisor Date MO' I Recommended CNb -Date O!Aok7 '
Functional Unit Manager hb'D3 PORC Commithee Meeting No. Date OMW MM Date ERC C ittee Mee in No. Approved by ,4dg rg G < g ~~ Date E if FORMNDAP-QA-0002-1,Rev.1,PageIof1
j. i
^ E0-100-113 f ! Revision 7 A Page 32 of 68 LQ/L-19 STOP INJECTION MD PREVENT INJECTION EXCEPT FROM:
e SLC e CRD e RCIC e HPCI , IF any system.is injecting, other than the exceptions listed, this step requires that these systems itan injection. All injection systems other than the exceptions listed, must 4 be orevented from injection. For feedwater, this would mean tripping feedwater pumps or closing their discharge valves. 4 For condensate, this would mean oreventina uncontrolled injection below RPV pressure 600 psig using valves or if needed, tripping condensate pumps. For RHR and Core Spray this would require preventina injection in accordance with overriding section of their , respective operating procedures. , Injection into the RPV is stopped and prevented, while rapid RPV depressurization proceeds, in order to prevent uncontrolled injection of cold water as RPV pressure decreases below the shutoff head of operating system pumps. Injection from boron injection systems and CRD is not terminated because operation of these systems may be needed to establish and maintain reactor shutdown. Further, the injection flowrates from these systems are small compared to those of the other Table 15 systems. Injection from RCIC is not stopped because .the injection flowrate from this system is small . Injection from HPCI is )ermitted to avoid potential isolation and minimize tie transient that may
. . occur when RPV injection is restored. It also helps reduce RPV pressure by spraying cold water into the steam space.
Continued operation of the RCIC and HPCI turbines aids in depressurizing the.RPV, and operation during RPV depressurization'is not expected to result in significant injection flowrate variations due to the stability of the system flow controllers. ~ n (
Reference:
SSES-EPG C5-3.1) 4
E0-100-113 I ' Revision 7 Page 33 of 68 CAUTION A RAPID INCREASE IN INJECTION MAY INDUCE LARGE PWR EXCURSION AND SUBSTANTIAL CORE DAMAGE This caution is applicable throughout this flowpath. The applicability of this Caution highlights the potential for large reactor power excursions and subsequent core damage from rapid injection of cold, unborated water when injection is commenced after depressurization is initiated. (
Reference:
SSES-EPG Caution #7) LQ/L-20 WHEN RAPID DEPRESS HAS BEEN INITIATED COMMENCE AND IRRESPECTIVE OF VORTEX LIMITS SLOWLY INCREASE INJECTION q TO RESTORE AND MAINTAIN LVL BETWEEN -60" AND -161" USING TABLi 15 SYSTEMS Re-establishing injection into the RPV is required in order to adequately cool the core and to make up the mass of steam being rejected through open SRVs. Since the reactor may become critical during this evolution, injection into the RPV is increased slow!y to preclude the possibility of large power excursions caused by rapid injection of cold unborated water. The level control band of -60" to -161" is the widest, acceptable water level control band. Although level fluctuations within this band are safe, it is very desirable to maintain level within the more restrictive target area of
-110" to -80". The target area and expanded band are shown in Figure 8, Water Level Operation Guidance. Operation outsido of the target area has numerous disadvantages which are described in LQ/L-14. The intent of this step is to restore and maintain level within the target band at all times unless prohibited by system perturbations or inadequate vessel injection, and remain within the expanded band at all times.
v
I NUREG-1021 Rev.7,Supp.1 1 I
.g Operator Licensing Examiner Standards I
1l l Manuscript Completed: June 1994 Date Published: June 1994 Division of Reactor Controls and Human Factors Omce of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 J j .
,... ~ .u, f
ES-303 (d DOCUMENTING AND GRADING OPERATING TESTS ADMINISTERED AT POWER REACTORS A. PURPOSE This standard describes the procedures for documenting all categories of the ' operating test, collating the data to arrive at a pass or fail recommendation, and reviewing the documentation to ensure quality. B. GENERAL INSTRUCTIONS AND GUIDELINES
'l. In this standard, it is assumed that the examiner prepared and administered the operating test in accordance with ES-301 and ES-302, respectively.
4 2. As stated in ES-302, it is essential that the simulator scenario documentation for the applicants in an operating crew is consistent and mutually supportive. Operating errors that involved more than one applicant should be noted by both evaluating examiners. If it was not possible for the examination team members to discuss and compare their observations before leaving the site, the chief examiner shall schedule a conference call after the examiners return to their respective offices.
- 3. The pro:edures contained herein require the examiner to evaluate the O(/ applicant's performance on the operating test and make a judgement whether the applicant's level of knowledge and understanding meet the minimum requirements to safely operate the facility for which the license is sought. The examiner should evaluate each noted deficiency in light of the total field of knowledge and ability demonstrated by the applicant in that subject area.
The examiner should consider the accuracy of the applicant's responses and actions and the importance factors of the K/As associated with the questions and tasks administered. For example, if an applicant is asked 2 questions in a subject area and gives a weak response to the question with a K/A importance rating of 2.7 and a correct response to the question with a K/A importance rating of 3.9, a satisfactory grade would normally be assigned for the subject arsa. The terms " satisfactory" and " unsatisfactory" when used to evaluate an applicant's performance on all or part of the operating test as discussed herein are defined as follows:
- a. S - Satisfactory Working Knowledge and Understanding The applicant may have some slight or minor difficulty describing system interactions. Competence in the operation of equipment associated with ti.e system is very good, although there may be Examiner Standards 1 of 27 Rev. 7, January 1993
l ES-303 l some hesitation while' discussing or perfoming'some tasks. The applicant appears to be familiar with the equipment and'
- procedures.
L b. U - Unsatisfactory or Poor Working Knowledge and Understanding The applicant has difficulty answering questions in depth and
- describing the interactions of systems. Discussions or behavior
' while operating equipment show lack of familiarity with the equipment and procedures. The applicant is unable to answer questions or provides incorrect or incomplete answers. The j applicant is unfamiliar with the subject or system, as evidenced
- by hesitant answers, inability to locate information, inability to i locate control board indications or controls, and lack of
{_ knowledge of system operating procedures. i ! 4. To maximize accuracy and completeness, the examiner shall review,
- evaluate, and finalize each applicant's operating test documentation in l accordance with the procedures in Section C as soon as possible after administering the test.
C. GRADING AND DOCUNENTATION PROCEDURE f
- 1. Review and Cateaorize Rouah Notes and Documentation l a. . Review the job performance measures (JPMs) and simulator scenarios that were performed and the questions that were asked. The examiner will review all the rough notes and documentation generated while administering the operating test to determine the L areas in which the applicant was deficient.
If the applicant generated or used any material such as figures, drawings, flowcharts, or forms during the operating test, the material may be used to aid in documenting the applicant's ! performance. If it contributes to an unsatisfactory performance evaluation, it shall be appropriately marked and cross-referenced
- to the applicable deficiency and attached to the examination package for retention.
- b. Verify the validity and technical accuracy of any questions that i were asked during the walk-through test but had not been
, prescripted and any unexpected events or actions that occurred during the simulator test. . c. On the rough notes and documentation, label or highlight every action, response, note, or comment that may constitute a
- performance deficiency.
- d. ' Label each deficiency related to the applicant's administrative )
./ and plant system knowledge and abilities with the alphanumeric
'* Examiner Standards 2' of 27 Rev. 7, January 1993 I i
1 I i j )O i V ES-303
- code of the administrative topic (e.g., A.1) or the control room
- or plant system (e.g., 8.1.3 or B.2.1) to which it applies.
- e. Review each simulator performance deficiency and, using the i competency and rating factor descriptions on Attachment 1 of ES-
' 301 and Form'ES-303-3 (RO) or ES-303-4 (SRO) as guides, code it ! with the number and letter of the rating factor (s) it most i accurately reflects (e.g., C.4.A). . Whenever possible, the
- examiner should attempt to identify the root cause of the 4 applicant's deficiencies and code each deficiency with no more than 2 different rating factors.
4 2. Evaluate the Anolicant's Performance ! } After categorizing and coding the rough notes, the examiner shall ' review, evaluate, and grade the applicant's performance in Category A,
" Administrative Topics," Category B " Control Room Systems / Facility
, Walk-through," and Category C, " Integrated Plant Operations," of the operating test as follows:
- a. Form ES-303-1, Category A l Review the identified deficiencies and make a decision whether the applicar.t's knowledge and understanding of each administrative topic was satisfactory or unsatisfactory (refer to the discussion 4
O- in Section B.3) and' document the grade by placing an "S" or "U" in the appropriate block or, page 2 of Form ES-303-1. Every
. unsatisfactory grade must be supported with detailed documentation as discussed in Section C.3. l 7
When all the topics in Category A have been graded, assess the-applicant's topic grades and deficiencies and assign a single "S" or "U" grade for the category. If the applicant has a "U" in only b i one administrative. topic, the examiner may fail the applicant in l Category A depending on the importance of the identified y
- deficiency. However, if the applicant has a "U" in two or more of !
l the administrative topics, the examiner must assign a grade of "U"
- for Category A. Place the assigned grade in the " Administrative Topics" block of the " Operating Test Summary" on page 1 of Form ;
ES-303-1.
- b. Form ES-303-1, Category B L On page 2 of the applicant's Form ES-303-1, enter the names of the systems and JPMs examined during Subcategories B.1, " Control Room Systems". and B.2, " Facility Walk-through,"-of the operating test. J Also enter the safety function number from the appropriate !
, Knowledge and Abilities Catalog (NUREG-1022 for PWRs and NUREG- ; 1023-fcr.BWRs). l Examiner Standards- 3 of 27 Rev. 7, January 1993 r q
i ES-303 l Evaluate each deficiency coded for Category B in the rough notes ~ to determine if the JPMs listed on Form ES-303-1 were properly performed. If the following criteria are met, assign a > satisfactory grade by placing an "S" in the "JPN Evaluation" column for that system, otherwise enter a "U": ' [ - if the JPM was time-critical, it must be completed within-the allotted time I - all the critical steps identified for a JPM must be l
- completed correctly- ,
- If the applicant missed a critical step but later performed it a correctly and accomplished the task standard without degrading the ,
i condition of the system or the plant, the applicant's performance on that JPN should be graded as satisfactory. I Further evaluate the Category B documentation to determine the number of prescripted questions that the applicant answered satisfactorily for each system listed on Form ES-303-1. It is permissible to allot partial credit, when appropriate, to determine whether or not the applicant satisfactorily answered a , question. The applicant must demonstrate at least 70 percent-knowledge in order for his or her answer to a question to be O considered satisfactory. If the applicant satisfactorily answered both prescripted questions, enter an "?" in the " Question Grade" column for that 4- system on page 2 of Form ES-303-1. If the applicant answered only one of the prescripted questions correctly, enter an "S" unless-the knowledge or ability that was missed is of such safety
- significance that an unsatisfactory system grade is justified, then enter a "U". If the applicant's answers to both prescripted questions were unsatisfactory, enter a "U" in the " Question Grade"
- column for the system, i
Every grade of "U" on a JPN or the system questions must be supported with detailed documentation as discussed in Section C.3.
