Text

Initial Submittal of the Outline and NRC Comments for the Perry Initial Examination - March 2002
	ML021190438
Person / Time
Site:	Perry
Issue date:	03/05/2002
From:	Dante Johnson; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
References
	50-440/02301, ES-201-2
	Download: ML021190438 (53)
	v • d • e

{{#Wiki_filter:INITIAL SUBMITTAL OF THE OUTLINE AND NRC COMMENTS FOR THE PERRY INITIAL EXAMINATION - MARCH 2002

PERRY INITIAL LICENSE EXAM MARCH 5 THRU 13, 2002 Form ES-201-2, "Examination Outline Quality Checklist," along with the written examination and operating test outline(s)

ES-201 Examination Outline Form ES-201-2 (R8,S1) Quality Checklist Facilit Date of Examination: Initials Item Task Description a b* c# 1

a. Verify that the outline(s) fit(s) the appropriate model per ES-401.

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b. Assess whether the outline was systematically and randomly prepared in accordance with Section D.1 of ES-401 and whether all K/A categories are appropriately sampled.

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c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics.

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d. Assess whether the justifications for deselected or rejected K/A statements are appropriate.

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2.

a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, and major transients.

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b. Assess whether there are enough scenario sets (and spares) to test the projected number and M

mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity; ensure each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s)*, jj f and scenarios will not be repeated over successive days.

c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.

3.

a. Verify that:

(1) the outline(s) contain(s) the required number of control room and in-plant tasks, 1 W (2) no more than 30% of the test material is repeated from the last NRC examination,-"' (3)* no tasks are duplicated from the applicants' audit test(s), and V, T (4) no more than 80% of any operating test is taken directly from the licensee's exam banks. v-

b. Verify that:

(1) the tasks are distributed among the safety function groupings as specified in ES-301," (2) one task is conducted in a low-power or shutdown condition,, V (3) 40% of the tasks require the applicant to implement an alternate path procedure, V (4) one in-plant task tests the applicant's response to an emergency or abnormal condition, andV (5) the in-plant walk-through requires the applicant to enter the RCA. -'

c. Verify that the required administrative topics are covered, with emphasis on perform ance-basedw activities.

d. Determine if there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on successive days.

4.

a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the i{4X appropriate exam section.

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b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate.

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c. Ensure that K/A importance ratings (except for plant-specific priorities) are at least 2.5.

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d. Check for duplication and overlap among exam sections.

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e. Check the entire exam for balance of coverage.

f. Assess whether the exam fits the appropriate job level 20 or SRO).

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a.

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Printed eISign tr

a. Author

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b. Facility Reviewer Ae '

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c. NRC Chief Examiner (#)

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d. NRC Supervisor 4

(.L~ Note:

Not applicable for NRC-developed examinations.

Independent NRC reviewer initial items in Column "c;" chief examiner concurrence required.

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ES 401 BWR P0 Examinat on Outline Form ES 401 Facility Perry Tier 1

Fm~rni~nrx Abnormal Plant Evolutions Date of Exam-3/4/2002 Group 1

2 3 Tier Totals Exam Level RO K/A Category Points T T I T I r y K1 2 1 K2 K3 2 K4 K5 K6 Al 5 A2 2 A3 A4 G T f tt-1 I F F- + F-7 4 4 2 +/- ~ + 4-+- f 4 4-9 0 3 1 9 1 7 10 5 0 2 Point Total 13 19 4 36 2.1 4 2 4 4 2 2 3T2 1 3 1 28

2.

Plant 2 4 0 0 4 2 1 1 3 1 2 1 19 Systems 3 0 0 0 1 0 1 1 0 0 0 1 4 Tier 8 2 4 9 4 4 5 5 2 5 3 51 Totals

3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 4

3 2 4 13 Note:

1.

'Ensure that at least two topics from every K/A category are sampled within each tier (i.e., the "Tier Totals" in each K/A category shall not be less than two).

2.

The Jpoint total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/- 1 from that specified in the table based on NRC revisions The final exam must total 100 points.

3.

Select topics from many systems: avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.

4.

Systems/evolutions within each group are identified on the associated outline.

5.

The shaded areas are not applicable to the category/tier. 6.* The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.

7.

On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the RO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above. Page 1 of 15 NUREG-1021, Revision 8, Supplement 1 Perry Final - Revision 1 ES-401 bS-4U1 BWVR RO Examination Outline Form ES-401-2 £ 1

Form ES-401-2 BWR RO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 295 295c 295* 295( 295 295 29,5, 295¢ 295C:

Abo.

50cA K/A KiILK2 1K3 Al A2 G G X X X X -4 .4 -4 .4 4 4-X 'AP,_P:: / Name / Safety Function

Mar, urbine Generator Trip /3 SCRAM / 1 High Reactor Pressure / 3 Low Reactor Water Level / 2 Low e'actor Water Level / 2 i4(J. rywellPress~ure / 5 ina tent Reactivity Addition /1 Inco mlete SCRAM / 1 Higih Drywell Pressure / 5 High Reactor Pressure/ 3 Reactor Low Water Level / 2 SCRAM Condition Present and Power D,Pr.! Downscale or Unknown/1

-Icl ontainment Hydrogen Conc. / 5 X X -4 4 4 4-4- X X I I j i I I X 4 1 1 1 F +/- X I I___ I I___ 4 ~ X 4 4 4 +/- .I X -2 1 I 2 5 2 1 Totas: _l... K/A Topic(s111 0ln. o Point AA1.07 Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: AC Electrical Distribution AA2.06 Ability to determine and/or interpret the following as they apply 3.5 1 to SCRAM: Cause of Reactor SCRAM AA2.01 Ability to determine and/or interpret the following as the apply to 4 1 HIGH REACTOR PRESSURE: Reactor Pressure AA1.04 Ability to operate and/or monitor the following as they apply to 2.7 1 LOW REACTOR WATER LEVEL: Reactor Water Cleanup I AK1.01 Knowledge of the operational implications of the following concepts as they apply to LOW REACTOR WATER LEVEL: Steam 2 7 1 Carryunder 2.4.1 Knowledge of EOP entry conditions and immediate action steps AK1.03 Knowledge of the operational implications of the following concepts as they apply to INADVERTENT REACTIVITY ADDITION: Shutdown Margin AA1.04 Ability to operate and/or monitor the following as they apply to INCOMPLETE SCRAM: Rod Control and Information System EK2.05 Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: RPS 4.3 i I 3.7 17 1 3.4 39 EK3.09 Knowledge of the reasons for the following responses as they ] 3.7 apply to HIGH REACTOR PRESSURE: Low-Low set initiation EK3.03 Knowledge of the reasons for the following responses as they i 4.1 1 apply to REACTOR LOW WATER LEVEL: Spray Cooling EA 1.04 Ability to operate and /or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE 4.5 APRM DOWNSCALE OR UNKNOWN: SBLC EA1.03 Ability to operate and monitor the following as they apply to HIGH CONTAINMENT HYDROGEN CONTROL Containment Atmosphere Control system 34 3roup Point Total: NUREG-1021, Revision 8, Supplement 1 i r r Page 2 of 15 Pert'.

ial Revision 1 J

29,,, 295 Cot 295 29, 29' 29 295C 29,5ný

295{

295. 2952 29-.2 295r. ,!295022-, Ki [ K2 BWR RO Examination Outline BWR RO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 K3 Al A2 0 K/A Toprc(s) X -4 4 4 4-4- X + 4 X -_4 4 SAPi /Name/ Safety. Function c~rv. or Complete Loss of Forced ,wC -jlation / 1 & 4 Los Main Condenser Vacuum / 3 Par:. or Complete Loss of AC Pwr / 6

9a': :i or Complete Loss of DC Pwr / 6 H.c.

eactor Water Level/ 2 Hig T-MT Temperature /5. High Drywell Temperature/5 High, uppression Pool Temp./ 5 Con: -,! Room Abandonment / 7 --4g:' ff-site Release Rate / 9 Pa*-*.,; or Complete Loss of CCW / 8 Par Or Comp. Loss of Inst. Air/ 8 X 4 4- +/- X 4 4-4- 4- .4-x -1. I-i--i ---- 4 -s i X X

4.

4 X 4 4 4-x 4 - 4 4 .4- .4 4 4 4-- Inad ertent Cont. Isolation / 5 & 7 Loss of CRD Pumps / 1 29 5. Surpession Pool High Water Temp. /5 X -4 .4 4 i-x I-I- -l

i

i t X 2 29 AK3.06 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Core Flow indication AK3.04 Knowledge of the reasons for the following responses as they apply to LOSS OF MAIN CONDENSER VACUUM: Bypass valve 3 4 closure AA1.04 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OFAC POWER. DC Electrical 3.6 distribution systems AA1.01 Ability to operate and/or monitor the following responses as rile~y apply to PARTIAL OR COMPLETE LOSS OF DC POWER DC Electrical distribution systems AK3.08 Knowledge oft he-reasons for the folowing resp onses as the, apply to HIGH REACTOR WATER LEVEL RCIC Steam supply vav'e closure 3.3 3.4 AA2.03 Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT TEMPERATURE (Mark III Containment Only): 2.8 Containment Humidity AK1.01 Knowledge of the operational implications of the following 3 concepts as they apply to HIGH DRYWELL TEMPERATURE: I3.3 Pressure/temperature relationship AK2.01 Knowledge of the interrelations between HIGH SUPPRESSION 3.6 POOL TEMPERATURE and the following: Suppression Pool Cooling 2.4.34 Knowledge of RO tasks performed outside the main control ro om during emergency operations including system geography and system 38 implications AK2.14 Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: PCIS/NSSSS AK2.01 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the following: System Loads AK3.03 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Service Air isolations 4.0 33 32 1 AK2.12 Knowledge of the interrelations between INADVERTENT I CONTAINMENT ISOLATION and the following: Instrument Air/Nitrogen 3.1 1 AA1.02 Ability to operate and/or monitor the following as they apply to I 3.6 1 LOSS OF CRD PUMPS: RPS EA1.03 Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Temperature 3.9 1 Monitoring NUREG-1021, Revision 8, Supplement 1 X Page 3 of 15 Perrv,.tiaý -- Revision I I f FormES.C"

ES

_Name / Safety Function

TMT Temperature / 5 High -rywell Temperature / 5 High Suppression Pool Water Level / 5 Low Suppression Pool Water Level I 5 High Sec. Cont. Area Rad. Levels / 9: Sec Cont. Ventilation High Rad. / 9 H'cgh Off-site Release Rate / 9 Planý Fire On Site / 8 egoi Point Totals:

