ML031130114

Draft - Outlines

ML031130114
Person / Time
Site:	Hope Creek
Issue date:	04/01/2003
From:	Reid J Public Service Enterprise Group
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
50-354/03-301 50-354/03-301
Download: ML031130114 (18)

Text

Alan Blarney Chief Examiner Division of Reactor Safety US Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 HOPE CREEK LIMITED SRO LICENSE EXAMINATION OUTLINE Enclosed is our proposed outline for the LSRO license examination to be conducted for Hope Creek candidates during the week of March 10 and 17, 2003. Included are:

• Form ES-201-2, Examination Outline Quality Checklist: Form ES-201 was pen and ink changed to reflect LSRO Exam requirements of ES-701.

Proposed Schedule: Currently there are 4 LSRO candidates. The operating exam is estimated to take 2 - 3 days. The written examination will be given the day after completion of the operating exam.

LSRO Written Examination Outline: The 50 question Written Exam outline for the LSRO exam was randomly generated using the "token method". K/A's that were not consistent with ES-701 Attachment 1 were reselected using the random process.

LSRO Administrative Topics Outlines: There are 3 Admin JPMs and 2 sets of 2 questions outlined on Form ES-301-1.

Discussion Scenario Outlines: There are 2 scenarios, each outlined on Form ES-D-1 IAW ES-701.

FacilitV Walk-Through Test Outlines: There are 5 Plant JPMs outlined on Form ES-301-2 IAW ES-701.

The examination team is currently developing the written and operating examination. If you have any questions or comments, please call me at 856-339-3966. For major issues, the Operations Training Manager, Jim Reid, can be reached at 856-339-3896. Jim is on the Examination Security Agreement.

Sincerely, I Archie E. Faulkner Training Supervisor /Exam Development

BWR LSRO NRC Examination Outline Facility: Date of Exam: Exam Level:

Hope Creek 3/10/2003 LSRO K/A Category Points Point Tier Group K K K K K K A A A A G Total 1 l 2 l 3 l 4 5 6 1 2 3 4

• Eegny&&

Emergency 1 0 l 1 l0 l 1 l2 l 2 6 Abnormal Plant 2

2 0 5 g j4 2 T 1 14 Evolutions EvltosTotals Tier 2U 2 1 5 5 4 ME'. 11 3 20 P2 Plant n1 0 0 0 0 0 1 0 0 0 0 1 2 Systems 2 1 1 0 2 2 1 0 2 0 0 0 9 3 1 0 0 0 0 2 0 1 0 0 0 4 Totals 3 1 0 1 2 4 0 3 0 0 1 15

3. Reactor and fuel characteristics and physical aspects of core 8 8 construction important to fuel handling or shutdown activities

4. Health Physics and Radiation Protection for fuel handling 7 7 activities and general employee responsibilities Note:

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

The final exam must total 50 points.

2. 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.

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

4. The shaded areas are not applicable to the category/tier.

5.

• 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.

6. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the SRO 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.

ES-401 ( BWR SRO Examiiwjcion Outline ( ES-401 -11 Emergency and Abnormal Evolutions - Tier 1/Group 1 System # Name Kl K2 K3A1 A2 G KA Topic(s) Imp. Pts.

295003 Partial or Complete Loss of A.C. Power lX AA1.01 AC. electrical distribution system 3.8 I 1 295003 Partial or Complete Loss of A.C. Power X 2.4.32 Knowledge of operator response to loss of all annunciators. 3.5 1 295006 SCRAM I 295007 High Reactor Pressure 295009 Low Reactor Water Level 295010 High Drywell Pressure 295013 JHigh Suppression Pool Temperature 295014 Inadvertent Reactivity Addition - AK2.05 Neutron monitoring system 4.1 1 295014 Inadvertent Reactivity Addition X AA2.01 Reactor power 4.2 1 295015 Incomplete SCRAM I 295016 Control Room Abandonment 295017 High Off-Site Release Rate 295023 Refueling Accidents .X, AA2.04 Occurrence of fuel handling accident 4.1 I 295023 'Refueling Accidents X 2.1.20 Ability to execute procedure steps. 4.2 1 I

295024 High Drywell Pressure 295025 High Reactor Pressure 295026 Suppression Pool High Water Temperature 295027 High Containment Temperature (Mark ill Containment Only) 295030 Low Suppression Pool Water Level Wednesday, December 18, 2002 8:09:02 AM Page 1

lES-401 BWR SRO Examination Outline

y and Abnormal Evolutions - Tier 1/Group 1 K ES-401 -1 System # Name I A2 G KA Topic(s) I Imp. IPts.