- For each Category B system listed on page 2 of Form ES-303-1, enter a " System Grade" of "S" if the applicant's performance on both the JPM and the prescripted questions was satisfactory. If
- the applicant was assigned a "U" for either the JPM or the .
prescripted questions, then enter a "U" for the " System Grade." ! When all the systems in Category B have been graded, determine an overall grade for Category B by calculating the percentage of , satisfactory system grades. If the applicant has an "S" on 80 l' percent or more of the systems examined the applicant passes Category B and gets an "S" overall. If fewer than 80 percent of
~k Examiner Standards 4 of 27 Rev. 7, January 1993 l p
1
-, - - . - - - - , . - . _ . . y - m-._ .,_r
s ES-303 the system grades are "S's" the applicant fails Category B and 4 shall be given a "U" overall. Document the applicant's grade by ' placing an "S" or a "U" in block "B. Control Room Systems and Facility Walk-through," in the
" Operating Test Summary" on page 1 of Form ES-303-1.
~
- c. Fom ES-303-1, Category C Using Form ES-303-3 or ES-303-4, depending on the applicant's license level, evaluate any deficiencies coded for Category C and '
circle the integral rating value (1 through 3) corresponding to the behavioral anchor that most accurately reflects the applicant's performance. As discussed in ES-301, Competency 5 is optional for SRO upgrade applicants. Multiply each integral rating value by its associated weighting factor to obtain a numerical measure of the applicant's l performance on each rating factor, then circle the corresponding numbers on page 3 of the RO or SRO applicant's Form ES-303-1.
- For each rating factor, place check marks in the columns on page 3 l of Form ES-303-1 corresponding to the numbers of the scenarios in !
i which the rated behavior was observed.
/ For each competency on page 3 of Form ES-303-1, sum the circled 4 rating factor grades and enter the resulting competency grade in j the " Total" column (the grades should range between 1 and 3).
{ Using the following evaluation criteria, determine if the , applicant's overall performance in Category C is satisfactory or unsatisfactory and document the grade by placing an "S" or a "U" I
- in block "C. Integrated Plant Operations," in the " Operating Test Summary" on the page 1 of Form ES-303-1.
- If the " Total" grade for All competencies is greater than 1.8, the applicant is satisfactory.
- If the " Total" grade for Competency 6. " Communications and Crew Interactions," is less than or equal to 1.8 but greater than 1.0 And the " Total" grades for All other competencies are 2.0 or greater, the applicant is satisfactory.
- If the " Total" grade for Competency 6 is 1.0 9.t the " Total" grade for any other competency is 1.8 or less, the applicant is unsatisfactory.
~ NOTE: Competency 5, " Control Board Operations," is optional for SRO upgrade applicants. However, if it is evaluated, it shall be factored into the applicant's final grade.
Examiner Standards 5 of 27 Rev. 7, January 1993
d ES-303 m i - i . Justify in detail, as discussed in Section C.3,_ each rating factor that is assigned an integral rating value of 1, regardless of the
" Total" grade determined for the associated competency. If the i applicant's performance in Category C is unsatisfactory overall, justify in detail' every rating factor assigned an integral rating
.value of I a.t 2 for each competency that has an unsatisfactory score. j
- 3. Finalize the Documentation
~
- a. Review and finalize the simulator scenarios that were run for '
Category C of the operating test. I; Complete Form ES-301-3 by entering the applicants' names, the positions they occupied during the scenario, and-the facility's name on the top of the form. Enter any scenario revisions made during the test on Form ES-301-3 so that each form accurately shows all the events that actually occurred during each scenario. Change the event numbers, malfunction numbers, malfunction type, and descriptions as necessary to reflect the "as run" conditions. j i This may be done by making pen-and-ink changes or retyping the
- scenario, provided the final form is clear and legible.
! Update each Form ES-301-4 to reflect the "as run" conditions. Any events that were not run shall be marked "not used" or discarded, and new forms shall be filled out for events that were not >
- originally planned. 'The examiner may neatly enter notes, l comments, and additional actions in the spaces between the expected operator actions.
The final Forms ES-301-3 and ES-301-4 must be a clear, legible, and sequential record of the actual events and actions that occurred during the simulator test. The forms sent to the applicant shall not contain any rough notes or irrelevant i comments. Any events or malfunctions that did not function as expected or I were not useful in evaluating the applicants (e.g., a surveillance i test that required a long time to perform) should be noted on the master copy of the scenarios to aid in future scenario preparation.
- b. Review the applicant's Form ES-303-1 and the rough documentation and justify in detail on Form ES-303-2, " Operating Test Comments,"
every knowledge or ability deficiency that contributed to the assignment of a "U" in any administrative topic in Category A, a "U" for any JPM in Category B, or an unsatisfactory grade for any prescripted JPM question. Deficiencies that contributed to ! Category C integral rating factor grades of I and 2, as discussed ' L , in Section C.2.c, must also be justified in detail if they Examiner Standards 6 of 27 Rev. 7, January 1993 i
, - , . , o _
4 i e i j .. ES-303 ! resulted in an unsatisfactory score in the associated competency.
; Deficiencies that do not contribute to an unsatisfactory grade p a nywhere on the operating test may be documented at the examiner's option.
Provide the following specific information, as applicable to each
- noted deficiency: l 1
- the question asked or task administered (i.e., describe the ; JPM or the simulator scenario and event and the applicant's
! position on the operating crew)
- the applicant's incorrect answer or action and an indication whether the action was a JPM critical step l
- the lack of knowledge or ability that the applicant
- demonstrated
- the consequences of the applicant's incorrect answer or action
- the correct answer or action with an appropriate facility i
reference (e.g., lesson plan, system description, procedure name and number)
- the K/A number and its importance rating (as given in NUREG-1122 or NUREG-1123) and the facility's learning objective 4 - the item from 10 CFR 55.45(a) that the ' applicant did not understand or was unable to perform General statements such as "did not know decay heat removal system" are inadequate.
Use printouts or strip chart recordings generated during the 4 simulator test and drawings and illustrations generated by the
- applicant to substantiate comments whenever possible.
The examiner should retain his or her rough documentation until the chief examiner and management have reviewed the examiner's recommendations and concurred in the results (refer to Section D and ES-501). l. 4
- c. Cross-reference each comment on Form ES-303-2 with the specific task, subject, or competency rating factor to which it applies on
, the applicant's Form ES-303-1. Do this by entering the applicable , alphanumeric subject reference from Form ES-303-1 (e.g., A.2, B.I.3.C. C.4.B) in the left-hand column of Form ES-303-2, and entering the page number on which the comment is found in the appropriate block on Form ES-303-1. D.- EXAMINER RECOMMENDATIONS
- 1. 'After grading and documenting the operating test in accordance with Section C, the examiner who administered the test shall make an overall recommendation by checking the " Pass" or " Fail" block, signing, and
\ Rev. 7, January 1993 Examiner Standards 7 of 27 )
i i
i ES-303 , , l dating the " Examiner Recomme.ndations" section on the applicant's Form i ES-303-1. The examiner ma) make a pass recommendation only if all summary blocks of the operating test contain satisfactory (S) grades or , the letters "N/E."
- 2. If the written examination was not waived and the written examination i data has not yet been entered on Form ES-303-1, the examiner will route
- the examination package to the written examination grader for processing i in accordance with ES-403. If the written examination results have already been entered or the examination was developed by a contract examiner, forward the examination package to the chief examiner.for review.
- 3. The chief examiner, or a management-approved designee, will review the ,
. grading of the operating test to verify that the examiner's comments , appropriately support his or her recommendation and to ensure that the ). operating test meets the requirements of ES-301. If the chief examiner ' 1 or designee does not agree with any of the examiner's recommendations,
- he or she shall confer with the examiner before overturning the ,
Such disagreements are not common and usually arise i recommendation. !. because an unsatisfactory grade is not adequately justified. It is,
- therefore, very important for examiners to be complete and accurate in
- their grading and documentation.
- / \ 4. The chief examiner'or designee shall'make an independent pass or fail
\s_/ recommendation, sign the " Final Recommendation" block on Form ES-303-1, and forward the package to the responsible section chief for review in j accordance with ES-501. The section chief, or higher, must concur in any recommendation to overturn the examiner's results, and the specific reasons for this action will be explained on Form ES-303-2. ,
i j ATTACHMENTS / FORMS: ! Form ES-303-1, " Operator Licensing Examination Report"
- Form ES-303-2, " Operating Test Comments" i Form ES-303-3, "R0 Competency Grading Worksheet for Integrated Plant Operations" Form ES-303-4, "SRO Competency Grading Worksheet for Integrated Plant Operations" 1
4 iOi E xaminer Standards 8 of 27 Rev. 7 January 1993 i 1
l i l ES-303 Operator License Examination Report Form ES-303-1 i. ! U.S. NUCLEAR REGULATORY COMISSION OPERATOR LICENSE EXANINATION REPORT l APPLICANT'S NAME DOCKET NUMBER 55- __ I R EXAMINATION TYPE (INITIAL OR RETAKE) FACILITY NAME ! REACTOR OPERATOR HOT SENIOR REACTOR OPERATOR (SRO) INSTANT COLD FACILITY I SR0 UPGRADE BWR DESCRIPTION : ! SRO LIMITED TO FUEL HANDLING PWR .: i WRITTEN EXANINATION SUMARY WRITTEN BY TOTAL EXAMINATION POINTS GRADED BY TOTAL APPLICANT POINTS . DATE ADMINISTERED APPLICANT GRADE % OPERATING TEST SUMARY ! ADMINISTERED BY DATE ADMINISTERED
- A. ADMINISTRATIVE TOPICS B. . CONTROL ROOM SYSTEMS AND FACILITY WALK-THROUGH l
C. INTEGRATED PLANT OPERATIONS (SIMULATOR TEST) EXANINER RECOMEMATIONS l CHECK BLOCKS PASS FAIL WAIVE SIGNATURE DATE WRITTEN EXAMINATION OPERATING TEST FINAL RECOMENDATION LICENSE RECOMEWATION ISSUE LICENSE SIGNATURE - SECTION CHIEF DATE DENY LICENSE O Examiner Standards 9 of 27 Rev. 7 January 1993
ES-303 2 Form ES-303-1 APPLICANT DOCKET NUMBER: 55- PAGE OF A. ADMINISTRATIVE TOPICS EVALUATION COMMENT PAGE (S OR U) NUMBER
- 1. CONDUCT OF OPERATIONS
- 2. EQUIPMENT CONTROL
- 3. RADIATION CONTROL
- 4. EMERGENCY PLAN B.1 CONTROL ROOM SYSTEMS SAFETY JPN GRADE FUNCTION (S'OR U)
QUESTION GRADE ., (S OR U) l SYSTEM COMMENT GRADE PAGE
- , SYSTEM /JPM TITLE (S OR U) NUMBER 4
1 1. 2. 3. 4. I 5. j 6. I 7.
~
B.2 FACILITY WALK-THROUGH 1. 2, 3. Examiner Standards 10 of 27 Rev. 7, Sup. 1, June 1994
L N ES-303 3.a Form ES-303-1 APPLICANT DOCXET NUMBER: 55- PAGE OF i C. REACTOR OPERATOR INTEGRATED PLANT OPERATIONS (SINULATOR TEST) GRADING SUMARY SCENARIO $ COMMENT-COMPETENCIES / RATING FACTORS WEIGNT 3.0 2.0 1.0 TOTAL OBSERVED PAGE No.