BWR RO Examination Outline Form ES-40"-2 Emermeneqcy and Abnormal Plant Evolutions - Tier 1/Group 2 K1 K2 K3 Al A2 G K/ATokic(s) ImiD P or's EK2.01 Knowledge of the interrelations between HIGH CONTAINMENT X I TEMPERATURE (Mark III Containment Only) and the following:

3. 2 Containment Spray x

-------4 .4 4 4 x r

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.i I~- -I I__ I4 1-1 1 t t 1 / 7 x 1 _1 1 4 4 2 EK2.07 Knowledge of the interrelations between HIGH SUPPRESSION POOL WATER LEVEL and the following: Drywell/Containment Water 3.1 1 Level EK2.08 Knowledge of the interrelations between LOW SUPPRESSION 3.5 1 POOL WATER LEVEL and the followin

gj___SRV Dischaýrge_submergence 2.1T Ability to determine and interpret the foWtoing-as-they-ta-pply to 31 PLANT FIRE ON SITE:

Systems that may be affected by the fire1 Group Point Total 19 NUREG-1021, Revision 8. Supplement 1 295(1 "2951

2959 5

295" '1 600C, K/A 1

Pert,

.a; Revision 1 Page 4 of 15

S APL Name/ Safety Function i 295'. Loss of Shutdown Cooling / 4 2951.- Refueling Accidents / 8 29'*" H tgi econdary Containment Area Tell'.: aturl 5 D ii.' 29, K/A Secc,, *ry Containment High ý:lai Po ssure / 5 Secondary Containmen e1z'..I etr Level / 5

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nt Totals: nt High I BWR RO Examination Outline Emeraencv and Abnormal Plant Evolutions - Tier 1/Group 3 K1 K-K3 X Al X A) G + - I t t r X X 1 1 1 0 K/A Tooic(s) AK3.01 Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING: Raising reactor water level Form ES-40 1-2 33 _P o1,:.s 3,3 1 AA1.03 Ability to operate and/or monitor the following as they apply to 3 REFUELING ACCIDENTS: Fuel Handling equipment .3 EK2.08 Knowledge of the interrelations between HIGH SECONDARY CONTAINMENT AREA TEMPERATURE and the following Systems 3.8 8 required for safe shutdown Suppressed EA2.03 Ability to determine and/or interpret 1 e following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL-3,4 Cause of the high water level Group Point Total: NUREG-1021, Revision 8, Supplement 1 Page 5 of 15 Pert. "-la! Revision I K3 Al A2 G I I I I I I I P

Form ES-,a i-2 tem # / Name 2C r" vdraulc System RrV1 M, 20'. Pc ,' iltion Flow Control RH}I LPCI Injection Mode HPCý Iso.ation (Emergency) "set HPS

SLPC, SLC x

x K/A Topa~ic(s) A3.05 Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including: Reactor water level Suppressed K6.04 Knowledge of the efect hat a loss malfunction of t;e following v,:ii have o. tn: ROD CONTROL AND INFORMATION SYSTEM IRM channel K3.06 Knowledge of the effe-t that a loss or malfunction of !he RECIRCULATION FLOV7 CONTROL SYSTEM will have on the following Recirc laVto, f.ow control vav, position A4.06 Ability to manually operate and/or monitor in the control room System reset following automatic initiation Suppressed Suppressed K5 01 Knowledge f the operat onal implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM Indicat o s of pump Imo 2, cavitation K3 01 Kinowledge of the effect that a loss or malfunction of the HIGH PRESSURE CORE 3.9 SPRAY SYSTEM (HPCS) will have on the following-Reactor Water level K2.03 Knowledge of electrical power 2.8 supplies to the following: Initiation logic K2.02 Knowledge of electrical power 3.1 1 supplies to the following: Explosive valves NUREG-1021, Revision 8, Supplement 1 20 20 Poi ts 205: 207

20S Co 20

209, 21 1 I

Per:- - -ai Revision 1 Page 6 of 15

c, ,stem # / Name SLU K1 K2 K 3 K4 BWR RO Examination Outline Plant Systems - Tier 2/Group 1 K5 K6 Al A2 A3 A4 I G x _______ -I _______ _______ I x 4 -t -I 2'. 21, ,Rr% IRM FR. SR,. 215 APF'1, LPRM x Ir-rr 1 1 1, x I x x X Form. ES-- K!A T i ImpT

p.

2.1.33 Ability to recognize indications for system operating parameters which are I entry-level conditions for technical specifications Al. 11 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR PROTECTION 3.4 SYSTEM controls including System status lights and alarms K5 02 Knowledge 3f the operational implications of tie followw ng concepts as 3.3 they apply to REACTOR PROTECTION SYSTEM. Specific logic arrangements P* 06 Ability to (a) predict the impacts of the wing on the INTERMEDIATE RANGE NITOR (IRM) SYSTEM and (b) based on 1 e predictions, use procedures to correct, l 3 0 rol, or mitigate the consequences of e abnormal conditions or operations: l ty Range switch 11 Knowledge of INTERMEDIATE OGE MONITOR (IRM) SYSTEM design 3.7 ure(s) and/or interlocks which provide for ollowing: Rod withdrawal blocks 14 Knowledge of the effect that a loss or unction of the SOURCE RANGE 3.7 NITOR (SRM) SYSTEM will have on the wing. Reactor power and indication 2 Knowledge of the effect that a loss or unction of the following will have on the 3 1 JRCE RANGE MONITOR (SRM) TEM: 24/48 volt DC power 6 Knowledge of AVERAGE POWER GE MONITOR/LOCAL POWER RANGE NITOR SYSTEM design feature(s) 2.6 or interlocks which provide for the wing Effects of detector aging on I M/APRM readings NUREG-1021, Revision 8, Supplement 1 EE 21i 212 21

Perr,

hal -- Revision 1 3

i X Page 7 of 15

.sem # / Name K1 K2 K" N'. . - Boiler Instr,.mientation X RCIC 21 7

.ýCl ADS K4 K5 BWR RO Examination Outline BWR RO Examination Outline Plant Systems_- Tier 2/GrouLp 1 K6 I Al A2 A3 A4 0

G t I r -i 1 1 1 1 X - 4- A + X X --i 1 r r P ri-.r;v CTMT and Auxiliaries "CIS Nuclear Steam Supply X ýýeac-r!Turbine Pressure X X X I4 i [ I 1' t 1 J. _________ K/A To ic s) K1 09 Knowledge of the physical connections and/or cause-effect relationships between NUCLEAR BOILER INSTRUMENTATION and the following: Redundant reactivity control/ alternate rod insertion Form ES-401-2 I m p_ Po 37 1 Ability to (a) predict the impacts of the ng on the REACTOR CORE TION COOLING SYSTEM (RCIC) and ed on those predictions, use 3.0 ures to correct, control, or mitigate the uences of those abnormal conditions ations: Valve openings___ Ability to manually operate and/or r in the control room : Manually 36 d controls Knowledge of AUTOMATIC ESSURIZATION SYSTEM design 3.8 1 (s) and/or interlocks which provide for owinQ AADS__o ic control Ability to manually operate and/or r in the control room Containment 4.0 re (Mark ill) Knowledge of the physical connections and/or cLuse-e ffect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the foilowlig Hligh pressure core sP ray -------............ K4.09 Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or interlocks which provide for the following: Manual opening_ of the SRV K3 30 Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on the followilng3 _OC 3.7 30 NUREG-1021. Revision 8, Supplement 1 ES.1 21ý 217: 223C Shu: 239 24. Rec: i _7

R\\'

Page 8 of 15 Perý.

al-Revision 1

ES ';tem # /Name 25 Feedwater 25,.ý Ke- ,r Water Level Control 2o1 260 K/A SGT X IK K2 K3 K4 BWR RO Examination Outline Plant Systems - Tier 2/Group 1 K5 K6 Al A2 I A3 A4 1 G X x -F-----i-- -+/- 1 1 i-i------r r EDGs X 0go, Point Totals 4 2 4 4 2 2 3 2 3 ! 1 Form ES-401-2 K/A TOpcs) Imp Por'_ s A1.02 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER 3 2 SYSTEM controls including Feedwater inlet temperature A102 Ability to predict and/or monitor changes in parametcrs associated with operating the REACTOR WATER LEVEL 3 6 CONTROL SYSTEM contrc's ncldmng Reactor feedwator flow K1+K07 Knowlocige of tile physical connections and/or cause-edect relationships between STANDBY GAS 3.1 TREATMENT SYSTEM and the following: Elevated stack release KI 06 Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY 3.2 GENERATORS (DIESEL/JET) and the following: Starting system Group Point Total 20ý NUREG-1021, Revision 8, Supplement i Per'. ai -Revision 1 Page 9 of 15 I I I I I I 4

BWR RO Examination Outline Plant Systems - Tier 2/GrouR2 KI K2 K 3 K4 K5 K6 A 1 A2 A3 A4 G. K/AuTooic s T-K504 Knowledqe Iui eono rationa implications of t fowtI con.cepts as they app!y to CONTRO)L RUD AND DR V:. _C MECHANISM Rod sejuence patteis Suppressed -T -{ t t X t T t t x I-I I. i-205 h n Cooling System (Rr' ,J t Cooling mode) 2'I~ 21 i P -;i I 21 BE 21 RiR PCI Torus/Pool Cooling X Moc 225', RHR LPCI Containment Spray Sys' x I I ýr ",, 7 -, Suppressed I K4.01 Knowledge of RECIRCULATION System design features) and,'or interlocKs which provide for th*e following 2/3 core coverage K1.08 Knowledge of the physical connections and/or cause-effect relationships between RECIRCULATION SYSTEM and the followin _AC Electrical I-i- - I- .ffi L X X ~ 3.1 A3.03 Ability to monitor automatic operations of the REACTOR WATER CLEANUP SYSTEM including Response to system solations K4.05 Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) desi gn feature(s) and/or 3.5 interlocks which provide for the following: Reactor cooldcwn rate Suppressed Suppressed KI 04 Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI. TORUS/SUPPRESSION POOL COOLING MODE and the following: LPCI/RHR pumps A4.08 Ability to manually operate and/or monitor in the control room System flow 2 NUREG-1021, Revision 8, Supplement 1 Pert-. -al-Revision 1 E,* 20 Rod and Drve 20 SCr 20' .O rX 202 -lation 202 Recir ulation I I Page 10 of 15 I ýern