295031 Reactor Low Water Level 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 295038 High Off-Site Release Rate 500000 High Containment Hydrogen Concentration Wednesday, December 18, 2002 8:09:03 AM Page 2

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ES-401 BWR SRO Examination Outline K ES-401 -1 Emergency and Abnormal Evolutions - Tier 1/Group 2

,System # Name K1.K2 7K3 A1 A2 G KATopic(s) Imp. iPts.

295001 Partial or Complete Loss of Forced Core Flow X AA2.04 Individual jet pump flows: Not-BWR-1&2 3.1 1 Circulation 295001 Partial or Complete Loss of Forced Core Flow X 1.01 Recirculation system 3.6 1 Circulation 295002 Loss of Main Condenser Vacuum 295004 Partial or Complete Loss of D.C. Power 295005 Main Turbine Generator Trip 295008 High Reactor Water Level 295011 [High Containment Temperature (Mark ll I Containment Only) 295012 High Drywell Temperature 295018 Partial or Complete Loss of Component Cooling Water 295018 Partial or Complete Loss of Component 7( _ AK1.01 Effects on component/system operations 3.6 X AA2.01 Component temperatures Cooling Water - 3.4 1 295019 Partial or Complete Loss of Instrument Air ___

295020 Inadvertent Containment Isolation 295021 Loss of Shutdown Cooling 3.02 Feeding and bleeding reactor vessel I

295021 !Loss of Shutdown Cooling .1.04 Alternate heat removal methods 3.7 1 295021 Loss of Shutdown Cooling ,.35 Knowledge of local auxiliary operator tasks during emergency operations including _3.5 1 stem geography and system implications.

295022 Loss of CRD Pumps .3.02 CRDM high temperature I _

3.1 1 295022 Loss of CRD Pumps .1.02 Reactivity control 3.7 1 295028 High Drywell Temperature 295029 High Suppression Pool Water Level Wednesday, December 18, 2002 8:09:03 AM Page 3

i IES-401 l BWR SRO Exam6iduon Outline ES-401 -1 i Emergency and Abnormal Evolutions - Tier 1/Group 2

!System # IName Ki K2 K3 A1 A2GIKA Topic(s) Imp. Pts.1 295032 High Secondary Containment Area Temperature 295033 High Secondary Containment Area Radiation X IEK3.04 Personnel evacuation Levels 295033 High Secondary Containment Area Radiation - X EA1.01 Area radiation monitoring system 4.0 1 Levels 295034 Secondary Containment Ventilation High X IEK3.01 Isolating secondary containment ventilation 4.1 1 Radiation 295034 Secondary Containment Ventilation High X l EA1.02 Process radiation monitoring system l 4.0 1 Radiation 295035 Secondary Containment High Differential Pressure 295036 Secondary Containment High Sump/Area Water Level 600000 Plant Fire On Site i~

xi

.X EK3.04 Actions contained in the abnormal procedure for plant fire on site 3.4-L-Wednesday, December 18, 2002 8:09:03 AM Page 4

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!ES-401 BWR SRO ExamiK )n Outline k ES-401-1 Plant Systems - Tier 2/Group 1 System !Name K1 K2 K3K4 K5*K61AIA2TA3IA4 G KATopic(s) t -

lImp. - Pts. l 201005 Rod Control and Information System f I (RCIS)

- I I-I 202002 Recirculation Flow Control System I 203000 RHR/LPCI: Injection Mode (Plant Specific)

I 206000 High Pressure Coolant Injection System i 207000 Isolation (Emergency) Condenser

-i L-L-L 209001 Low Pressure Core Spray System 209002 High Pressure Core Spray System (HPCS) 211000 Standby Liquid Control System 212000 Reactor Protection System 215004 Source Range Monitor (SRM) System 2.2.32 Knowledge of RO duties in the control room during fuel handling such as 3.3 alarms from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