- 1. ALARM 5/ ANNUNCIATORS 1 2 3
, A. NOTICE / ACKNOWLEDGE 0.30 0.90 0.60 0.30 _,,,. ,,,,,,,. __,, ,_
- 5. INTERPRET / VERIFY 0.40 1.20 0.00 0.40 ,, _ _ _
C. Pal 0RITIZE 0.30 0.90 0.60 0.30 ( ) _ _ _ _ 1 2 3
- 2. DIAGNOSIS A. RECOGNIZE 0.40 1.20 0.80 0.40 _ ___ _
B. USE OF REFERENCE MATERIAL 0.20 0.60 0.40 0.20 _, , ,,,,,,, C. DIAGNO$f 0.40 1.20 0.80 0.40 ( ) _ ,,_,, _,, ._. 2 3
- 3. SYSTEM RESPONSE 1 A. LOCATE / INTERPRET 0.33 1.00 0.67 0.33 ._ , ,,,,,_ _, ,
0.33 1.00 037 0.33 ( B. SYSTEM OPERATION KNOWLEDGE _,,,,,, ,, ,,,,. __, C. EFFECT OF ACTIONS 0.33 1.00 0.67 0.33 ( ) _,,,,, ,_, ,,,_., 4 PROCEDURES /TECM SPECS 1 2 3 A. REFERENCE 0.20 0.60 0.40 0.20 , , , , __, B. E0P ENTRY /IMMEDIATE ACTIONS 0.40 1.20 0.80 0.40 __,_ ,,,_. ,,,,,, C. PROCEDURE COMPLIANCE 0.20 0.60 0.40 0.20 _,,,_ ,,,,,,,, _ .,,,_ D. TECH $PEC ENTRY 0.20 0.60 0.40 0.20 ( ) ,,,,,,,, _,,,,, _,,,_ , , , , , , ,
- 5. CONTROL BOARD OPERATIONS 1 2 3 A. LOCATE 0.25 0.75 0.50 0.25 ,,, ,_,,,, .,._
B. MANIPULATE 0.25 0.75 0.50 0.25 _,,_, _. ,_,,,, ,. C. RESPONSE 0.25 0.75 0.50 0.25 ,,,,,, ,,,, .,,,,_ ,,,_. D. MANUAL CONTROL 0.25 0.75 0.50 0.25 ( ) , _ _ _
- 6. COMMUNICATIout 1 2 3 A. PROVIDE INFORMAfl0N 0.33 1.00 0.67 0.33 ,,,,, . _ _ _,_, ,,
B. RECEIVE INFORMATION 0.33 1.00 0.67 0.33 .,,,,,, ,,,_, ,, _ C. CARRY CUT INSTRUCTIONS 0.33 1.00 0.67 0.33 ( ) _,,_ ,,,,,_ ,,,,,_ ,_ O i i V Examiner Standards 11 of 27 Rev. 7, January 1993
1 ES-303 3.b Form ES-303-1 v APPLICANT DOCKET NUMBER: 55- PAGE OF C. SENIOR REACTOR OPERATOR INTEGRATED PLANT OPERATIONS i (SIMULATOR TEST) GitADING
SUMMARY
COMPETENCIES / SCENARIOS COMMENT RATING FACTORS WEIGHT 3.0 2.0 1.0 TOTAL 00 SERVED PAGE NO. l l
- 1. ALARMS / ANNUNCIATORS 1 2 3 l A. PRIORITIZE 0.30 0.90 0.60 0.30 __, __,, _ __.
B. INTERPRET 0.35 1.05 0.70 0.35 __,, _ _,_ ,,_, 0.70 0.35 C ) ! 0.35 [C. 'AssFY 1.05 ,_._ _,_ ,,,,,_ _
- 2. DIAGN0515 1 2 3 A. RECOGNIZE 0.25 0.75 0.50 0.25 ,,,,,,_ _ __, ___
B. ACCURACY 0.25 0.75 0.50 0.25 __ _ ,,._, __,, 1 C. DIAGNOSE 0.25 0.75 0.50 0.25 _,_ _.,_ _ , _ _,
-)
D. CREW RESPONSE 0.25 0.75 0.50 0.25 ( ) _ __. _ ,,_.
]
- 3. STITEM RESPONSE 1 2 3 A. INTERPRET 0.35 1.05 0.70 0.35 _ ,__, ,_ _.,
B. ATTENTIVE 0.20 0.60 0.40 0.20 _ _ _ _ C. PLANT EFFECTS 0.45 1.35 0.90 0.45 ( ) ,_, _ .__ _._
- 4. PROCEDURES 1 2 3 l
t A. REFERENCE 0.25 0.75 0.50 0.25 _,_,
- 8. CORRECT USE 0.50 1.50 1.00 0.50 ,_ _,_. _ _ , _
C. CREW IMPLEMENTATION 0.25 0.75 0.50 0.25 C ) _ _ _
- 5. CONTROL 804RD OPERATIONS 1 2 3 A. LOCATE 0.25 0.75 0.50 0.25 ,,_ _ __. _
B. MANIPULATE 0.25 0.75 0.50 0.25 ,__ _ _._ C. RESPONSE 0.25 0.75 0.50 0.25 _,,,,,, __,. _ ,_ D. MANUAL CONTROL 0.25 0.75 0.50 0.25 ( ) _ ___
- 6. COMMUNICATIONS 1 2 3 A. CLARITT 0.45 1.35 0.90 0.45 ,_
B. CREW INFORMED 0.35 1.05 0.70 0.35 _ . , , _, _, C. RECEIVE INFORMATION 0.20 0.60 0.40 0.20 ( ) _,_ , ___ _._
- 7. DIRECTING OPERAT!ONS 1 2 3 A. TIMELY ACTION 0.20 0.60 0.40 0.20 ,_, ,_.,, _
B. SAFE DIRECTIONS 0.40 1.20 0.80 0.40 _, _ _ ___ C. OVERSIGHT 0.20 0.60 0.40 0.20 , _, .,,,,,,, D. CREW FEEDBACK 0.20 0.60 0.40 0.20 ( ) _ _
- 8. TECHNICAL SPECIFICATIONS 1 2 3 A. RECOGN!ZE 0.40 1.20 0.80 0.40 _ __,. .,_,, _._
B. LOCATE 0.20 0.60 0.40 0.20 _ _ _ _ C. COMPLIANCE 0.40 1.20 0.80 0.40 ( ) _ _,_ .__ ._, l Euniner Standeres 12 of 27 Rev. 7, s w. 1, June 1994
l
)
J ES-303 Operating Test Comments Form ES-303-2 l l
- l APPLICANT DOCKET NUMBER: 55- PAGE OF
- i. 1 2
FORM ES-303-1 CON 9ENTS l I CROSS REFERENCE l i s i i 4 1 i . 1 1 1 i i. Examiner Standards 13 of 27 Rev, 7, January 1993 l
(. ES-303 R0 Competency Grading Worksheet Form ES-303-3 i For Integrated Plant Operations i
- 1. ' UNDERSTAND AND INTERPRET ANNUNCIATORS AND ALARM SIGNALS I
DID THE APPLICANT: l l (a) NOTICE and ACKNOWLEDGE alarms? 3 2 1 Consistent and Minor difficulties Failed to notice and ! timely or lapses in awareness acknowledge important i acknowledgement or response alarms; distracted by nuisance alarms, etc. [.30 X =
]
(b) Correctly INTERPRET and VERIFY that annunciators and alarm signals were consistent with plant and system conditions (including the use of alarm response procedures.(ARPs), when necessary)? 3 2 1 Consistent and Minor inaccuracies Significant inaccur-efficient in interpreting acies resulted in interpretation and verifying signals plant degradation; and verification poor use of ARPs [.40X =
]
, (c) ATTEND to ANNUNCIATORS and ALARM SIGNALS in order of importance and
- severity?
3 2 1 l Consistent Minor inaccuracies Did not prioritize attention in and oversights attention to signals; , all cases inattentive to ! important alarms [.30 X =
] <
O Examiner Standards 14 of 27 Rev. 7, January 1993
ES-303 2 Form ES-303-3
- 2. DIAGNOSE EVENTS AND CONDITIONS BASED ON SIGNALS AND READINGS DID THE APPLICANT:
(a) RECOGNIZE off-normal trends and status? 3 2 1 Quick and accurate Some delays in Serious omissions, recognition recognizing off- delays, or normal conditions inaccuracies in recognizing events
=
[.40 X ] (b) Correctly USE REFERENCE MATERIAL (prints, books, charts) to aid in diagnosing and classifying events and conditions? I 3 2 1 Correctly used Minor errors in Did not use or d references, using or relying incorrectly used
- when necessary on references references to diagnose events
[.20 X =
]
(c) Correctly DIAGNOSE plant conditions based on control room indications? 3 2 1 Diagnoses were Minor errors or Faulty diagnoses accurate difficulties adversely affected in diagnoses plant status [.40 X =
]
b
\- Examiner Standards- 15 of 27 Rev. 7, January 1993 0
l
/'^
ES-303 3 Form ES-303-3 I
- 3. UNDERSTAND PLANT AND SYSTEM RESPONSE DID THE APPLICANT: l (a) LOCATE and correctly INTERPRET relevant instruments and other indicators of plant and system response (s)? l 3 2 1 Accurate and Minor errors in Serious omissions I efficient location locating and inter- or inaccuracies ,
and interpretation preting instruments in interpreting I of instruments and displays instruments [.33 X =
]
. (b) Demonstrate KNOWLEDGE of SYSTEM OPERATION, including set points, 2 interlocks, and automatic actions? 3 2 1
' O Demonstrated Minor instances of Inadequate knowledge iO thorough under- errors due to in- resulted in plant j standing of adequate knowledge degradation !
system operations , i [.33 X =
]
(c) Demonstrate an understanding of how his or her ACTIONS (or inaction) AFFECT PLANT and SYSTEM CONDITIONS? 3 2 1 l Understood the Minor misunder- Appeared to act effect of actions standing of effect without knowledge l on plant and of actions on plant of or regard for J systems and systems effect on plant and systems ; [.33 X =
]
Examiner Standards 16 of 27 Rev. 7, January 1993
l l ES-303 4 Form ES-303-3
- 4. - COMPLY WITH AND USE PROCEDURES AND TECHNICAL SPECIFICATIONS ]
l DID THE APPLICANT: (a) REFER.T0 the appropriate procedure in a timely manner? 3 2 1 i
. Quickly located Minor difficulties and Problems and failures appropriate oversights in referring in referring to procedures to appropriate procedures in 2
procedures important instances t' [.20 X =
]
f (b) RECOGNIZE E0P ENTRY CONDITIONS and carry out appropriate immediate l actions without the aid of references or other forms of assistance? l 3 2 1 Consistent, accurate Minor lapses or errors, Did not accurately i and timely but actions generally execute actions , recognition appropriate [.40 X -
) i (c) COMPLY WITH procedures (including precautions and limitations) in an accurate and timely manner?