/ Namre SJi I

I

ES M' 23- '.1a 24r A> -s

.stem

/ Namne K1 K2 K,

'.PCI: TorusiPool Spray K4

-,d Reheat S:eam

' Turbine Gen And 25nr. Rea.ýor Condensate 262 " AC - ectrical Distribution 262., UPS ,C!DC) DC F ectrical Distribution Offgas BWR RO Examination Outline Plant Systems - Tier 2/Group 2 K5 K6 Al I A2 A3 A4 G IX X 1 1 t X X X Form-ES-V. K/A Topic(s) Suppressed A4.09 Ability to mian:aIly o oer an tid/or monitor in the contri room Reactor pressure A2,07 Ability to (a) predict tie impacts ol the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and (b) based on "nose prei ct:ons, use procedures to correct, contiro or mitigate the consequences of those abnormal conditions or operations Loss of reactor,,turbiue pressure control systems A1.04 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CONDENSATE 2 2 SYSTEM controls including. Hotwell level 2.4.50 Ability to verify system alarm setpoints and operate controls identified ir the alarm response manual K4.01 Knowledge of UNINTERRUPTIBLE POWER SUPPLY (AC/DC) design feature(s) and/or interlocks wiich provide for the 3 following. Transfer from preferred power to alternate power supplies K4.02 Kiiowledge of DC ELECTRICAL ,I DISTRIBUTION design feature(s) and/or interlocks which provide for the following: 3.1 1 Breaker interlocks, permissives, bypasses and cross ties -A2.14 Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM and (b) based on those predictions, use procedures X to correct, control, or mitigate the 2.6 consequences of those abnormal conditions or operations: Offgas filter high differential pressure Rad ation Monitoring X K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the RADIATION MONITORING SYSTEM. Reactor Protection Systemr_ 30 NUREG-1021, Revision 8, Supplement 1 263'

271C, 272" Page 11 of 15 Pert,.

i al - Revision 1 1

S. S.. S286: 29C Secc-d BWR RO Examination Outline Plant Systems - Tier 2/Group 2 T ~K A2 A3!K4'K em# / Name KI K3K4 K6 AlA2 3 tection ary CTMT 290,-. Conrol Room HVAC x 3001' . Instrument Air x Comconent Cooling Water KiA " egor Point Totals: Form ES-40-2 A4 G +-1 I I t t r-I I I I r r r r r x 4 4 0 04 2 1 113 K/A Tjic2ls)-

Imp, Poinrs K5.02 Knowledge of the operational implications of the following concepts as 2.6 they apply to FIRE PROTECTION SYSTEM.

Effect of halon on fires Suppressed K1.03 Knowledge of the physical connections and/or cause-effect relationships between CONTROL ROOM HVAC and the following: Remote Air intakes K1.04 Knowledge of the connections and/or cause/effect relationships between INSTRUMENT AIR SYSTEM and the followingc Cooling _water to compressor A2.03 Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those abnormal operations: High/Low CCW temperature 1 2.8 2.8 i 1 2.9 1 Group Point Total: NUREG-1021, Revision 8, Supplenient i 400-ý Fire P-o Page 12 of 15 Pert-. -ýal - Revision 1

BWR RO Examination Outline Plant Systems - Tier 2/Group 3 S. stenn # /Name 21' Trav-simg In-Core Probe 23" nol Cooling and Cleanup 234- .e -andling Equipment K2 K3 K4 I K5 ".,S!'. Leakage ConrIol ac. aste Pat-: Ventilation Reac:or Vessel Internals -eao:. Point Totals: x K6 Al _ A2 __A3 __A4 Foro ES-:0C G LK/A Topic(s) Imp Suppressed -i 4 r + I x K4.06 Knowledge of FUEL POOL COOLING AND CLEANUP design featuie s) and/or 20 interlocks which provide for tie following: Maintenance of adequate pool level 01 Ability to predict and/or monitor changes in parameters associated with operating the FUEL HANDLING 3.1 i EQUIPMENT controlsincIdAI n Spent Fuel Pool level Suppressed I

t It I

i 1 4 t t 0o 0o 0 0 x 0 0 0 1i x 2.1.27 Knowledge of system purpose and/or 2.3 I function K6. 15 Knowledge of the effect that a loss or malfunction of the following w I have on the 3.1 REACTOR VESSEL INTERNALS: ADS Group_ Point Total Plant-Specific Priorities System / Topic oimary Containment and Auxiliaries / Containment pressure ,0 CIC / Manually initiated controls Recommended Replacement for Reason 223001K6 12 Plant Specific high importance PSA for manually venting containment on high pressure 217000K5.03 Plant Specific high mpor'ance PSA requiring r-nanucal I initiation of RCIC Pa 1,, c oority Total (limit 10) 2 NUREG-1021, Revision 8, Supplement 1 PC -! 23c. 26

288, 290K

, K 22 21 [ i 1 1 1

Pert,

-al -- Revision I Page 13 of 15

Generic Knowledge and Abilities Outline (Tier 3) C -egory Operations

1-Control K/A#

Topic Imp. 2 1 11 Knowledge of less than one hour technical specification action statements for systems I 3.0 21.14 Knowledge of system status criteria which require the notification of plant personnel 2.5 2.1 29 i Knowledge of how to conduct and verify valve lineups 34 2.1.30 Ability to iocate and operate components / including local controls 3.9 Total 2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown 4.0 and designated power levels 22.13 Knowledge of tagging and clearance procedures 3.6 2.2.34 Knowledge of the process for determining the internal and external effects on core reactivity Points 2.8 1 Total 2 2 ir! : F, Control 3 1 Knowledge of 10CFR20 and related facility radiation control requirements .3.9 Knowledge of the process for performing a containment purge L 2 Total Ability to recognize abnormal indications for system operating parameters which are entry-level 4,0 conditions for emergency and abnormal operating procedures 2.4.11 Knowledge of abnormal condition procedures 34 1 Er. ,icv P-ocedures/Plan 2.4.12 Knowledge of general operating crew responsibilities during emergency operations 3.4 1 2.4.19 Knowledge of EOP layout/symbols/ and icons 2.7 Total 4 To!il (RO/SRO) lal - Revision 1 Page 14 of 15 2.6 2.5 Tier NUREG-1021, Revision 8, Supplement 1 Perry FORM ES-401-5 ES-Z 1 1 I 1

Record of Rejected K/As ic-t,*dmviSelct~ed K/A ES-4 2 2 2 12 2/21 2T1 1/2 2/ 1 21000K1.09 2 2 -,3000K4 01 Form ES-401-10 Reason for Rejection No wetwell sprays at facility RPIS not relevant to selected APE for facility Facility does not have ADS acoustic sensors Facility does not have FWCI (BWR-6) Facility does not h ve RHR Loop selection logic (BWR-6) Single unit faciity does not have unit cross-ties for instrument air Tier I Group 1 has more questions than available systems for RO level. EK2 for same EPE was previously selected. This topic was rejected to minimize chance for redundant questions. Additionally, this topic has a lower K/A value System interaction with suppressed system (MSIV Leakage control) Replaced by plant soecific priority Replaced by plant specific priority After further scrutiny, unable to write a viable exam question. It was determined this APE topic not applicable to Perry. The new topic (AA2.17) also has a higher Importance Rating. After further scrutiny unable to write a viable exam question. It was determined this APE topic not applicable to Perry. There is no relationship between AEGTS and Primary Containment Isolation System at Perry. After further scrutiny, unable to write a viable exam question, It was determined this APE topic not appr cable to Perry The new topic (K4.02) also has the same Importance Rating. 1.. Perry Final - Revision 1 NUREG 1021, Revision 8, Supplement 1 (*:M0000EA1 05

c,516AA1 03

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ýo02A2 08 S000K4 i I 9000K3Q 03
95024EK2 01 S239001K1.13 223001 K6.12 7

.7000K5.03 600000AA2 06 Page 15 of 15 i :-ýý--ý,domlv Selected K/A I i I

ES-401 BWR SRO Examination Outline Form ES-401-1 Facility: Perry Date of Exam: 3/4/2002 Exam Level: SRO K/A Category Points Tier Group Point K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G Total 1 3 7 2 8 4 2

1.

2 2 26 Emergency& 2 1 4 4 4 3 17 Abnormal Plant Evolutions Tier 4 11 6 12 7 3 43 Totals 1 2 2 3 2 2 2 1 2 0 4 3 23

2.

Plant 2 1 0 0 2 0 1 2 4 2 0 1 13 Systems 3 0 0 0 1 1 0 1 0 0 1 0 4 Tier 3 2 3 5 3 3 4 6 2 5 4 40 Totals

3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 4

5 2 6 17 Note:

1.

Ensure that at least two topics from every K/A category are sampled within each tier (i.e., the "Tier Totals" in each K/A category shall not be less than two).

2.

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/- 1 from that specified in the table based on NRC revisions. The final exam must total 100 points.

3.

Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.

4.

Systems/evolutions within each group are identified on the associated outline.

5.

The shaded areas are not applicable to the category/tier. 6.* The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.

7.

On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the RO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above. NUREG-1021, Revision 8, Supplement 1 Perry Final - Revision 1 Page 1 of 15

1 ES-401 E/APE # / Name / Safety Function 295003 Partial or Complete Loss of AC Pwr/ 6 295003 Partial or Complete Loss of AC Pwr/ 6 295006 SCRAM / 1 295006 SCRAM / 1 295007 High Reactor Pressure / 3 295009 Low Reactor Water Level / 2 K1 x BWR SRO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 K2 x K3 Al x x A2 x x G K/A Topic(s) Form ES-401-1 Imp. Points AK2.03 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF AC POWER and the following: A.C. electrical distribution system AA1.04 Ability to operate and/or monitor the following as they apply to PERTIAL OR COMPLETE LOSS OF AC POWER: DC Electrical distribution systems AK1.03 Knowledge of the operational implications of the following concepts as they apply to SCRAM: Reactivity Control AA2.06 Ability to determine and/or interpret the following as they apply to SCRAM: Cause of Reactor SCRAM AA2.01 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor Pressure AA1.04 Ability to operate and/or monitor the following as they apply to LOW REACTOR WATER LEVEL: Reactor Water Cleanup 3.9 3.7 1 4.0 3.8 4.1 2.7 AK1.01 Knowledge of the operational implications of the following 2 295009 Low Reactor Water Level / 2 [ concepts as they apply to LOW REACTOR WATER LEVEL: Steam 2.9 Carryunder AA2.01 Ability to determine and/or interpret the following as they 3.8 1 295010 High Drywell Pressure / 5 -apply to HIGH DRYWELL PRESSURE: Leak rates 295010 High Drywell Pressure/ 5 295013 High Suppression Pool Temp. / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 295016 Control Room Abandonment / 7 29501 7 High Off-site Release Rate / 9 295023 Refueling Accidents Cooling Mode / 8 295024 High Drywell Pressure / 5 x x X I- i 4-x x x x x x 2.4.1 Knowledge of EOP entry conditions and immediate action steps AA1.01 Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Suppression Pool Cooling AK2.01 Knowledge of the interrelations between HIGH SUPPRESSION POOL TEMPERATURE and the following: Suppression Pool Cooling AK1.03 Knowledge of the operational implications of the following concepts as they apply to INADVERTENT REACTIVITY ADDITION: Shutdown Margin AA1.04 Ability to operate and/or monitor the following as they apply INCOMPLETE SCRAM: Rod Control and Information System 2.4.34 Knowledge of RO tasks performed outside the main control room during emergency operations including system geography and system implications AK2.14 Knowledge of the interrelations between HIGH OFF-SI RELEASE RATE and the following: PCIS/NSSSS AA1.03 Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: Fuel Handling equipment EK2.05 Knowledge of the interrelations between HIGH DF PRESSURE and the following: RPS 4.6 1 NUREG-1021, Revision 8, Supplement 1 Page 2 of 15 Perry Final - Revision 1 1 1 1 1 1