3 2 1 . Accurate and Few errors; corrections Many significant ' timely compliance made in sufficient time errors; excessive to avoid adverse effect assistance required I [.20 X -
]
(d) RECOGNIZE plant conditions that are addressed in technical specifications? 3 2 1 , Recognized and Minor assistance Did not recognize complied with required to recognize conditions and/or LCOs and-action conditions and/or comply with LCOs and statements comply with LCOs and action statements action statements , [.20 X -
]
- Examiner Standards 17 of 27 Rev. 7, January 1993 I i
.% -. 4,_ . . - . . ,.m.., , .._ . ,
.. . . . . . . . . - -_ . .- .-. -. -. . . ~ -.
t i l
.Y Form ES-303-3
-ES-303 5
- 5. OPERATE THE CONTROL BOARD DID THE APPLICANT:
(a) ' LOCATE CONTROLS efficiently.and accurately? 3 2 1 4 Promptly located Some minor hesitancy Unable to locate , i appropriate controls and difficulty in controls without ! in all instances locating controls- assistance [.25 X =
]
l (b) MANIPULATE CONTROLS in an accurate and timely manner? .i 3 2 1 i Control Manipulations Minor shortcomings, Improper manipul-
- were consistently but efficiently ations resulted in accurate and mitigated any major system timely resulting consequences perturbations
[.25 X = J (c) ACT appropriately in response to INSTRUMENT READINGS 7 3 2 1 Responses were Generally adequate Failed to react appropriately to F appropriate ~and response; some timely errors and lapses instrument readings without assistance f [.25 X -
]
(d) Take MANUAL CONTROL of automatic functions when appropriate? l 3 2 1 Took manual control Minor delays and some Depended on automatic when appropriate prompting necessary actions; had to be s before overriding prompted to take automatic functions manual control [.25 X =
]
Examiner Standards 18 of 27 Rev. 7, January 1993
e (m ES-303 6 Form ES-303-3 .
- 6. C0ltiUNICATE AND INTERACT WITH OTHER CREW MEMBERS DID THE APPLICANT:
(a) PROVIDE clear and accurate INFORMATION on system status to others for the performance of their jobs? 3 2 1 Provided others Minor instances of Failure to accurately with accurate and needing to be provide important pertinent information prompted for input; information to others some incomplete and jeoptrdini plant inaccurate information status 1 [.33 X ] (b) Effectively RECEIVE INFORMATION from others (including requesting, acknowledging, and attending to information)? 3 2 I f Responded and Minor instances Inattentive to (- reacted appropriately of failure to information to information acknowledge or provided by others from others respond to information [.33 X -
]
(c) CARRY OUT the INSTRUCTIONS of the supervisor successfully? 3 2 1 Ably carried out Minor hesitancy and Failed to promptly all supervisory difficulty in following and accurately follow objectives; dis- orders, but ultimately directions; blindly cussed instructions complied successfully complied with when questionable erroneous orders [.33 X -
]
O Examiner Standards 19 of 27 Rev. 7, January 19C
l l l
- ES-303 SRO Competency Grading Worksheets Form ES-303-4 For Integrated Plant Operations
- 1. UNDERSTAND AND INTERPRET ANNUNCIATORS AND ALARM SIGNALS DID THE APPLICANT:
(a) NOTICE and ATTEND to annunciator and alarm signals in order of their importance and severity? l 3 2 1
~
Responded accurately Minor difficulties Failed to attend to and efficiently in in attending to signals or prioritize import- l 4 all instances or prioritizing attention ant alarms; responded j slowly; distracted by i nuisance alarms [.30 X -
]
l (b) Correctly INTERPRET the meaning and significance of alarms and annunciators (including the use of alarm response procedures (ARPs), when necessary)? 3 2 1 O Understood and quickly determined Minor inaccuracies or delays in Misinterpretations, delays, or misuse of
- what failures alarms interpreting alarms ARPs resulted in were indicating plant degradation
[.35 X -
]
(c) VERIFY that annunciator and alarm signals were consistent with plant and system conditions? l 3 2 1 Ensured proper Minor lapses in Failed to correctly j verification when alarm verification, verify signals on necessary but no inappropriate important occasions; actions were taken as did not notice incon-a result of inadequate sistencies between verification alarms and plant conditions [.35 X -
]
V' - Examiner Standards 20 of 27 Rev. 7, January 1993
- v. ,,- -- -- - - - - - - - =w -
- wer-w v .-
ES-303 2 Form ES-303-4
- 2. DIAGNOSE EVENTS AND CONDITIONS BASED ON SIGNALS AND READINGS ;
DID THE APPLICANT: (a) RECOGNIZE off-normal trends and status? 3 2 1 1 Quick and accurate Some delays in Serious omissions, recognition recognizing off- delays, or normal conditions inaccuracies in recognizing trends [.25 X =
]
(b) Ensure the collection of CORRECT, ACCURATE, and COMPLETE information and reference material on which to base diagnoses? 3 2 1 : Ensured that all Minor instances of over- Serious instances of relevant indications looking, overrelying on, misusing or failing and references were or misinterpreting indic- to use important O checked ations and/or referances information or data [.25 X -
]
(c) Correctly DIAGNOSE plant conditions based on control room indications? 3 2 1 Diagnoses of plant Minor errors or Faulty diagnoses 1' conditions were difficulties in adversely affected accurate diagnosing conditions plant status [.25 X -
]
(d) Ensure that CORRECT and TIMELY DIAGNOSTIC ACTIVITIES were carried out by the CREW 7 3 2 1 Ensured effective Minor errors or Faulty diagnostic ! diagnostic activities difficulties in activities by crew and diagnoses by crew diagnosing by crew adversely affected plant status [.25 X =
]
Examiner Standards 21 of 27 Rev. 7, January 1993
.( ES-303 3 Form ES-303-4
- 3. UNDERSTAND PLANT AND SYSTEM RESPONSE DID THE APPLICANT:
(a) INTERPRET control room indicators correctly and efficiently to ascertain and verify the status and operation of plant systems? 3 2 1 Accurate and efficient Minor errors in Serious omissions, interpretation of interpreting delays, or inaccur-instruments and instruments and acies in interpreting displays displays instruments and displays
=
(.35 X ] Remain ATTENTIVE to control room indications? : (b) 3 2 1 Regularly scanned Sporadically scanned Rarely scanned indic-p)
,V indications; antici-pated changes in plant conditions due to indications; minor lapses in antidipating predictable changes ations; failed to anticipate predict-able changes in plant events in progress status
[.20 X =
]
(c) Demonstrate, through directives and actions, a thorough UNDERSTANDING of how the PLANT, SYSTEHS, and COMPONENTS OPERATE AND INTERACT (including set points, interlocks, and automatic actions)? 3 2 1 Demonstrated thorough Minor errors because of Inadequate knowledge understanding of how gaps in knowledge of of system and com-systems and components how systems and ponent operation operate and interact components operate resulted in serious mistakes or plant degradation [.45 X -
]
O Examiner Standards 22 of 27 Rev. 7, January 1993
! ES-303 4 Form ES-303-4
- 4. COMPLIANCE WITH AND USE OF PROCEDURES DID THE APPLICANT: ;
(a) REFER to correct procedure: and procedural steps when appropriate? 3 2 1 Requested or readily Minor lapses in Failed to correctly located all appro- referring to or refer to procedures priate procedures locating appropriate in important as necessary procedures instances [.25 X -
]
l (b) USE PROCEDURES CORRECTLY, including following procedural steps in correct sequence, abiding by procedural cautions and limitations, , selecting correct paths on decisions blocks, and correctly transitioning t between procedures? 3 2 1 , l O Accurately and promptly executed procedural steps Minor errors, but made necessary corrections in Significant errors impeded or slowed recovery or degraded I a timely fashion plant unnecessarily
=
[.50 X ] (c) Ensure the safe, efficient IMPLEMENTATION of procedures BY THE CREW 7 l I 3 2 1 i Kept crew informed of Crew occasionally had Read procedures to l procedural status; got to question SR9 him/herself; failed acknowledgment from regarding statos; to coordinate or crew when reading allowed lapses in verify crew's use of procedures implementation by crew procedures [.25 X -
]
l l l
\--
Examiner Standards 23 of 27 Rev. 7, January 1993 i
,7 .
ES-303 5 Form ES-303-4
- 5. OPERATE THE CONTROL BOARDS DID THE APPLICANT:
(a) I.0CATE CONTROLS efficiently and accurately? 3 2 1 Promptly located _ Some minor hesitancy or Unable to locate ; appropriate controls difficulty in locating controls without in all instances controls assistance ! [.25 X =
]
(b) MANIPULATE CONTROLS in an accurate and timely manner? 3 2 1 1 Manipulations were Minor shortcomings, Improper manipul-consistently accurate but any resulting ations caused major and timely consequences were system perturbations
- readily mitigated A
[.25 X =
]
(c) ACT appropriately in response to INSTRUMENT READINGS? ) 3 2 1 l Responses were Generally responsive, Failed to react appropriate and but some minor appropriately to timely errors and lapses instrument readings without assistance [.25X =
]
4 (d) Take MANUAL CONTROL of automatic functions when appropriate? I 3 2 1 Took manual control Minor delays; some Depended on automatic ) , as appropr W prompting necessary actions; required ) before overriding prompting to take automatic functions manual control [.25 X =
]
)
.s
' Examinee Standards 24 of 27 Rev. 7, January 1993
i Form ES-303-4 Q ES-303 6
]
- 6. . COMUNICATE AND INTERACT WITH THE CREW AND OTHER PERSONNEL )
DID THE APPLICANT: ,
- (a) Communicate in a clear, easily-understood manner?
l 3 2 1 i-Communications were At times, communi- Communications were ; j timely, clear, and cations were ill-timed, vague , easy to hear and - confusing, hard to or difficult to J understand hear or understand hear or understand l [.45 X =
]
i (b) Keep crew members and those outside the control room informed of plant status? , 3 2 1 Provided others with Had to be prompted for Failed to provide ; accurate, pertinent information in some minor important information i
; information throughout instances; gave some ;
E . scenarios incomplete or ihaccurate j information - t [.35X =
] ]
(c) ENSURE RECEIPT of clear, easily-understood communications from crew and i others? ; 3 2 1 [ Requested information failed to require Failed to request , ! or clarification when or acknowledge inform- needed information; 1 necessary; understood mation from others inattentive when . communications from information was
- others provided; failed to >
correct serious I misunderstandings among crew members [.20 X =
.] !
e Examiner Standards 25 of 27 Rev. 7, January 1993
.c 4
i !
- _ _ _ _ _* .!__i.__, __ . . . . . . .,
- j V
- y. '
Q ES-303 7 Form ES-303-4 -
- 7. . DIRECT SHIFT OPERATIOui . y DID THE APPLICANT:
(a) Take TIMELY and DECISIVE ACTION when problems arose? i 3L 2 1 . Took early remedial Minor instances of Failed to take timely action when necessary -failure to take action- action; resulted in within a reasonable deterioration of period of time plant conditions [.20 X =
]
(b)~ Provide TIMELY, WELL THOUGHT OUT DIRECTIONS that facilitated CREW PERFORMANCE and demonstrated appropriate CONCERN for the SAFETY of the plant, staff, and'public? j 3 2 1 Directives enabled Minor instances of Directivcs inhibited < safe, integrated incorrect, trivial, safe performance-crew performance or difficult-to-carry- crew had to explain out orders why orders could not or should not be followed [.40 X =
]
(c) Stay in a position of OVERSIGHT and provide an APPROPRIATE AdOUNT of DIRECTION and GUIDANCE? 3 2 1 Stayed involved but Crew occasionally had to lost the big picture; not intrusive; request assistance, crew had to repeat-anticipated crew's which interfered edly request or needs and )rovided with their ability to provide guidance; guidance wien carry out actions failed to verify that necessary directives were correctly implemented [.20 X =
]
(d) SOLICIT and INCORPORATE FEEDBACK from the crew to foster an effective, team-oriented approach to problem solving and decision' making? 3 2 1
. Involved crew in At times, failed to Made decisions with-problem-solving process' . involve crew in decision out crew participa-as appropriate,= leading making when it would tion or consultation; to effective team have been appropriate, crew divisiveness was decision making. detracting from team- counter productive oriented approach
[.20 X =
]
g
~ Examiner Standards 26'of 27 Rev. 7, January 1993 cf :
7
ES-303- 8 Form ES-303-4
- 8. COMPLY WITH AND USE TECHNICAL-SPECIFICATIONS DID THE APPLICANT:
(a) RECOGNIZE when conditions were covered by technical specifications (TS)? 3 2 1 Recognized TS . Minor errors and Failed to correctly limiting conditions for misunderstand 1ngs recognize situations operation and action with respect to covered by TS and statements without use TS applications action statements of references J [.40 X -
]
(b) LOCATE the appropriate technical specifications quickly and efficiently? 3 2 1 Lncated applicable Had difficulty locating Could not locate l TS quickly and TS; had to search appropriate TS accurately through index and body of document lO [.20 X =
]
(c) Ensure correct COMPLIANCE with technical specifications and limiting i condition for operation action statements?
- 3. 2 1 )
1 Directives were based Needed some assistance Apr11ed incorrect TS l l on correct under- from crew to ensure to situation; standing of TS compliance allowed crew to J action statements violate TS [.40 X -
]
i Examiner Standards. 27 of 27 Rev. 7, January 1993 l-
I NUREG-1123 ep . 4 Cd Knowledge and Abilities Catalog for l\uclear Power Plant Operators: Boiling Water Reactors l 4 4 i
' CPMN-Manusce,t Completed: September 1986 Date Published: September 1986
- i j
Division of Human Factors Technology Office of Nuclear Reactor Regulation
, U.S. Nuclear Regulatory Ceinmission Washington, DC 20555 !
l N\Yf] s...../ i i i I
\
'~'
1 ORGANIZATION OF THE BWR CATALOG ;
~~a_, ,
The Knowledge and Abilities Catalog for Nuclear Prser Plant Operators: Boil- i ing Water Reactors.is organized into five major etions. Knowledge and abil- , ity statements (K/As) are grouped according to the major section to which they pertain. .lhis organization is shown' schematically below. l t PLANT-WIDE GENERIC KNOWLEDGE AND ABILITIES (33) PLANT SYSTENS (54) Knowledge Categories (K1 - K6) i Ability Categories (Al - A4) System-wide Generics (15) . EMERGENCY AND A8 NORMAL PLANT EVOLUTIONS (38). ' Emergency Plant Evolutions Abnormal Plant Evolutions ' Knowledge Categories (E/A K1 - E/A K3) Ability Categories (E/A A1 - E/A A2) System-wide Generics (12) , COMPONENTS Component Categories (8) f -'s Knowledge Abilities ( 2 THEORY Reactor Theory Categories (8) Knowledge Abilities Thermudynamics Categories (10) Knowledge Abilities 1,1 Plant-wide Generic Knowledge and Abilities A group of 33 knowledge and abilities has been identified as generic to all plants (e.g.,, knowledge of safety proc edures related to high pressure). These are generally administrative knowledge and abilities with broad applica-bility across systems and/or plants. They are listed in Section 2 of this BWR catalog. I' ( K/A Catalog: BWR 1-1 i
, r 1 ORGANIZATION OF THE BWR CATALOG The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boil-ing Water Reactors is organized into five major sections. Knowledge and abil-ity statements (K/As) are grouped according to the major section to which they pertain. This organization is shown schematically below.
PLANT-WIDE GENERIC KNOWLEDGE AND ABILITIES (33) PLANT SYSTEMS (54) Knowledge Categories (K1 - K6) Ability Categories (Al - A4) System-wide Generics (15) ENERGENCY AND ABNORNAL PLANT EVOLUTIONS (38) Emergency Plant Evolutions ; Abnormal Plant Evolutions Knowledge Categories (E/A K1 - E/A K3) Ability Categories (E/A A1 - E/A A2) System-wide Generics (12) - CONPONENTS Component Categories (8) (q Knowledge Abilities THEORY Reactor Theory Categories (8) Knowledge Abilities Thermodynamics Categories (10) Knowledge Abilities ! 1.1 Plant-wide Generic Knowledge and Abilities 4 A group of 33 knowledge and abilities has been identified as generic to all plants (e.g.,, knowledge of safety procedures related to high pressure). These are generally administrative knowledge and abilities with broad applica-bility across systems and/or plants. They are listed in Section 2 of this BWR catalog. 1 "t i i V... K/A Catalog: BWR 1-1
1.2 Plant Systems Major safety functions must be maintained to ensure safe nuclear power plant operation. The nine safety functions required for a BWR plant are: Reactivity Control
. Reactor Water Inventory Control Reactor Pressure Control Heat Removal From Reactor Core Containment Integrity Electrical Instrumentation Plant Service Systems Radioactivity Release the ir Fif ty-four plant systems have been included in the BWR Catalog based on relationship and importance to the safety functions. Table I contains a list of these plant systems by safety function. It should be noted that ten plant systems each contribute to two safety functions. Each plant system has a six-digit code number. The first three digits are the same as those used by INP0 to identify SYSTEM / DUTY areas. See pages 1-13 to 1-15 for an alphabeti-cal and numerical listing of the plant systems included in the BWR Ca tal og.
See Section 3 of the BWR Catalog for the delineations of the K/As for the plant systems. Table 1 Plant Systems by Safety Function Safety Function I: Reactivity Control 201001 Control Rod Drive Hydraulic System 201003 Control Rod and Drive Mechanism 201002 Reactor Manual Control System 202002 Rec irculation Flow Control System 202001 Recirculation System 201005 Rod Control and Information System (RCIS) 211000 Standby Liquid Control System Safety Function II: Reactor Water Inventory Control 206000 High Pressure Coolant Injection System 209002 High Pressure Core Spray System (HPCS) 209001 Low Pressure Core Spray System 256000 Reactor Condensate System 217000 Reactor Core Isolation Cooling System (RCIC) 259001 Reactor Feedwater System 204000 Reactor Water Cleanup System 259002 Reactor Water Level Control System 203000 RHR/LPCI: Injection Mode (Plant Specific) O K/ A Catalog: BWR 1-2
l i r [ Safety Function III: Reactor Pressure Control . 218000 Automatic Depressurization System 239001 Main and Reheat Steam System 241000 Reactor / Turbine Pressure Regulating System 239002 Relief / Safety Valves Safety Function IV: Heat Removal From Reactor Core 206000- High Pressure Coolant Injection System 209002 High Pressure Core Spray System (HPCS) 207000 Isolation (Emergency) Condenser 209001 Low Pressure Core Spray System 239001 Main and Reheat Steam System 245000 Main Turbine Generator and Auxiliary Systems 217000 Reactor Core Isolation Cooling System (RCIC) l 202001 Recirculation System 203000 RHR/LPCI: Injection Mode (Plant Specific) 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) Safety Function V: Containment Integrity 223001 Primary Containment System and Auxiliaries 223002 Primary Containment Isolation System / Nuclear Steam Supply Shut-Off ' 290002t Reactor Vessel Internals
/ 219000 RHR/LPCI: Torus / Suppression Pool Cooling Mode t 226001 RHR/LPCI: Containment Spray System Mode i 230000 RHR/LPCI: Torus / Suppression Pool Spray Mode 1 290001 Secondary Containment Safety Function VI: Electrical 262001 A.C. Electrical Distribution 263000 D.C. Electrical Distribution 264000 Emergency Generators (Diesel / Jet) i 262002 Uninterruptable Power Supply (A.C./D.C.)
Safety Function VII: Instrumentation i 215005 Average Power Range Monitor / Local Power Range Monitor System 215003 Intermediate Range Monitor (IRM) System 216000 Nuclear Boiler Instrumentation i 272000 Radiation Monitoring System- ; 212000 Reactor Protection System 215002 Rod Block Monitor System 201005 Rod Control and Information System (RCIS) ; 214000 Rod Position Information System 201004 Rod Sequence Control System (Plant Specific) 201006 Rod Worth Minimizer System (RWM) (Plant Specific) 215004. Source Range Monitor (SRM) System 215001- Traversing In-Core Probe , tThis system number does not correspond to an INP0 SYSTEM /0UTY area
~
number. K/A Catalog: BWR 1-3
Safety Function VIII: Plant Service Systems O 286000 Fire Protection System 234000 Fuel Handling Equipment Safety Function IX: Radioactivity Release 239003 MSIV Leak--- Control System ' 271000 Offgas System 288000 Plant Ventilation Systems 272000 Radiation Monitoring System 268000 Radwaste 290002t Reactor Vessel Internals 233000 Fuel Pool Cooling and Clean-up 261000 Standby Gas Treatment System 290003t Control Room Heating, Ventilating and Air Conditioning tThis system number does not correspond to an INP0 SYSTEM /0VTY area number. O O l K/A Catalog: 8WR 1-4 I i I
)) 1.2.1 Plant System Definitions The BWR Catalog defines the plant systems listed below as including:
Primary Containment System and Auxiliaries (223001) Drywell Hydrogen Recombiners- ' Torus ACAD/ CAM Suppression Pool Vacuum Breakers Ven til ation Combustible Gas Control Ventilation and Purge Drywell Coolers Hydrogen Igniters Drywell Chillers Main Turbine Generators and Auxiliaries (245000) Turbine Generators Main Tube Oil Turbine Drains Hydrogen Seal Oil Hydrogen Gas Cooling Stator Water Cooling Voltage Regulation Moisture Separators Rod Control and Infonnation System (201005) Rod Position Information System Rod Action Control System o Rod Pattern Controller Rod Gang Drive System Rod Interface System Reactor Condensate System (256000) Air Ejectors Condenser Low Pressure Feed Heaters Extraction Steam Heater Level Control Condensate Makeup and Eject Condensate System Reactor Feedwater System (259001) Feedwater System Feedwater Regulating Valves Reactor Feed Pump (Turbine and Motor as applicable) High Pressure Feed Heaters Extraction Steam High Pressure Feed Heater Level Control Reactor Vessel and Internals (290002) Fuel Reactor Vessel Internal Vessel Components RHR/LPCI: (System) Mode The convention "RHR/LPCI:" is used to refer to a particular sys-(o'y[_
) tem function regardless of the mode or system's relationship to RHR or LPCI.