ES-401 E/APE # / Name / Safety Function 295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp. /5 295027 High CTMT Temperature / 5 295030 Low Suppression Pool Water Level/5 29503 1 Reactor Low Water Level / 2 2950 '7 SCRAM Condition Present and Power Above APRM Downscale or Unknown / 1 295038 High Off-site Release Rate / 9 500000 High Containment Hydrogen Conc. / 5 KI BWR SRO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 K2 K3 Al A2 G 4 4 4 4

i.

I-X X I I I r F 4- +/- X 1 4 r F 4- x I-I- F + + 4 4-x t t I I I-X X X X K/A Topic(s) EA2.01 Ability to determine and/or interpret the following as apply to HIGH DRYWELL PRESSURE: Drvwell Pressure EK3.09 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE. Low-Low set initiation EA1.03 Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Temperature Monitorinq EK2.01 Knowledge of the interrelations between HIGH CONTAINMENT TEMPERATURE (Mark Ill Containment Only) and the following: Containment Spray EK2.06 Knowledge of the interrelations between LOW SUPPRESSION POOL WATER LEVEL and the following: Suppression Pool makeup EK3.03 Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL: Spray Coolinq EA1.04 Ability to operate and /or monitor the following as they apply SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN, SBLC EA.03 Ability to operate and monitor the following as they apply to HIGH CONTAINMENT HYDROGEN CONTROL: Containment Atmosphere Control System Form ES-401-1 Imp. Points K/A Category Totals: 1 2 8 4 2 Group Point Total: NUREG-1021, Revision 8, Supplement 1 EK2.10 Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Condenser Air Removal System Perry Final - Revision 1 Page 3 of 15

ES-401 E/APE # Name / Safety Function 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 295002 Loss of Main Condenser Vacuum / 3 295004 Partial or Complete Loss of DC Pwr / 6 295005 Main Turbine Generator Trip / 3 295008 High Reactor Water Level / 2 295011 High CTMT Temperature / 5 295012 High Drywell Temperature / 5 295018 Partial or Complete Loss of CCW/ 8 1 295019 Part. Or Comp. Loss of Inst. Air / 8 295020 Inadvertent Cont. Isolation / 5 & 7 295021 Loss of Shutdown Cooling / 4 295022 Loss of CRD Pumps / 1 295028 HighLDrywell Temperature / 5 295029 High Suppression Pool Water Level / 5 2950,2 High Secondary Containment Area Temperature /5 K1 X BWR SRO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 K2 X X K3 X X X X Al X X A2 x G X K/A Topic(s) AK3.06 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Core Flow indication AK3.04 Knowledge of the reasons for the following responses as they apply to LOSS OF MAIN CONDENSER VACUUM: Bypass valve closure AA1.01 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF DC POWER: DC Electrical distribution systems AA1.07 Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: AC Electrical Distribution AK3.08 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR WATER LEVEL: RCIC Steam supply valve closure AA2.03 Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT TEMPERATURE (Mark III Containment Only) Containment Humidity AK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Pressure/temperature relationship AK2.01 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the followinq: System Loads AK3.03 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Service Air isolations AK2.12 Knowledge of the interrelations between INADVERTENT CONTAINMENT ISOLATION and the following: Instrument Air/Nitrogen 2.4.9 Knowledge of low power/shutdown implications in accident (e.a., LOCA or Loss of RHR) mitioation strateciies Form ES-401-1 Imp. Points 3.0 1 3.6 3.4 3.3 3.5 1 3.2 3.5 3.4 3.2 1 3.9 1 -I-F + I + + . X X x AA1.02 Ability to operate and/or monitor the following as they apply to LOSS OF CRD PUMPS: RPS EK2.07 Knowledge of the interrelations between HIGH SUPPRESSION POOL WATER LEVEL and the following: Drywell/Containment Water Level EK2.08 Knowledge of the interrelations between HIGH SECONDARY CONTAINMENT AREA TEMPERATURE and the following: Systems required for safe shutdown 3.6 1 NUREG-1021, Revision 8, Supplement 1 Page 4 of 15 Perry Final - Revision 1 I I -Jý 1 1 1 1 1 1

ES-401 E/APE # / Name!/ Safety Function 295033 High Sec. Cont. Area Rad. Levels / 9 295034 Sec. Cont. Ventilation High Rad. / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 600000 Plant Fire On Site / 8 K/A Category Point Totals: K1 I K2 BWR SRO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 K3 Al A2 E/P

i 4

Nam / i-e ucto x x x G -4 4 I-I- I-I- -I-1 4 4 4 3 Form ES-401-1 I Imp Points K/A Topic(s) EA1.02 Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Process Radiation Monitoring System Suppressed EA2.03 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Cause of the high water level AA2.15 Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Requirements for establishing a fire watch 4.0 3.8 3.5 1 Group Point Total: NUREG-1021, Revision 8, Supplement 1 17 1 Page 5 of 15 Perry Final - Revision I 1

ES-401 Sy 201005 RCIS BWR SRO Examination Outline Plant Systems - Tier 2/Group 1 9 9 r vstem # / Name 202002 Recirculation Flow Control 203000 RHR/LPCI: Injection Mode 206000 HPCI 207000 Isolation (Emergency) Condenser K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G K1 K2 K3--4 I x x 4 4 4 4 4 4 4 -I-x 4 .4-K/A Topic(s) K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the ROD CONTROL AND INFORMATION SYSTEM: IRM channel K3 06 Knowledge of the effect that a loss or malfunction of the RECIRCULATION FLOW CONTROL SYSTEM will have on the following: Recirculation flow control valve position A4.06 Ability to manually operate and/or monitor in the control room: System reset following automatic initiation Suppressed Suppressed K5.01 Knowledge of the operational 20900", LPCS implications of the following concepts as they apply to LOW PRESSURE CORE 2.7 SPRAY SYSTEM: Indications of pump cavitation 209002 HPCS X K2.03 Knowledge of electrical power 2.9 1 supplies to the following. Initiation logic 211000 SLC 2.1.33: Ability to recognize indications 2 for system operating parameters which 4.0 1 are entry-level conditions for technical specifications K5.02 Knowledge of the operational 212000 RPS implications of the following concepts as 34 they apply to REACTOR PROTECTION SYSTEM: Specific logic arrangements K3.04 Knowledge of the effect that a loss or 215004 SRM malfunction of the SOURCE RANGE 37 MONITOR (SRM) SYSTEM will have on the following: Reactor power and indication K6.02 Knowledge of the effect that a loss or 215004 SRM malfunction of the following will have on the 33 SOURCE RANGE MONITOR (SRM) SYSTEM: 24/48 volt DC power NUREG-1021, Revision 8, Supplement 1 Form ES-401-1 Points -I imp. 3.2 3.7 3.9 1 Perry Final - Revision 1 r 1_v. Page 6 of 15

ES-401 System # / Name K1I K2 LK3 215005 APRM/LPRM 216000 Nuclear Boiler Instrumentation 217000 RCIC X K4 X BWR SRO Examination Outline Plant Systems - Tier 2/Grouo 1 K5 K6 Al A2 X A3 A4 Form ES-401-1 G K/A Topic(s) K4.06 Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following: Effects of detector aging on LPRM/APRM readings K1.09 Knowledge of the physical connections and/or cause-effect relationships between NUCLEAR BOILER INSTRUMENTATION and the following: Redundant reactivity control/ alternate rod insertion A2.12 Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve openinqs Imp.j Points 2.8 1 4.0 3.0 217000 RCIC A4.04 Ability to manually operate and/or X monitor in the control room: Manually 3.6 1 initiated controls 2.4.4 Ability to recognize abnormal 218000 ADS indications for system operating X parameters which are entry-level 4.3 1 conditions for emergency and abnormal __operating procedures A4.06 Ability to manually operate and/or 223001 Primary CTMT and Auxiliaries X monitor in the control room: Containment 4.0 1 ____pressure (Mark Ill) i K1.15 Knowledge of the physical 223002 PCIS/Nuclear Steam Supply I connections and/or cause-effect Shutoff relationships between PRIMARY X CONTAINMENT ISOLATION 3.4 1 SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following: High pressure core spray 226001 RHR/LPCI: Containment Spray Syste irode X A4.08 Ability to manually operate and/or monitor in the control room: System flow 3.1 NUREG-1021, Revision 8, Supplement 1 Page 7 of 15 Perry Final - Revision 1

ES-401 BWR SRO Examination Outline Plant Systems - Tier 2/Grouo 1 Form ES-401-1 System # Name K1 K2 K3 K4 -K K6 Al A2 A3 A4 G K/A Topic(s)

Imp, Points

_K4,09 Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or interlocks 36 239002 SRVsI i Xwhich provide for the following: Manual ___opening of the SRV K3.30 Knowledge of the effect that a loss or 241000 Reactor/Turbine Pressure malfunction of the REACTOR/TURBINE 30 1 Regulator i PRESSURE REGULATING SYSTEM will have on the following: EGC A1.02 Ability to predict and/or monitor 259002 Reactor Water Level Control changes in parameters associated with X operating the REACTOR WATER LEVEL 3,5 1 CONTROL SYSTEM controls including: Reactor feedwater flow K2.03 Knowledge of electrical power 2.5 1 261000 SGTS supplies to the following: Initiation logic 2.4.50 Ability to verity system alarm 262001 AC Electrical Distribution X setpoints and operate controls identified in 3.3 1 the alarm response manual A2.03 Ability to (a) predict the impacts of 264000 EDGs the following on the EMERGENCY GENERATORS (Diesel/Jet) and (b) based on those predictions, use procedures to X correct, control, or mitigate the 3.4 1 consequences of those abnormal conditions or operations: Operating unloaded, lightly loaded, and highly loaded L290001 Secondary CTMT Suppressed____ 2 2 3 22 1 '2 0 4 23 K/A Category Point Totals: I Group Point Total: NUREG-1021, Revision 8, Supplement 1 Page 8 of 15 Perry Final - Revision 1

ES-401 System # / Name 201001 RMC 201004 RSC: 201006 RWIM Fydrauiic System S BWR SRO Examination Outline Plant Systems - Tier 2/Group 2 ______ F V 1 1-ri 1 4 KI K2 K3 K4 K5 K6 Al A2 A3 X A4 G A3.05 Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including: Reactor water level F r T 1 1 1 1-t-T F F .1 Suppressed _s Suppressed 202001 Recirculation 204000 RWCU 205000 Shut (RHR Shutdc 214000 RPIS 215002 RBM down Cooling System wn Cooling mode) Suppressed I I I +/- + F F F F F + X I I r F F F + 4 4-x r F F +-+ +/- I 4 4 4-X F t 1 1 I 4 4 F F F F F F 1* 1 - t-1 4 4 F F F K4.01 Knowledge of RECIRCULATION System design feature(s) and/or interlocks which provide for the following: 2/3 core coverage A3.03 Ability to monitor automatic operations of the REACTOR WATER CLEANUP SYSTEM including:

Response

to system isolations A2.02 Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low shutdown cooling suction pressure Suppressed Suppressed NUREG-1021, Revision 8, Supplement 1 K/A Tooic(s) Form ES-401-1 Inn P ints !I 2.8 _____i -I__~ I__L__ Perry Final - Revision 1 Page 9 of 15 K/A TODiC(Sl I Imp ý Poin

ES-401 System # / Name 215003 IRM 219000 RHR/LPCI: Torus/Pool Cooling Mode K1 BWR SRO Examination Outline Plant Svstems - Tier 2/Group 2 Form ES-401-1 9 '4 K2 K3 K4 K5 K6 A1 A2 X A3 A4 G 4 4 - -4 4 4 9 1-X 4 4 4 4 - +/- + 4 4 4 9 '4 230000 RHR/LPCI: Torus/Pool Spray Mode -I- - 9 4 9 4-234000 Fuel Handling Equipment 239003 MSIV Leakage Control 245000 Main Turbine Gen. And Auxiliaries X K/A Tocicts) A2.06 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty Ranqe switch K1.04 Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE and the following: LPCI/RHR pumps Imp. Points 32 39 1 S3.9 1 Suppressed A1.01 Ability to predict and/or monitor changes in parameters associated with operating the FUEL HANDLING EQUIPMENT controls including: Spent Fuel Pool level 3.4 + + I I I I I I 1 ______ t t* r + + '4 9 I I I t X Suppressed A2.07 Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of reactor/turbine pressure control systems 39 A1.02 Ability to predict and/or monitor 259001 Reactor Feedwater changes in parameters associated with operating the REACTOR FEEDWATER 3.3 SYSTEM controls including: Feedwater inlet temperature K4.01 Knowledge of UNINTERRUPTIBLE 262002 UPS (AC/DC) POWER SUPPLY (AC/DC) design feature(s) X and/or interlocks which provide for the 3.4 1 following: Transfer from preferred power to alternate power supplies NUREG-1021, Revision 8, Supplement 1 Page 10 of 15 Perry Final - Revision I K4 A2 L I ,,v

BWR SRO Examination Outline Plant Svstems - Tier 2/Grouo 2 System # / Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KJA Topic(s) Imp. Points I 2.4.45 Ability to prioritize and interpret 263000 DC Electrical Distribution X the significance of each annunciator or 3.6 alarm A2.02 Ability to (a) predict the impacts of 271000 Offgas the following on the OFFGAS SYSTEM and (b) based on those predictions, use X procedures to correct, control, or mitigate 3.1 1 the consequences of those abnormal conditions or operations: Low dilution steam flow K6.01 Knowledge of the effect that a loss or 272000 Radiation Monitoring malfunction of the following will have on the 2 RADIATION MONITORING SYSTEM: Reactor Protection System 286000 Fire Protection I 290003 Control Room HVAC 300000 Instrument Air 400000 Component Cooling Water II K!A Category Point Totals: 0 0 2 0 2 4 2 0 Group Point Total: NUREG-1021, Revision 8, Supplement 1 ES-401 Form ES-401-1 13 if Perry Final - Revision I 1 1 1 Page 1 1 of 15

ES-401 BWR SRO Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 3 System # / Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G K/A Topic(s)

Imp, Points K5.04 Knowledge of the operational 201003 Control Rod and Drive X

implications of the following concepts as 134 Mechanism I they apply to CONTROL ROD AND DRIVE ~MECHANISM: Rod sequence patterns 215001 Traversing In-Core Probe Suppressed s K4.06 Knowledge of FUEL POOL COOLING 233000 Fuel Pool Cooling and Cleanup AND CLEANUP design feature(s) and/orI X AN LAU einfaues n/r 3.2 1 interlocks which provide for the following: Maintenance of adequate pool level A4.09 Ability to manually operate and/or 239001 Main and Reheat Steam X monitor in the control room: Reactor 3.9 1 _____pressure A1.04 Ability to predict and/or monitor changes in parameters associated with 256000 Reactor Condensate X operating the REACTOR CONDENSATE 2 9 SYSTEM controls including: Hotwell level 268000 Radwaste 288000 Plant Ventilation 290002 Rector Vessel Internals K/A Category Point Totals: 0 01 0 1 0 0 1 0 Group Point Total: 4 Plant-Specific Priorities System Topic Recommended Replacement for Reason Points 2-23001 A4.06 Primary Containment and Auxiliaries /Containment pressure 223001 K6.12 Plant Specific high importance PSA for manually 1 venting containment on high pressure 217000A4.04 RCIC / Manually initiated controls 217000K5.03 Plant Specific hiqh importance PSA requiring manual 1 initiation of RCI C Plant-Specific Priority Total (limit 10) 2 PPge 12 of 15 Perry Final - Revision 1 NUREG-1021, Revision 8, Supplement 1

ES-4( Generic Knowledge and Abilities Outline (Tier 3) FORM ES 401-5 Category Conduct of Operations Equipment Control Radiation Control Total K/A# 2.1.11 Topic Knowledge of less than one hour technical specification action statements for systems Imp. 3.8 Points I 2.1,14 Knowledge of system status criteria which require the notification of plant personnel 3.3 1 II 2 1 29 Knowledge of how to conduct and verify valve lineups 3.3 1 2.1.30 Abiiity to locate and operate components / including local controls 3.4 1 Total 2.22 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels 4 I 351 2.2.12 Knowledge of surveillance procedures 3.4 1 2.2.13 Knowledge of tagging and clearance procedures 3.8 2.2.29 Knowledge of SRO fuel handling responsibilities 2.2.34 Knowledge of the process for determining the internal and external effects on core reactivity Knowledge of I0CFR20 and related facility radiation control requirements Knowledge of the process for performing a containment purge NUREG-1021, Revision 8, Supplement 1 Tnf*l Page 13 of 15 Perry Final - Revision 1 I Total I I 1 ES-4031 Generic Knowledge and Abilities Outline (Tier 3) FORM ES-401-5 1

Emergency Procedures/Plan Tier 3 Point Total (RO/SRO) 2.4.39 r r Knowledge of the RO's responsibilities in Emergency Plan implementation 3.1 I 2.4.11 Knowledge of abnormal condition procedures 3.6 1 2.4.3 Ability to identify post accident instrumentation 3.8 1 2.4.41 Knowledge of the emergency action level thresholds and classifications 4.1 1 2.4 12 Knowledge of general operating crew responsibilities during emergency operations 3.9 1 1 i. 2.4.19 Knowledge of EOP layout/ symbols/ and icons 3.7 I F 6 Total I1 NUREG-1021, Revision 8, Supplement 1 Perry Final-Revision 1 Page 14 of 15 L_17 ___j ,3.1 1 ii

Kecord OT i-(electeo rJAS Or~cncminl, £eiprftpc K/A R-nH-mI Selected K/A 5OOnnOn.A1 05 Reason for Rejection No wetwell soravs at facilitv rnnr-l ewl spay at7 fac7. iity 295016AA1.03 2__ .1-- 2180IOA1 02 RPIS not relevant to selected Ai-'_ tor taciity Facility does not have ADS acoustic sensors 4 ?sqNNA? No WY'n)OOVA 11 Facilitv does not have FWCI (BWR-6) -cIiItv does not nave i-<MK LOOO seiecuon icuic iIivvr-uJ 2 / 2 300000K3.03 Single unit facility does not have unit cross-ties for instrument air 2/3 239001K1. 13 System interaction with suppressed system (MSIV Leakage control) 1 /1 295037EA1.07 EPE topic area interfacing with suppressed system (RMCS) 1!1 295014AK1.04 Facility does not use PCIOMR 2 / 1 223001K6.12 Replaced by plant specific priority 2/1 217000K5.03 Replaced by plant specific priority This APE should have been a 'common' question (RO and SRO). This change will reduce the total number of exam questions by one. The 1 / 1 295007AK2.04 new topic (AA2.01) also has a higher Importance Rating. After further scrutiny unable to write a viable exam question. It was determined this APE topic not applicable to Perry. The new topic 1/2 600000AA2.06 (AA2.15) also has a higher Importance Rating. 2/2 /2 271000A2.14 After further scrutiny, unable to write a viable exam question. The new topic (A2.02) also has a higher Importance Rating. I__ _ ~~I Perry Final - Revision 1 NUREG 1021, Revision 8, Supplement 1 ES-401 p Tier/Grr '1/1 -1/1 -2/1 2/1 o/ 1 Facilitv does not have RHRI Loop selection logic tB VV-b) Page 15 of 15 r-I C: C' A -1 -1 r) Pecorr' of Reiected K/As t-Uo II E - -IU I-U m I Ou

ES-301 Administrative Topics Outline Form-ES-3 01-1 Facility: Perry Date of Examination: 3/4/2002 Examination Level: RO Operating Test Number: 1 Administrative Describe method of evaluation: Topic/Subject

1. ONE Administrative JPM, OR Description

2. TWO Administrative Questions A. 1 Shift Turnover 2.1.3 (3.0) - Knowledge of Shift Turnover Requirements JPM: Complete a Shift Relief/Turnover checklist as the oncoming operator Jet Pump 2.1.7 (3.7) - Ability to Evaluate Plant Performance and Make Operational Judgements Based on Operability Operating Characteristics / Reactor Behavior / Instrument Interpretation JPM: Determine Jet Pump operability A.2 Tagging 2.2.13 (3.6) - Knowledge of Tagging and Clearance Procedures JPM: Establish equipment isolation boundaries A.3 Radiation Control 2.3.1 (2.6) - Knowledge of IOCFR20 and Related Facility Radiation Control Requirements Requirements JPM: Evaluate requirements of RWP and perform appropriate actions for personnel contamination A.4 Personnel 2.4.39 (3.3) - Knowledge of RO's Responsibilities in Emergency Plan Implementation Accountability JPM: Perform Site Accountability actions from outside the Control Room.