K/A Catalog: BWR 1-5
m 1.2.2 Knowledge and Ability Stem Statements for Plant Systems The information delineated within each plant system is organized into six dif-ferent types of knowledge and four different types of ability. Iforthere are ability no knowledge or ability statements following one of the knowledge The knowledge stem statements, it indicates that no statements were rated. and ability stem statements for the plant systems are listed in Table 2. Table 2 Knowledge and Ability Statments for Plant Systems Ki Knowledge Stem Statement Kl. Knowledge of the physical connections and/or cause-effect r(lationships between (SYSTEM) and the following: K2. Knowledge of electrical power supplies to the following: K3. Knowledge of the effect that a loss or malfunction of the (SYSTEM) will have on following: K4. Knowledge of (SYSTEM) design feature (s) and/or interlocks which provide for the following: K5. Knowledge of the operational applications of the following concepts as they apply to (SYSTEM): K6. Knowledge of the effect that a loss or malfunction of the following will have on the (SYSTEM): Al Ability Stem Statement A1. Ability to predict and/or monitor changes in parameters associated with operating the (SYSTEM) controls including: A2. Ability to (a) predict the impacts of the following on the (SYSTEM) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A3. Ability to monitor automatic operations of the (SYSTEM) including: A4. Ability to manually operate and/or monitor in the control room: O l K/A Catalog: BWR 1-6 l
, ~. . . - - - -
1.2.3 System-wide Generic Knowledge and Abilities for Plant Systems
.n.
Fifteen knowledge and abilities have been identified as generic to all sys-1 (,2g ( tems. They are generally administrative in nature. The fif teen system-wide , generic K/As are repeated at the end of each plant system delineation in the BWR Catalog. Table 3 contains a list of the fifteen system-wide generic (SG) K/As. Table 3 ! System-wide Generic Statements for Plant Systems SG# Statement
- 1. Knowledge of operator responsibilities during all modes of plant opera-tion. ,
- 2. Knowledge of system status criteria which require the notification of plant personnel.
- 3. Knowledge of which events related to system operation / status should be ,
reported to outside agencies.
- 4. Knowledge of system purpose and/or function.
- 5. Knowledge of limiting conditions for operations and safety limits.
O 6. Knowledge of bases in technical specifications for limiting conditions Q for operations and safety limits.
- 7. Knowledge of purpose and function of major system components and con-trols.
- 8. Knowledge of the annunciator alarms and indications, and use of the response instructions.
- 9. Ability to locate and operate components, including local controls.
- 10. Ability to explain and apply all system limits and precautions.
- 11. Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.
- 12. Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
- 13. Ability to perform specific system and integrated plant procedures dur-ing all modes of operation.
- 14. Ability to perform without reference to procedures those actions that require immediate operation of system components or controls.
- 15. Ability- to recognize abnonnal indications for system operating parame-
.O ters which are entry-level conditions for emergency and abnormal operat-Q _..
ing procedures. K/A Catalog: BWR 1-7
l 1.3 Emergency and Abnormal Plant Evolutions Section 4 of the BWR Catalog contains 15 emergency plant evolutions and 23 abnormal plant evolutions. The listing of emergency and abnormal plant evolu-tions was developed to include those integrative situations crossing several plant systems and/or safety functions. (Knowledge and abilities related to an abnormal situation within one specific plant system are delineated within that system and are incluceu I7 the plant systems section of the BWR Catalog.) An emergency plant evolution is a_ny n condition, event or symptom whir.h leads to entry into Emergency Procedure Guidelines (EPGs). Three broad areas of emer-rad ioac-gency evolutions relate to reactor cor. trol, containment control, and tivity releases. An abnormal e,.Mion is any degraded condition, event or symptom not directly leading to an EPG entry condition nor related to an operational condition as: power operation, start-up, hot shut down, cold shut down, and refuel. It is recognized that for each condition there are degrees of severity. The EPG entry guidelines were used as the bases of classifying each condition Any as either an abnormal plant evolution or an emergency plant evolution. abno rmal condition which so degrades as to threaten plant safety will result in entry into the EPGs and at that time be treated as an emergency condition. Table 4 contains a list of the emergency and abno rmal plant evolutions included in the BWR Catalog. Each evolution has a unique six digit code number. Table 4 Emergency and Abnormal Plant Evolutions Emergency Plant Evolutions 295027 digh Containment Temperature ; Mark Ill Containment Only) 295024 High Drywell Pressure 295028 High Drywell Temperature 295038 High Off-Site Release Rate 295025 High Reactor Pressure 295033 High Secondary Containment Area Radiation Levels 295032 High Secondary Containment Area Temperature 295029 High Suppression Pool Water Level 295030 Low Suppression Pool Water Level 295031 Reactor Low Water Level 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 295035 Secondary Containment High Differential Pressure 295036 Secondary Containment High Sump / Area Water Level 295034 Secondary Containment Ventilation High Radiation 295026 Suppression Pool High Water Temperature Abnormal Plant Evolutions 295016 Control Room Abandonment 295011 High Containment Temperature (Mark Ill Containment Only) 295010 High Drywell Pressure K/A Catalog: BWR 1-8
) Abnonnal Plant Evolutions (Continued)
(Lum ~Q. 295012 High Drywell Temperature 295017 High Off-Site Release Rate 295007 High Reactor Pressure ' 295008 High Reactor Water Level 295013 High Suppression Pool Temperature 295020 Inadvertent Containment Isolation 295014 Inadvertent Reactivity Addition 295015 Incomplete SCRAM 295022 Loss of CR0 Pumps 295002 Loss of Main Condenser Vacuum 295021 Loss of Shutdown Cooling 295009 Low Reactor Water Level 295005 Main Turbine Generator Trip 295003 Partial or Complete Loss of A.C. Power 295018 Partial or Complete Loss of Component Cooling Watert
~295004 Partial or Complete Loss of D.C. Power 295001 Partial or Complete Loss of Forced Core Flow Circulation 1 295019 Partial or Complete Loss of Inst.ument Air 295023 Refueling Accidents >
l 295006 SCRAM ) TThe BWR Catalog defines component cooling water as including: reactor build-ing component cooling water, nuclear component closed cooling water, service
- [, water, nuclear service water, turbine building closed cooling water system, main condenser circulating water, emergency service water, containment closed j cooling water system, and RHR service water. '
1.3.1 Knowledge and Ability Statements for Emergency and Abnormal Plant Evolutions. The emergency and abnonnal plant evolution knowledge and ability statements are organized into three knowledge types and two ability types. If there are no knowledge or ability statements following a knowledge or ability stem j statement, it indicates that no statements were rated in that category. The i knowledge and ability stem statements for the Emergency and Abnormal Plant Evolutions are listed in Table 5. Table 5 Knowledge and Ability Sten Statements for Emergency and Abnormal Plant Evolutions l E/A Ki Knowledge Stem Statement E/A Kl. Knowledge of the operational implications of the following conc epts
- as they apply to (EMERGENCY OR, ABNORMAL PLANT EVOLUTION):
k E/A K2. Knowledge of the interrelations between (EMERGENCY OR ABNORMAL PLANT m EVOLUTION) and the following: r (d E/A K3. Knowledge of the reasons for the following responses as they apply to (EMERGENCY OR ABNORMAL PLANT EVOLUTIQN): K/A Catalog: BWR 1 l
E/A Af Ability Stem Statement (Continued) _ n E/A A1. Ability to operate and/or monitor the following as they apply to ( EMERGENCY AND ABNORMAL PLANT EVOLUTION): E/A A2. Ability to determine and/or interpret the following as they apply to ( EMERGENCY AND ABNORMAL PLANT EVOLUTION): 1.3.2 System-wide Generic Knowledge and Abilities for Emergency and Abnormal Plant Evolutions Twelve knowledge and abilities have been identified as generic to all emer-gency and abnormal plant evolutions. These system-wide generic statements are repeated at the end of each emergency and abnonnal plant evolution in the BWR Ca tal og. Table 6 contains a list of the system-wide generics. Table 6 System-wide Generic Statements for Emergency and Abnormal Plant Evolutions SGf Statement
- 1. Knowledge of system status criteria which require the notification of plant personnel.
- 2. Knowledge of which events related to system operation / status should be reported to outside agencies.
- 3. Knowledge of limiting conditions for operations and safety limits.
- 4. Knowledge of bases in technical specifications for limiting cond itions for operations and safety limits.
- 5. Knowledge of the annunciator alarms and indications, and use of the response instructions,
- 6. Aisility to locate and operate components, including local controls.
- 7. Ability to explain and apply all system limits and precautions. )
- 8. Ability to recognize indications for system operating parameters which '
are entry-lvel conditions for technical specifications.
- 9. Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
- 10. Ability to perform without reference to procedures those actions that require immediate operation of system components or controls.
- 11. Ability to recognize abnonnal indications for system operating parameters !
which are entry-level conditions for emergency and abnonnal operating ; proced ures. I
- 12. Ability to utilize symptom based procedures.
K/A Catalog: BWR 1-10
i
- 1. 4 Components m '
Basic components such as valves and pumps are found in many systems. The fol-lowing eight categories of components, for which additional K/As sre presented, are delineated in Section 5 of the BWR Catalog. These additional l K/As are more detailed or specific than those appropriate for system listings, ! yet at the same time they are generic to the component types. Each component - 1 type has a unique six-digit code number. 291001 Valves 291002 Sensors / Detectors 291003 Controllers and Positioners Pumps l 291004 l 291005 Motors and Generators 291006 Heat Exchangers and Condensers 291007 Demineralizers and Ion Exchangers 291008 Breakers, Relays and Disconnects 1.5 Theory Fundamental theoretical knowledge which underlies safe performance on the job is delineated in Section 6 of the BWR Catalog. These K/As represent basic concepts without which other knowledge or abilities related to the operational , .[mC/) implications of plant operation could not be mastered. Each theory topic has a unique six-digit code number. i Reactor Theory 292001 Neutrons 292002 Neutron Life Cycle ! 292003 Reactor Kinetics and Neutron Sources l 292004 Reactivity Coefficients i 292005 Control Rods 292006 Fission Product Poisons l 292007 Fuel Depletion and Burnable Poisons 292008 Reactor Operational Physics Thermodynamics 293001 Thermodynamic Units and Properties 293002 Basic Energy Concepts 293003 Steam 293004 Thermodynamic Processes 293005 Thermodynamic Cyclas 293006 Fluid Str. tics , 293007 Heat Transfer and Heat Exchanges y 293008 Thermal H/draulics ! f T 293009 Core Thermal Limits V ... 293010 Brittle Fracture and Vessel Thermal Stress K/A Catt. log: BWR 1-11
1.6 Knowledge and Ability Statements q 1.6.1 Numbering Each kno.wledge listed in the BWR Catalog is identified by two characters which reference the knowledge category, and two digits which reference the number of the statement within the category. For example, K2.03 references the third statement in the second knowledge ca tegory. Similarly, the first ability category is identif'" as A1. To distinguish emergency and abnormal plant evolution K/As from those for plant systems, an E (emergency) or an A (abnor-mal) has been placed before the knowledge (E/A K1) or ability (E/A A1) charac-ters. 1.6.2 Importance Ratings Importance ratings of the K/As are given for R0s and for SR0s next to each knowledge and ability statement in the BWR Catalog. These ratings reflect the average ratings of individual NRC/ utility panel members. The rating scale is presented in Table 7. Table 7 R0 and SR0 Importance Ratingst Rating Importance for Safe Operation 5 essential 4 very important 3 fairly important 2 of limited importance 1 insignificant importance
*/? indicates variability in the responses tImpor'ance includes direct and indirect impact of the K/A on safe plant operations in a manner ensuring personnel and public health and safety. 'Therefore, the raf g of 2.0 or below represents a statement of limited or insignificant irrgartance for the safe operation of a plant. Such statements are generally considered as inappropriate content for NRC licensing examina-tions. (See Sections 1.6.3 and 1.6.4 below for qualifications of importance ratings related to variability of the ratings and plant specific data.)