NUREG-1021, Revision 8 Final-Revision I

ES-301 Administrative Topics Outline Form-ES-301-1 Facility: Perry Date of Examination: 3/4/2002 Examination Level: SRO Operating Test Number: 1 Administrative Describe method of evaluation: Topic/Subject

1. ONE Administrative JPM, OR Description

2. TWO Administrative Questions A. 1 Shift Turnover 2.1.3 (3.4) - Knowledge of Shift Turnover Requirements JPM: Complete a Shift Relief/Turnover Checklist as the oncoming operator Feedwater 2.1.7 (4.4) - Ability to Evaluate Plant Performance and Make Operational Judgements Based on Temperature Operating Characteristics / Reactor Behavior / Instrument Interpretation Reduction Ops JPM: Prepare for Feedwater Temperature Reduction Operations A.2 Risk Assessment 2.2.17 (3.5) - Knowledge of the Process for Managing Maintenance Activities During Power Operations JPM: Perform an On-Line Risk Determination A.3 Radiation Control 2.3.1 (3.0) - Knowledge of 10CFR20 and Related Facility Radiation Control Requirements requirements JPM: Evaluate requirements of RWP and perform appropriate actions for personnel contamination A.4 Emergency Plan 2.4.29 (4.0) - Knowledge of the Emergency Plan JPM: Classify an Emergency event, make Protective Action Recommendations, and complete paperwork for notification of Off-Site authorities NUREG-1021, Revision 8 Final - Revision 0

ES-301 Control Room Systems and Facility Walk-Through Test Outline Form-301-2 Facility: Pe Date of Examination: 3/4/2002 Exam Level RO/SROI Operating Test No.: 1 B.1: Control Room Systems System JPM Description Type Safety Code* Function RRS Shift Recirculation Pump B from Slow Speed to SNA 1 Si 202001 Fast Speed and Raise Reactor Power using Recirculation Flow CRDH CRD Alternate Injection for Level Control MSA 2 S2 201001/295031 Control Room HVAC Shift CR HVAC & Emergency Recirculation from SD 9 S3 290003 Emergency to Normal RCIC RCIC Startup from Standby Readiness (CST to MASL 4 S4 217000 CST) RHR/LPCI Terminate Containment Spray RHR Loop A NS 5 S5 226001 DG Remotely Transfer Bus EH 12 to the Alternate SN 6 S6 264000 Preferred Source from the DG MRSS Opening Inboard Main Steam Line Drain Valve SN 3 S7 239001 B.2: Facility Walk-Through SLCS Commence Alternate Boron Injection DR 1 P1 211000 RHR Perform RHR Loop B Alternate Injection DR 2 P2 203000/295031 Fire Protection Initiate CR Subfloor C02 from Outside CR MA 8 P3 286000

Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol Room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021, Revision 8 Final - Revision 1

ES-301 Control Room Systems and Facility Walk-Through Test Outline Form-301-2 Facility: Per1y Date of Examination: 3/4/2002 Exam Level: SROU Operating Test No.: 1 B.1: Control Room Systems System JPM Description Type Safety Code* Function Si S2 S3 RCIC RCIC Startup from Standby Readiness (CST to MASL 4 S4 217000 CST) S5 DG Remotely Transfer Bus EH12 to the Alternate SN 6 S6 264000 Preferred Source from the DG MRSS Opening Inboard Main Steam Line Drain Valve SN 3 S7 239001 B.2: Facility Walk-Through P1 RHR Perform RHR Loop B Alternate Injection DR 2 P2 203000/295031 Fire Protection Initiate CR Subfloor C02 from Outside CR MA 8 P3 286000 NUREG-1021, Revision 8

'ype Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol Room, (S)imulator, (L)ow-Power, (R)CA Final - Revision 0

Appendix D Scenario Outline Form ES-D-1 Facility: Perry Scenario No.: 1 Examiners: Op-Test No.: 2002-01 Operators: Initial Conditions: A Reactor startup is in progress following a brief outage. Reactor power is being held at 70% power per SCC request. A xenon transient is in progress. RHR B is in secured status for preventive maintenance on the pump breaker. RHR B was declared inoperable five hours ago per Tech. Spec.3.5.1, Action A; 3.6.1.7, Action A; and 3.6.2.3, Action A. The MFP is in secured status to support recirc valve actuator work. SRV F041 E is weeping and causing the Suppression Pool to slowly heatup. The OPRMs are functional but are inoperable per Tech. Spec. 3.3.1.3. Required Action A.3 has been implemented. Turnover: 1. BOP operator place RHR A in suppression pool cooling mode and lower Suppression Pool temperature to 75 F. ESW A and ECC A are in operation. 2. Maintain 70% power. Target Critical Tasks: Emergency Depressurization, Restore RPV water level Event Malf. No. Event Event No. Type* Description 1 N (BOP) Startup RHR A in Suppression Pool cooling mode 2 MV06: C (BOP) RHR Min Flow valve (F064A) fails due to mechanical binding after RHR flow is established (TS 3.5.1.C, 3.6.11.7.B, 3.6.2.3.8) 1 E 12-F064A 3 CN02: I (BOP) CRDH flow controller failure in Auto mode 1C11R0600 100% RD12R1447 C (RO) Single control rod drift inward (14-47) 4 Various C (RO) Reactor Feed Pump A bearing failure R (RO) Lower reactor power to 63% using recirc flow N (RO) Remove RFPT A from service ZL1N27R0425A I (RO) RFP A manual speed control dial pot failure (as is) 5 FW02 -50% M (All) Feedwater System Pipe Break inside Drywell I Reactor Scram CP01: C (BOP) HPCS Pump shaft breaks 1E22C0001 MV04: 1E51F0013 C (BOP) RCIC Injection Valve (F01 3) failure to Auto open RC03 C (BOP) RCIC Turbine mechanical trip latch failure 6 TH02C C (All) Recirc Bottom Head Drain pipe break (5 minute ramp) 75% ZD1 B21 S34B C (BOP) ADS B Inhibit Switch failure in Normal position 7 M RPV emergency depressurization / Inject with I~w pressure ECCS to maintain adequate core cooling

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Final - Revision 1

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 1 Objectives: The BOP operator places RHR A in Suppression Pool Cooling mode to lower Suppression Pool temperature to 75 F. When RHR is initiated, the RHR minimum flow valve will remain open after flow is established. This will make RHR inoperable and the Tech. Specs. will be consulted. Immediately after Tech Specs are referenced, the CRDH Flow Controller will fail in the Auto mode causing a single control rod to drift inward. The crew will recognize the drifting control rod and enter ONI-C51, Unplanned Change in Reactor Power or Reactivity. When the crew identifies the failed CRDH Flow Controller, the controller will be placed in Manual and CRDH parameters will be returned to normal allowing the drifted control rod to settle at notch 00. After a plan is implemented to recover the drifted control rod, RFP A will experience sustained high bearing vibration requiring reactor power to be lowered to 63% using recirc flow to allow the RFP to be removed from service. After the power reduction, as RFP A is being removed from service, the manual speed control dial potentiometer will fail requiring a trip of the RFP to remove it from service. When RFP A is tripped, a feedwater pipe breaks in the Drywell resulting in a reactor scram. Following the scram, the HPCS pump will be unavailable due to a shaft break. When RCIC is initiated for level control, the RCIC injection valve will fail to open automatically requiring operator action to manually open the injection valve. After injection is established with RCIC, the RCIC Turbine mechanical trip latch will fail making RCIC unavailable for injection, A small Recirc pipe break in the Drywell will develop and slowly increase in severity resulting in rising drywell temperature and pressure and lowering RPV water level. As reactor level continues to lower, the crew will place alternate injection systems in service, emergency depressurize the RPV, and maintain adequate core cooling with low pressure ECCS systems in accordance with PEI-B13, RPV Control (Non-ATWS). However, when ADS is inhibited, the ADS Inhibit Switch B will fail which may require the crew to take action to prevent an unintended ADS blowdown. Discussion of Safety Significance for scenario I The BOP operator will place RHR in Suppression Pool Cooling. When RHR flow is initiated, the RHR Min flow valve stays open after flow is established. The BOP operator must note that the min flow valve has failed to close and inform the SRO because with the minimum flow valve failed open, RHR A is inoperable. Final - Revision 0 Page I of 2

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario I The CRDH Flow Controller will fail open in the Auto mode causing a single control rod to drift inward. The RO must recognize that a control rod is drifting and determine which rod is drifting. The drifting control rod is safety significant because it directly effects core reactivity and core power distribution. Operators will be required to determine the cause of the drifting control rod (failed CRDH flow controller) and begin actions to recover the control rod. After a plan is implemented to recover the drifted control rod, RFP A will experience sustained high bearing vibration. This is safety significant because it will require the operators to lower reactor power in a controlled manner and remove the RFP from service. Removing the RFP from service is safety significant because the operation directly affects RPV water level control. As the RFP is being removed from service, the manual speed control dial pot will fail 'as is' requiring a trip of the RFP, directly leading to a Feedwater System pipe break in the Drywell and a reactor scram. Following the scram, the HPCS pump will be unavailable due to a shaft break. When RCIC is initiated for level control, the RCIC injection valve will fail to open automatically. The RCIC injection valve failure is safety significant because RCIC is the only normal high-pressure injection system available. Operator action will be required to manually open the injection valve allowing RCIC injection. After manual flow control is established with RCIC, the RCIC Turbine mechanical trip latch will fail making RCIC unavailable for injection. A small Recirc pipe break in the Drywell will develop and slowly increase in severity resulting in rising Drywell temperature and pressure and lowering RPV water level. The Recirc pipe break is safety significant because it directly contributes to the loss of coolant inventory requiring the crew to place alternate injection systems in service and prepare for RPV emergency depressurization to allow injection with low pressure ECCS systems. As RPV level continues to lower and ADS is inhibited, ADS Inhibit Switch B will fail. Failure of the capability to inhibit ADS is safety significant because it places the crew in a condition not anticipated by PEI-B 13, RPV Control (Non-A TWS), requiring the crew to take action to prevent unintended RPV emergency depressurization. Final - Revision 0 Page 2 of 2

Appendix D Scenario Outline Form ES-D-1 Facility: Perry Scenario No.: 2 Examiners: Op-Test No.: 2002-01 Operators: Initial Conditions: The plant is operating at 100% power. A xenon transient is in progress. RHR B is in secured status for preventive maintenance on the pump breaker. RHR B was declared inoperable five hours ago per Tech. Spec.3.5.1, Action A; 3.6.1.7, Action A; and 3.6.2.3, Action A. The OPRMs are functional but are inoperable per Tech. Spec. 3.3.1.3. Required Action A.3 has been implemented. HPCS testing is scheduled to support flow rate testing. Turnover: 1. Place HPCS in full flow test mode to the suppression pool. HPCS ESW and HPCS Pump Room Cooler are in operation. 2. Maintain 100% power. Target Critical Tasks: Initiate action to shutdown the reactor, Inhibit ADS, Terminate and Prevent injection into the RPV, Emergency Depressurization, Restore RPV water level. Event Malf. No. Event Event No. Type* Description 1 N (BOP) Start HPCS in full flow test mode to the Suppression Pool 2 CP03: C (BOP) HPCS Pump flow degradation (2 minute ramp) (TS 3.5.1. B and C) 1 E22C0001 100% 3 AD01N C (BOP) ADS/SRV B21-F047H cycling (TS 3.5.1.E, F and H / TS 3.0.3) R (RO) Lower reactor power to 90% using recirc flow 4 CN03: I (RO) Reactor Feed Pump Controller B oscillations (1 minute ramp) 1C34R060B 20% 5 ED061 C (BOP) Loss of 480Vac Bus F-1 -E TC05 C (ALL) Turbine Control EHC leak / Main Turbine trip and reactor scram 10% 6 RDI 5 C (RO) Failure of RPS and ARI to automatically shutdown the reactor M (All) A'WS SL01A C (BOP) SLC Squib Valve failures, C41-FO04A and C41-FO04B SL01B 7 CB01: C (RO) All RFBPs trip lN27CO001A M (All) Loss of all Feedwater capability CBOl: 1N27C0001B CB01: 1N27C0001C CB01: 1N27C0001 D 8 M (All) RPV emergency depressurization / Inject with low pressure ECCS to maintain adequate core cooling RV04: C (BOP) ADS/SRV B21-F041F failure closed 11B21 F0041F

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Final - Revision I

'S 2002 Perry NRC Examination Scenario Objectives Safety Significance Discussions Scenario 2 Objectives: The BOP operator will place HPCS in the full flow test mode to support flowrate testing. After HPCS flow is stabilized for the test, HPCS Pump flow will be degraded requiring the HPCS System to be declared inoperable. After Tech. Specs. have been consulted, an ADS/SRV will open and cycle due to shorted switch contacts; the crew will enter ONI-B21-1, SRV Inadvertent Opening/Stuck Open, evacuate the Containment, reduce reactor power to 90% using recirc flow, and de-energize the SRV solenoids by removing control power fuses. After the SRV closes, Tech Specs must again be consulted and the loss of HPCS and ADS will require a Tech. Spec. 3.0.3 entry. The crew will respond to a RPV level transient due to Feedwater Flow Controller A oscillations that require entry into ONI-C34, Feedwater Flow Control Malfunction, and manual control of feedwater flow. After conditions have stabilized, a loss of 480 Volt Bus F-1-E will result in a loss of the running EHC Pump B, TBCC Pump C, and CVCW Chiller A (in addition to other loads). Following auto start of standby EHC Pump A, the EHC System will develop a leak which will slowly increase in severity until the reactor is manually scrammed and the Main Turbine is tripped or the Main Turbine trips automatically. When the reactor scrams, the control rods will fail to fully insert due to blockage in the scram discharge volume. PEI-B1 3 (RPV Control-ATWS) is entered and executed to stabilize the plant. SLC squib valves will fail to fire. After PEI ATWS actions are underway and RPV level has reached L2, (130" above TAF), Feedwater System capability is lost and the RCIC system will be unable to provide adequate makeup. Therefore, the crew will emergency depressurize the RPV to allow for low pressure ECCS injection. Two ADS/SRVs will fail to operate during the emergency depressurization. Discussion of Safety Significance for scenario 2 The BOP operator will place HPCS in the full flow test mode. After HPCS flow is stabilized for the test, the BOP operator must note that HPCS Pump flow is degraded. This is safety significant because the HPCS System is inoperable and unavailable for core cooling. Final - Revision 0 Page I of 2

r 2002 Perry NRC Examination Scenario Objectives Safety Significance Discussions Scenario 2 Next, an ADS/SRV will open and cycle due to shorted switch contacts; the crew will enter ONI-B21-1, evacuate the Containment, reduce reactor power, and deenergize the SRV solenoids by removing control power fuses. This is safety significant because the cycling SRV directly affects core reactivity and results in a rising Suppression Pool temperature. When SRV fuses are pulled, the crew must determine that the ADS/SRV is inoperable and unavailable. This is safety significant because it further degrades the status of the high-pressure ECCS systems, thus requiring entry into Tech. Spec. 3.0.3. The crew will then respond to a RPV level transient due to Feedwater Flow Controller A oscillations. This is significant because it requires manual control of RPV water level to terminate the level transient. After conditions have stabilized, a loss of 480 Volt Bus F-1-E will result in a loss of running EHC Pump B, TBCC Pump C, and CVCW Chiller A (in addition to other loads). This is safety significant because it requires the crew to evaluate plant status and enter the appropriate ONIs for the 480-Volt bus failure and TBCC Pump trip. Following the auto start of the standby EHC Pump A, a turbine control EHC leak will develop. This is safety significant because the crew must recognize the EHC leak, manually scram the reactor, and trip the Main Turbine before the Main Turbine trips automatically. When the reactor is scrammed, the control rods fail to fully insert due to blockage in the scram discharge volume. The failure of control rods to insert is safety significant because it will require the operators to take PEI A TWS actions to shutdown the reactor, control reactor level, and control reactor pressure. During A TWS actions, the SLC squib valves will fail to fire. This failure is safety significant because it will lengthen the time required to achieve reactor shutdown. Following the loss of the Feedwater System, the crew must determine that the RCIC System will be unable to provide adequate makeup, thereby challenging the ability to maintain adequate core cooling. This is safety significant because emergency depressurization will be required to establish controlled injection with low pressure ECCS to assure adequate core cooling. Two ADS/SRVs will fail to operate during the emergency depressurization. Failure of the ADS SRVs to open is safety significant because the crew must recognize the failure and open additional SRVs. Final - Revision 0 Page 2 of 2

"Appendix D Scenario Outline Form ES-D-1 Facility: Perry Scenario No.: 3 Examiners: Op-Test No.: 2002-01 Operators: Initial Conditions: Reactor startup is in progress with the plant at 5% of rated power. RHR B is in secured status for preventive maintenance on the pump breaker. The OPRMs are functional but are inoperable per Tech. Spec. 3.3.1.3. Required Action A.3 has been implemented. Turnover: Plant startup continues: withdraw control rods to 10% power, transfer the Reactor Mode Switch to RUN, and continue power ascension. All required MODE 1 change paperwork has been reviewed and approved. Target Critical Tasks: Emergency Depressurization, RPV Flooding to restore and maintain adequate core cooling Event Malf. No. Event Event No. Type* Description 1 R (RO) Increase reactor power to 10% using control rods RD01 :R1043 C (RO) Control rod 10-43 stuck at position 8 8% 2 NM02H I (RO) IRM H failure upscale (bypass failed IRM) 100% (TS 3.3.1.1 and OR 6.2.3) 3 N (RO) Verify NI overlap / Transfer Reactor Mode Switch to RUN I Withdraw IRMs 4 RD17A C (BOP) CRDH Pump A trip due to loss of lube oil. 50% N (BOP) Perform CRD Pump trip recovery RD05R5443 Accumulator fault HCU 54-43 (TS 3.1.5) (1 minute time delay) 5 CP02: C (BOP) Service Water Pump 'B' trip due to shaft seizure (start standby Service Water OP41 CO01 B Pump) 6 bat C (All) Seismic Event (OBE) or/seismic_2 RP01A C (All) RPS 'A' EPA Breaker Trip (loss of RPS Bus 'A') (30 second time delay) TH02A / TH02B M (All) Recirc Loop pipe rupture (reactor scram on high Drywell pressure) 100% (TH02A - 6 minute time delay & 5 minute ramp) (TH02B - 8 minute time delay & 5 minute ramp) MV08: C (BOP) NCC Drywell Isolation Valve P43-F215 failure when valve becomes fully dosed OP43F0215 RD01R4219 C (RO) Control rod 42-19 stuck at position 12 (during scram) 7 bat I (All) Loss of all RPV level indication ms/losslevel2 M (All) Emergency Depressurization / RPV Flooding to restore and maintain adequate core cooling

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Final - Revision 1

1 4 2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 3 Objectives: The crew will continue the startup. Prior to placing the Reactor Mode Switch in RUN, control rod 10-43 will not withdraw using normal drive water pressure. This will require the crew to take action per ONI-CI 1, Inability to Move Control Rods, and SOl-C11 (RCIS) to free the stuck control rod. After the control rod is moving normally, IRM H will fail upscale resulting in a RPS half-scram. The startup will be placed on hold while Tech Specs are referenced, the IRM is bypassed, and the half-scram is reset. The startup continues and after the Reactor Mode Switch has been placed in RUN, the running CRDH Pump will trip due to a loss of lube oil, requiring the standby CRDH Pump to be started. While performing CRD pump trip recovery, an HCU accumulator fault alarm will be received requiring the crew to monitor for additional accumulator fault alarms. After the standby CRDH Pump is started, Service Water Pump B will trip. ONI-P41, Loss of Service Water, will be entered requiring manual start of a standby Service Water Pump. Immediately after a standby Service Water Pump is started, a seismic event occurs which causes an RPS EPA breaker to trip and a Recirc Loop pipe break in the Drywell, resulting in rising Drywell pressure and temperature. NCC Drywell Isolation Valve P43-F215 will fail when it is fully closed further degrading the crew's ability to control Drywell temperature. The reactor will be manually scrammed or will automatically scram; however, one control rod will fail to insert. Following the scram, rising Drywell temperature will result in a loss of all level indication. RPV Flooding and Emergency Depressurization is performed and low pressure injection systems are used to maintain adequate core cooling. Discussion of Safety Significance for scenario 3: As the startup continues, prior to placing the Reactor Mode Switch in RUN, one control rod will not withdraw using normal drive water pressure requiring the crew to take action to get the control rod to move. This is safety significant because the actions directly affect core reactivity. After the control rod is unstuck, an IRM will fail upscale. This is safety significant because it will result in a RPS half-scram, require the crew to determine the IRM is inoperable, bypass the IRM, and reset the half-scram. Final - Revision 0 Page I of 2