1.6.3 Asterisk and Question Ratings Some importance ratings are followed by an asterisk or question mark. These indicate variability in the rating responses. An asterisk indicates the rat-ing spread was very broad. An asterisk can also signify that more than 15 percent of the raters indicated that the knowledge or ability is not required for the R0 or SR0 position at their plant because it is the responsibility of A question mark indicates that more than 15 l someone else (e.g., SRO vs. RO). > percent of the raters felt that they were not familiar with the knowl edge or ability as related to the particular system or design feature. These marks indicate the need for examiners to reyiew plant-specific materials to deter-mine whether or not that knowledge or ability is indeed appropriate for inclu-sion in any given examination. l l K/A Catalog: BWR 1-12 .
i
/ %
l 1.6.4 Plant Specific Data Some K/A statements apply only to specific BWR plants with applicable' design , features. These statements appear with h colon (:) and are followed by the i
. plant identifying information. For example, if a statement applies to con- '
tainment, it may refer only to plants with Mark I, II, or III containment j vessels. The statement would then be followed by ": Mark I, II, or III". ; Some statements are followed by ": Plant Specific" indicating that' they do : not relate to any particular plant design nor to a particular containment I structure. However, the knowledge or ability does not apply to all plants. 1.6.5 Difference ratings A dagger (t) to the left of an individual knowledge or ability statement indi-cates that more than 20% of the raters indicated that the level of knowledge or ability required by en SR0 is different than the level of knowledge or l ability required by an R0. I 1.7 List of Plant Systems Included in the *WR Catalog i 1.7.1 Alphabetical List of Plant Systems Page N_o. l 262001 A.C. Electrical Distribution 3.6-1
, 218000 Automatic Depressurization System 3.3-1
'\ 215005 Average Power Range Monitor / Local Power Range 3.7-1 Monitor System 201001 Control Rod Drive Hydraulic System 3.1-1 201003 Control Rod and Drive Mechanism 3.1-7 290003 Control Room HVAC 3.9-37 263000 D.C. Electrical Distribution 3.6-5
, 264000 Emergency Generators (Diesel / Jet) 3.6-9 286000 Fire Protection System 3.8-1 234000 Fuel Handling Equipment 3.8-5 233000 Fuel Pool Cooling and Clean-up 3.9-25 ;
206000 High Pressure Coolant Injection System 3.2-1 ! 209002 High Pressure Core Spray System (HPCS) 3.2-7 1 215003 Intermediate Range Monitor (IRM) System 3.7-7 207000 Isolation (Emergency) Condenser 3.4-1 209001 Low Pressure Core Spray System 3.2-13 239003 MSIV Leakage Control System 3.9-1 245000 Main Turbine Generator and Auxiliary Systems 3.4-7 239001 Main and Reheat Steam System 3.3-5 216000 Nuclear Boiler Instrumentation 3.7-11 271000 . Of fgas System 3.9-5 288000 Plant Ventilation Systems 3.9-11 223002 Primary Containment Isolation System / Nuclear -3.5-7 Steam Supply Shut-Off 223001 Primary Containment System and Auxiliaries 3.5-1 226001 RHR/LPCI: Containment Spray System Mode 3.5-23 O 203000 - RHR/LPCI: Injection Mode (Plant Specific) 3.2-49 tj 219000 230000 RHR/LPCI: Torus / Suppression Pool Cooling Mode 3.5-17 _. RHR/LPCI: Torus / Suppression Pool Spray Mode 3.5-29 272000 Radiation Monitoring System 3.9-15 K/A Catalog: BWR 1-13 l
m 268000 Radwaste 3.9-21 256000 Reactor Condensate System 3.2-19 217000 Reactor Core Isolation Cooling System (RCIC) 3.2-25 259001 Reactor Feedwater System 3.2-31 201002 Reactor Manual Control System 3.1-11 212000 Reactor Protection System 3.7-17 290002 Reactor Vessel Internals 3.5-13 204000 Reactor Water Cleanup System 3.2-37 259002 Reactor Water Level Control System 3.2-43 241000 Reactor / Turbine Pressure Regulating System 3.3-11 202002 Recirculation Flow soctrol System 3.1-15 202001 Recirculation Systs 3.1-19 239002 Relief / Safety Valves 3.3-19 215002 Rod Block Monitor System 3.7-23 201005 Red Control and Information System (RCIS) 3.1-27 214000 Rod Position Information System 3.7-27 201004 Red Sequence Control System (Plant Specific) 3.7-31 201006 Rod Worth Minimizer System (RWM) (Plant Specific) 3.7-35 290001 Secondary Containment 3.5-35 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) 3.4-13 215004 Source Range Monitor (SRM) System 3.7-39 261000 Standby Gas Treatment System 3.9-31 211000 Standby Liquid Control System 3.1-31 215001 Traversing In-Core Probe 3.7-43 262002 Uninterruptable Power Supply (A.C./D.C. ) 3.6-15 l 1.7.2 Numerical List of Plant Systems Page No. 201001 Control Rod Drive Hydraulic System 3.1-1 l 201002 Reactor Manual Control System 3.1-11 201003 Control Rod and Drive Mechanism 3.1-7 l 201004 Rod Sequence Control System (Plant Specific) 3.7-31 ! 201005 Rod Control and Information System (RCIS) 3.1-27 l 201006 Rod Worth Minimizer System (RWM) (Plant Specific) 3.7-35 202001 Recirculation System 3.1-19 l 202002 Recirculation Flow Control System 3.1-15 203000 RHR/LPCI: Injection Mode (Plant Specific) 3.2-49 204000 Reactor Water Cleanup System 3.2-37 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) 3.4-13 206000 High Pressure Coolant Injection System 3.2-1 207000 1 solation (Emergency) Condenser 3.4-1 209001 Low Pressure Core Spray System 3.2-13 209002 Hign Pressure Core Spray System (HPCS) 3.2-7 211000 Standby Liquid Control System 3.1-31 212000 Reactor Protection System 3.7-17 , 214000 Rod Position Infonnation System 3.7-27 ; 215001 Traversing In-Core Probe 3.7-43 215002 Rod Block Monitor System 3.7-23 215003 Intermediate Range Monitor (IRM) System 3. 7 -7 l 215004 Source Range Monitor (SRM) System 3.7-39 ! 215005 Average Power Range Monitor / Local Power Range 3.7-1 1 Monitor System 216000 Nuclear Boiler Instrumentation 3.7-11 K/A Catalog: BWR 1-14
,~.
C 217000 Reactor Core _ Isolation Cooling System (RCIC) 3.2-25 218000 Automatic Depressurization System 3.3-1 219000 RHR/LPCI: Torus / Suppression Pool Cooling Mode 3.5-17 223001 Primary Containment System and Auxiliaries 3.5-1 223002 Frimary Containment Isolation System / Nuclear 3.5-7 Steam Supply Shut-Off 226001 RHR/LPCI: Containment Spray System Mode 3.5-23 230000 RHR/LPCI: Torus / Suppression Pool Spray Mode- 3.5-29 233000 Fuel Pool Cooling and Clean-up 3.9-25 234000 Fuel Handling Equipment 3. 8 -5 239001 Main and Reheat Steam System 3.3-5 239002 Relief / Safety Valves 3.3-19 239003 MSIV Leakage Control System 3. 9-1 241000 Reactor / Turbine Pressure Regulating System 3.3-11 245000 Main Turbine Generator and Auxiliary Systems 3.4-7 256000 Reactor Condensate System 3.2-19 259001 Reactor Feedwater System 3.2-31 259002 Reactor Water Level Control System 3.2-43 261000 Standby Gas Treatment System 3.9-31 262001 A.C. Electrical Distribution 3.6-1 262002 Uninterrupta'ble Power Supply ( A.C./D.C. ) 3.6-15 263000 D.C. Electrical Distribution 3. 6 - 5 264000 Emergency Generators (Diesel / Jet) 3.6-9 268000 Radwaste 3.9-21 O 271000 Offgas System 3.9-5 (j 272000 Radiation Monitoring System 3.9-15 286000 Fire Protection System 3.8-1 288000 Plant Ventilation Systems 3.9-11 290001 Secondary Containment 3.5-35 290002 Reactor Vessel Internals 3.5-13 290003 Control Room HVAC 3.9-37 1.8 List of Emergency and Abnormal Plant Evolutions 1.8.1 Numerical List of Emergency Plant Evolutions Pag e N_o o. 295024 High Drywell Pressure 4.1-1 295025 High Reactor Pressure 4.1-5 295026 Suppression Pool High Wate~ Temperature 4.1-9 295027 High Containment Temperature (Mark III Containment Only) 4.1-11 295028 High Drywell Temperature 4.1-13 295029 High Suppression Pool Water Level 4.1-15 295030 Low Suppression Pool Water Level 4.1-17 295031 Reactor Low Water Level 4.1-21 295032 High Secondary Containment Area Temperature 4.1-25 295033 High Secondary Containment Area Radiation Levels 4.1-29 295034 Secondary Containment Ventilation High Radiation 4.1-31 n 295035 Secondary Containment High Differential Pressure 4.1-33 i 295036 Secondary Containment High Sump / Area Water Level 4.1-35 {Q 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 4.1-37 295038 High Off-Site Release Rate 4.1-41 K/A Catalog: BWR 1-15
Page No. O 1.8.2 Numerical List of ,'bnormal Plant Evolutions Partial or Complete Loss of Forced Core Flow Circulation 4.2-1 295001 4.2-5 295002 Loss of Main Condenser Vacuum 4.2-9 295003 Partial or Complete Loss of A.C. Power Partial or Complete Loss of D.C. Power 4.2-13 295004 4.2-15 295005 Main Turbit,. ;enerator Trip 4.2-19 295006 SCRAM 4.2-23 295007 High Reactor Pressure 4.2-25 295008 High Reactor Water Level 4.2-29 295009 Low Reactor Water Level 4.2-31 295010 High Drywell Pressure 4.2-35 295011 High Containment Temperature (Mark III Containment Only) 4.2-37 295012 High Drywell Temperature 4.2-39 295013 High Suppression Pool Temperature 4.2-41 295014 Inadvertent Reactivity Addition 4.2-45 295015 Incomplete SCRAlt 4.2-49 295016 Control Room Abandonment 4.2-51 295017 High Off-Site Release Rate Partial or Complete Loss of Component Cooling Water 4.2-55 295018 4.2-59 295019 Partial or Complete Loss of Instrument Air 4.2-63 295020 Inadvertent Containment Isolation 4.2-67 295021 Loss of Shutdown Cooling Loss of CRD Pumps 4.2-69 295022 4.2-71 295023 Refueling Accidents 0 1 i O' K/A Catalog: BWR 1-16
Emergency / Abnormal Plant Evolutions EPE: 295037 SCRAM Condition Present and Reactor Power Above APRM Downstale or Unknown INP0
REFERENCE:
2P*00501 Failure of reactor to SCRAM when required or failure of control rods to fully insert during SCRAM IMPORTANCE K/A NO. KNOWLEDGE R0 SR0 EKl. Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER AB0VE APRM DOWNSCALE OR UN' NOWN : EKl.01 Reactor pressure effects on reactor power............. 4.1
- 4.3*
EK1.02 Reactor water l evel ef fec t s on reactor power. . . . . . . . . . 4.1
- 4.3*
EK1.03 Boron effects on reactor power (SBLC)................. 4. 2 4.4* EK1.04 Hot shutdown boron weight: Plant-Specific............ 3.4 3.6 EK1.05 Cold shatdown boron weight: Plant-Specific........... 3.4 3.6 ' EK1.06 Cooldown effects on reactor power..................... 4.0* 4.2* EK1.07 Shutdown margin....................................... 3.4 3.8 ,m, . V) 6 EK2. Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER AB0VE APRM 00WNSCALE OR UNKNOWN and the following: EK2.01 RPS................................................... 4.2* 4.3* ; EK 2. 02 RRCS: Plant-Specific................................. 4. 0 4.2 i EK2.03 ARI/RPT/ATWS: Plant-Specific......................... 4.1 4.2* EK2.04 SBLC system........................................... 4.4* 4.5* i EK2.05 CRD hydraulic system.................................. 4. 0 4.1 ! EK2.06 CRD mechanisms........................................ 3.5 3.6 ) EK 2.0 7 Neutron monitoring system............................. 4. 0* 4.0 EK 2. 08 SPDS/ERIS/CRIDS/ GDS: Pl an t-Spec i f i c . . . . . . . . . . . . . . . . . . 2. 7 3.1 ! EK 2. 09 Reactor water level................................... 4.0 4.2 EK 2.10 Reactor pressure...................................... 3.8 4.1 EK 2.11 RMCS: Plant-Specific................................. 3. 8 3.9 i EK 2.12 Rod control and information system: Plant-Speci f ic. . . 3. 6 3.8 ! EK 2.13 Alternate boron injection methods: Plant-Specific.... 3. 4 4.1 1 EK 2.14 RPIS: Pl a n t -S pe c i f i c . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6 3.9 l EK3. Knowledge of the reasons for the following responses i as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER AB0VE APRM 00WNSCALE OR UNKNOWN : l 1 EK3.01 Recirculation pump trip / runback: Plant-Specific...... 4.1 4. 2 EK3.02 SBLC injection........................................ 4.3* 4.5* ['s EK3.03 EK3.04 Lower ing re ac tor wa te r l evel . . . . . . . . . . . . . . . . . . . . . . . . . . Hot shutdown boron weight: Plant-Specific............ 4.1* 3.2 4.5* 3.7 (
*~/ < l l
l K/A catalog: BWR 4.1-37 l l
i l Emergency / Abnormal Plant Evolutions q EPE: 295037 SCRAM Condition Present and Reactor Power Abcve iPRM Downscale or Unknown IMPORTANCE R0 SRO K/A NO. KNOWLEDGE
)
EK3.05 Cold shutdown boron weight: Plant-Specific........... 3.2 3. 7 EK3.06 Maintaining heat sinks external to the containment.... 3.8 4.1 EK3.07 Various alternate methods of control rod insertion: P l a n t - S pe c i f i c . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4. 2 4.3* EK3.08 ATWS circuitry: Plant-Specific....................... 3.6* 3.9* ABILITY EA1. Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR P'WER AB0VE APRM DOWNSCALE OR UNKNOWN : EA1.01 Reactor Prv ection System............................. 4.6* 4.6* EA1.02 RRCS: Plant.-Specific................................. 3. 8 4.0 EA1.03 ARI/RPT/ATWS: Plant-Specific......................... 4.1* 4.1* EA1.04 SBLC.................................................. 4.5* 4.5* EA1.05 CRD hydraulics systems................................ 3.9 4.0 EA1.06 Neutron monitoring system............................. 4.1
- 4.1 EA1.07 RMCS: Plant-Specific................................. 3. 9 4.0 EA1.08 Rod control and information system: Plant-Specific... 3. 6 3.6 EA1.09 SPDS/ERIS/CRIDS/ GDS: Pl an t-Spec i fi c . . . . . . . . . . . . . . . . . . 2.8* 3.0 EA1.10 Alternate boron injection methods: Plant-Specific.... 3.7 3.9 EA1.11 PCIS/NSSSS............................................ 3.5 3.6 EA2. Ability to determine and/or interpret the following as they aply to SCRAM CONDITION PRLSENT AND REACTOR POWER ABOYE APRM 00WNSCALE OR UNKNOWk :
EA2.01 Reactor pa;e ......................................... 4.2* 4.3* EA2.02 Reactor water 1evel................................... 4.1
- 4.2*
EA2.03 SBLC tank level....................................... 4.3* 4.4* EA2.04 Suppre ss ion pool temperat ure. . . . . . . . . . . . . . . . . . . . . . . . . . 4.0* 4.1* EA2.05 Control rod position.................................. 4.2* 4.3* EA2.06 Reactor pressure...................................... 4. 0 4.1 EA2.07 Conta inment condi tions/isolations. . . . . . . . . . . . . . . . . . . . . 4.0 4.2* SYSTEM GENERIC K/As
- 1. tKnowledge of system status criteria which require the notification of plant personnel. 3.3 4. 2
- 2. tKnowledge of which events related to system operation / status should be reported to outside agencies. 3.2* 4.6*
O K/A catalog: BWR 4.1-38
Emergency / Abnormal Plant Evolutions EPE: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown-IMPORTANCE SYSTEM GENERIC K/As R0 SR0
- 3. tKnowledge of limiting conditions for operations and safety limits. 3. 4 4.3*
- 4. tKnowledge of bases in technical specifications for limiting conditions for operations and safety limits. 3.1 4.2*
- 5. Knowledge of the annunciator alarms and indications, and use of the response instructions. 3. 9 3.8
- 6. Ability to locate and operate components, including local controls. 4.2* 4.1*
- 7. Ability to explain and apply all system limits and precautions. 3. 7 3.9 L
- 8. tAbility to recognize indications for system operating p.. parameters which are entry-level conditions for technical
('v'j specifications. 3.4 4.3*
- 9. Ability to verify system alarm setpo.ints and operate controls identified in the alarm response manual. 4.1 3. 9
- 10. Ability to perform without reference to procedures those actions that require immediate operation of system components or controls. 3.9* 3.8*
- 11. tAbility to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. 4.4* 4.7*
- 12. tAbility to utilize symptom based procedures. 3.9* 4.6*
(~ _
\.y/
K/A catalog: BWR 4.1-39 4
- - .. .. - . ~ .
i ODtA
\ "/
Emergency / Abnormal Plant Evolutions APE: 295015 Incomplete SCRAM INP0
REFERENCE:
2000100501 Control rod drive system failure (high SCRAM discharge header volume water level) 2000200501 Failure of reactor to SCRAM when required or failure of control rods to fully insert during SCRAM
- IMPORTANCE K/A NO. KNOWLEDGE RO- SRO AKl. Knowledge of the op rational implications of the (
following concepts as they apply to INCOMPLETE SCRAM : AKl. 01 Shutdown margin....................................... 3.6* 3.9* : AK1.02 Cooldown e f fect s o n reactor power. . . . . . . . . . . . . . . . . . . . . 3.9 4.1 AK1.03 Re a c t iv i ty e f fec ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8 3.9 AK1.04 Reactor pressure: P l a n t - S pe c i f i c . . . . . . . . . . . . . . . . . . . . . 3.8 3.8 AK2. Knowledge of the interrelations between INCOMPLETE SCRAM and the following: A- AK2.01 CRD hydraulics........................................ 3. 8 3.9 3.6 3.7 (U) AK 2. 02 AK2.03 RMCS: Plant-Specific................................. Rod control and information system: Pl ant-Spec i fic. . . 3.2 3.6 , AK2.04 RPS................................................... 4.0 4.1 l AK2.05 Rod worth minimizer: Plant-Specific.................. 2.6 2.8 AK 2. 06 RSCS: Plant-Specific................................. 2. 6 2.8 AK2.07 CRD mechanism.. ...................................... 3. 3 3. 4 AK2.08 Neu tro n moni tor i ng sy stem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6 3. 7 AK2.09 1PIS.................................................. 3.5 3.6 AK2.10 SPDS/ERIS/CRIOS/ GDS: Pl an t-S pec i f i c. . . . . . . . . . . . . . . . . . 2. 8 3.0 AK2.11 Instrumtat air........................................ 3. 5 3.7 AK3. Know h ge of the reasons for the following respnses as th .* apply to INCOMPLETE SCRAM : , AK3.01 Bypass ing rod insertio n bl ocks. . . . . . . . . . . . . . . . . . . . . . . . 3.4 3.7 l
)
ABILITY l AA1. Ability to operate and/or monitor the following as they apply to INCOMPLETE SCRAM : A A 1. 0.: C R0 hyd rau l i c s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8 3.9 AA1.02 RPS................................................... 4. 0 4.2* AA1.03 RMCS: Plant-Specific................................. 3. 6 3.8 AA1.04 Rod control and information system: Plant-Specific... 3.4 3.7
,/x .Y AA1.05 Rod worth minimizer: Pl an t-Spec i f i c . . . . . . . . . . . . . . . . . . 2. 5* 2.8* .k.w / AA1.06 RSCS: Plant-Specific................................. 2. 7 2.9 w
K/A catalog: BWR 4.2-45
Emergency / Abnormal Plant Evolutions APE: 295015 Incoiaplete SCRAM IMPORTANCE X/A NO. ABILITY R0 SRO AA1.07 Neutron monitoring system............................. 3. 6 3. 7 AA1.08 Process computer /SPOS/ERIS/CRIDS/ GDS: Plan t-Speci fic. 2. 7 2.9 AA2. Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM : AA2.01 Reactor power......................................... 4.l* 4.3* AA2.02 Control .ud position.................................. 4.1
- 4.2*
SY5 74 GENERIC K/As
- 1. tKnowledge of system status criteria which require the notification of plant personnel. 3.2 3.9
- 2. tKnowledge of which events related to system operation / status should be reported to outside agencies. 3.2* 4.5*
- 3. tKnowledge of limiting conditions for operations and safety limits. 3.2 4. 0
- 4. tKnowledge of bases in technical specifications for limiting conditions for operations and safety limits. 2.9 3.8
- 5. Knowledge of the annunciator alarms and indications, and use of the response instructions. 3. 8 3.8
- 6. tAbility to locate and operate compone'its, including local controls. 4.1 3.9
- 7. Ability to explain and apply all system limits and precautions. 3. 4 3.5
- 8. TAbility to recognize indications for system operating parameters which are entry-level conditions for technic 11 spec ifications. 3. 5 4.2
- 9. Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. 3. 7 3.7
- 10. Ability to perform without reference to procedures those actions that require immediate operation of system components or controls. 4.0* 3.9*
- 11. Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. 4.2* 4.4*
K/A catalog: BWR 4.2-46
t
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i l' l
'- W ~ Emergency / Abnormal Plant Evolutions
- . ' APE
- 295015-Incomplete SCRAM 4
e
; IMPORTANCE -!
SYSTEM GENERIC K/As R0 SRO -
- 12. tAbility to utilize sympt 6rc. based procedures. 3. 7 4.4* i 4 *
- +
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- i i
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i e I f i i-i i i. L K/A catalog: ,BWR . 4.2-47
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