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 3 As the startup continues and the Reactor Mode Switch has been placed in RUN, the running CRDH Pump will trip that requires the standby CRDH Pump to be started. An accumulator fault alarm will also occur. This is safety significant because Tech. Specs. would require a manual reactor scram after 20 minutes if an additional HCU accumulator fault alarm were to be received. After the standby CRDH Pump is started, a Service Water Pump will trip. This will require a manual start of a standby Service Water Pump to avoid high temperatures on components and systems cooled by the Service Water System. After a standby Service Water Pump has been started, a seismic event occurs which causes a RPS EPA breaker to trip and a Recirc pipe break in the Drywell. This is safety significant because the reactor will be manually scrammed or will automatically scram following the seismic event. One control rod will fail to insert following the scram. The failure of the control rod to insert is safety significant because it will require the crew to make a timely assessment of control rod positions to determine if A TWS actions are necessary. Following the scram, conditions in the Drywell will cause a loss of all level indication. This is safety significant because the crew must enter RPV Flooding and emergency depressurize the RPV in order to allow low-pressure injection systems to be used to maintain adequate core cooling. Final - Revision 0 Page 2 of 2

Appendix D Scenario Outline Form ES-D-1 Facility: Perry Scenario No.: 4 Examiners: Op-Test No.: 2002-01 Operators: Initial Conditions: The plant is operating at 100% power. RHR B is in secured status for preventive maintenance on the pump breaker. RHR B was declared inoperable five hours ago per Tech. Spec.3.5.1, Action A; 3.6.1.7, Action A; and 3.6.2.3, Action A. The OPRMs are functional but are inoperable per Tech. Spec. 3.3.1.3. Required Action A.3 has been implemented. Turnover: 1. Shift NCC Pumps (start NCC Pump C and shutdown NCC Pump A). Target Critical Tasks: Manually start RHR Pump A (failure to auto start), Emergency Depressurization Event Malf. No. Event Event No. Type* Description 1 N (BOP) Shift NCC Pumps (start NCC Pump C and shutdown NCC Pump A) 2 CN02: I (BOP) RFPT A Lube Oil Temp controller failure in Auto mode 1 P44R0450 0% 3 PT02: I (RO) Reactor Narrow Range Level Transmitter N004A Offset (3 minute ramp) 1C34N0004A (ORM 6.2.1.3) 17% 4 RF FW66 C (RO) RFPT A spurious trip TH12B C (RO) Reactor Recirculation FCV B runback failure (TS 3.4.1) 5 CP02: C (BOP) TBCC Pump A trip (start standby TBCC Pump) 1 P44C0001A SW03 C (ALL) TBCC System Process Piping Leakage (1 minute time delay and 3 minute ramp) 25% Fast reactor shutdown required R (RO) Decrease reactor power to 66% using recirc flow (58 Mlbs/hr) 6 TH28 M (All) MSL Break in Drywell 1% PC01A DWICNTMT Bypass Leakage (to be modified in Event #7) 0% CB04: C (BOP) RHR Pump A fails to auto start on Drywell high pressure (required for 1E12C0002A Containment Spray mode) 7 CB01: C (BOP) RHR Pump A trips when flow is aligned to containment spray 1E12CO002A M (All) RPV emergency depressurization to control Containment pressure PT01: I (RO) Reactor Level Transmitter N081 C failure downscale 1B21N0081C 5%

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Final - Revision 1

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 4 Objectives: With the plant operating at 100% power, the BOP operator will shift NCC Pumps. After NCC Pump C has been placed in service and NCC pump A has been shutdown, RFPT A Lube Oil Temp controller will fail closed in Auto requiring the crew to place the controller in Manual and restore lube oil temperature to normal. After lube oil temperature is restored, the Narrow Range level transmitter for the in-service channel fails low (offset). The crew enters ONI-C34, Feedwater Flow Control Malfunction, to select an operable Narrow Range level channel and restore level to normal. When conditions have stabilized, RFPT A will trip. During the transient, Reactor Recirculation FCV B will fail to automatically runback requiring Technical Specifications to be referenced for a loop flow mismatch. After level and power are stabilized, TBCC Pump A will trip. ONI-P44, Loss of TBCC, will be entered requiring manual start of the standby TBCC Pump. Immediately following start of the standby TBCC Pump, a TBCC System piping leak will be initiated and grow progressively worse until a fast reactor shutdown is required. Following the scram, a MSL pipe break in the Drywell occurs resulting in an MSIV isolation and rising Containment pressure. When signaled to start on high Drywell pressure, RHR Pump A will not automatically start and must be manually started. When aligned for containment spray, RHR Pump A breaker will trip. Eventually the RPV must be depressurized to control the Containment pressure rise. During emergency depressurization, Reactor Level Transmitter N081C will fail downscale. Discussion of Safety Significance for scenario 4 After NCC Pump C has been placed in service and NCC Pump A has been shutdown, RFPT A Lube Oil Temp controller will fail closed. This is safety significant because failure to recognize and correct the failure would eventually result in RFP bearing damage and a loss of the RFP, thereby challenging RPV water level control. Final - Revision 0 Page I of 2

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 4 After lube oil temperature is restored, the Narrow range level transmitter for the in-service channel fails low (off set). This is safety significant because it will require manual operation of the Feedwater System to select an operable Narrow Range level channel and return RPV level to normal. When conditions have stabilized, RFPTA will trip and during the transient, Reactor Recirculation FCV B will fail to automatically runback. The RFPT A trip is safety significant because a reactor scram could result if the expected plant response is not verified. Failure of Reactor Recirculation FCV B to automatically runback is safety significant because Recirc Loop Flows will not be matched as required by Tech. Specs. After level and power are stabilized, TBCC Pump A will trip and operator action will be required to manual start the standby TBCC Pump to allow for continued plant operation. Manual start of the standby TBCC Pump will also lead to a TBCC System piping leak that will grow progressively worse and require the crew to initiate a fast reactor shutdown. Following the scram, a MSL pipe break in the Drywell occurs resulting in an MSIV isolation and rising Containment pressure. This is safety significant because PEI-B 13, RPV Control (Non-A TWS,) and PEI-T23, Containment Control, must be entered and executed to maintain key Containment and RPV parameters. When signaled to start on high Drywell pressure, RHR Pump A will not automatically start. This is safety significant because RHR Pump A must be manually started to assure it will perform its LPCI design function if required. The RHR Pump A trip is safety significant because the crew must determine that RHR flow will be unavailable for containment spray, requiring RPV emergency depressurization to control the Containment pressure rise. The Reactor Level Transmitter downscale failure is safety significant because the failure could complicate RPV level control during and following the emergency depressurization. Final - Revision 0 14 Page 2 of 2

Perry Plant Initial License NRC examination Operating Examination Risk Significance Each candidate will be examined in a dynamic simulator setting on a minimum of 2 events identified in the Perry IPE. The dynamic simulator evaluation contains the following Perry IPE events. Transient with a Loss of Power Conversion System: "* MSIV Isolation "* Various High Pressure Injection Systems unavailable "* RPV Emergency Depressurization required "* RPV level control with the ECCS low pressure systems Intermediate LOCA Events: "* Steam Leak in Drywell "* Steam Break in Containment "* Failure of Long-Term Containment Heat Removal with RHR ATWS: "* Failure to Insert All Control Rods "* Failure of RPV Level Control "* Failure of Standby Liquid Control (SLC) System "* RPV Emergency Depressurization required "* RPV level control with the ECCS low pressure systems Additionally, two of four scenarios contain transient events that could lead to Perry IPE events if operator action is inappropriate or ineffective: "* IORV transient "* Trip of Service Water Pump SRO candidates will be required to determine Risk related to equipment availability as part of the Administrative Examination. Confidential Page 1 11/15/01

PERRY INITIAL LICENSE EXAM MARCH 5 THRU 13, 2002 NRC Comments and Resolution on licensee submitted test outlines

Comments on the PERRY Exam Outlines General Systems Deleted: General System 215001, "Traversing In-Core Probe" REQUIRED CHANGE1 System 290001, "Secondary Containment" REQUIRED CHANGE, Comment(s) NRC: LICENSEE

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NRC: LICENSEE

RESPONSE

Need a copy of the K/As/Systems deselected during the development of the outlines. Per discussions with the licensee and the contractor hired to develop the exam outlines, no K/As were "deselected" prior to developing the outline. Although licensed operators at Perry do not operate the system, knowledge of how the system interfaces with NIs, RPS, etc. would be testable knowledge. It may be appropriate to eliminate some K/As under this system but removing the entire system is not. The licensee took action to reintroduce this system in the K/A selection process. A partial random re selection was performed. This system was not selected. The system will be included in future outline development activities. This system contains K/As that specifically refer to BWR-6 design. Although it may be appropriate to eliminate some K/As under this system, eliminating the entire system is not. The licensee took action to reintroduce this system in the K/A selection process. A partial random re selection was performed. This system was not selected. The system will be included in future outline development activities. Sheet 1 of 2

Comments on the PERRY Exam Outlines JPM/Scenario Event Number A.2 (SRO) A.2 (SRO) (perform risk assessment) Comment All Scenarios Comment(s) NRC: LICENSEE

RESPONSE

NRC: LICENSEE

RESPONSE

The task should require the applicant to make an "SRO level" decision concerning the performance of maintenance. The licensee understood and will ensure the JPM includes a verifiable action that allows the examiners to evaluate the applicant's ability to perform "SRO level" actions in regards to maintenance planning. Although it appears that there are more than a sufficient number of events included in the scenarios, several of the events will be combined due to the difficulty in crediting individual actions once the event has been initiated. For example, in Scenario 1, Event 4, the RO would not be credited with an "N" for removing the RFPT from service, this is part and parcel to mitigating the event (RFPT bearing failure). The licensee understood. This issue will be revisited during review of the "as submitted" exam material. Sheet 2 of 2 I i}}

ML021190438

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