215005 Average Power Range Monitor/Local Power Range Monitor System 216000 Nuclear Boiler Instrumentation 217000 Reactor Core Isolation Cooling System (RCIC) 218000 Automatic Depressurization System 223001 Primary Containment System and Auxiliaries 223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off 226001 RHR/LPCI: Containment Spray System Mode 239002 Relief/Safety Valves 241000 Reactor/Turbine Pressure Regulating System Wednesday, December 18, 2002 8:10:02 AM Page 1

'ES-401 BWR SRO Exami( Jn Outline K ES-401-1 Plant Systems - Tier 2/Group 1 System [Name K1 K2 K3 K4 K5 K6!A1 A2TA3A4G KATopic(s) Imp. Pts.

259002 Reactor Water Level Control System 261000 Standby Gas Treatment System 262001 SA.C. Electrical Distribution 264000 Emergency Generators (Diesel/Jet) X- K6.09 D.C. power 3.5 1 290001 Secondary Containment l -

Wednesday, December 18, 2002 8:10:02 AM Page 2

ES-401 BWR SRO Examil in Outline K ES-401-1 Plant Systems - Tier 2/Group 2 System IName IK1 K2 K3 K4 K5 K6 A1lA2 A3 A4 G KATopic(s) Imp. Pts.

201001 [Control Rod Drive Hydraulic System X K5.03 Pressure indication 2.7 1 201002 Reactor Manual Control System 201004 Rod Sequence Control System (Plant Specific) 201006 Rod Worth Minimizer System (RWM)

(Plant Specific) 202001 Recirculation System X K4.01 2/3 core coverage: Plant-Specific 3.9 1 204000 Reactor Water Cleanup System i jX K.04 Heat exchanger operation 2.7 1 205000 Shutdown Cooling System (RHR X K6.03 Recirculation system 3.2 1 Shutdown Cooling Mode) 205000 Shutdown Cooling System (RHR X iA2.05 System isolation Shutdown Cooling Mode) 3.7 1 214000 Rod Position Information System -. i 215002 Rod Block Monitor System 215003 Intermediate Range Monitor (IRM) Systeem X iK2.01 IRM channels/detectors 2.7 1 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode _ _ _ I _ _

230000 RHR/LPCI: Torus/Suppression Pool Spray Mode 234000 Fuel Handling Equipment X K1.05 Reactor vessel components: Plant-Specific 3.3 1 239003 MSIV Leakage Control System 245000 Main Turbine Generator and Auxiliary Systems 259001 Reactor Feedwater System >P i 1 3 262002 Uninterruptable Power Supply (A.C./D.C 263000 D.C. Electrical Distribution

• 5 X I I 271000 Offgas System Wednesday, December 18, 2002 8:10:03 AM Page 3

ES-401 BWR SRO Examit on Outline Plant Systems - Tier 2/Group 2

( ES-401-1 System !Name K1 K2:K3 K4 K5K6A1,A2 A3A4 G KATopic(s) imp. PtsI' 272000 lRadiation Monitoring System X A2.01 Fuel element failure 4.1 1 1 286000 Fire Protection System j

290003 Control Room HVAC I

300000 1nstrument Air System (IAS) 3.9- 1 1 400000 Component Cooling Water System X 1K4.01 Automatic start of standby pump (CCWS) i j Wednesday, December 18, 2002 8:10:03 AM Page 4

ES-401 BWR SRO ExamK an Outline ES-401-1 l Plant Systems - Tier 2/Group 3 System 'Name K1 K2 K3 K4 K5 K6 A1lA2 A3 A4 G KATopic(s) imp. Pts.

201003 Control Rod and Drive Mechanism X K1.01 I Control rod drive hydraulic system 3.3 1 215001 Traversing In-Core Probe X K6.04 Primary containment isolation system: Mark-l&lI(Not- BWR1) I3.4 1 233000 Fuel Pool Cooling and Clean-up X A2.02 Low pool level 3.3 1 239001 IMain and Reheat Steam System I 256000 Reactor Condensate System j ___

268000 Radwaste.

I i 288000 Plant Ventilation Systems 290002 Reactor Vessel Internals II X K6.02 CRD mechanism 2.9 I1 J _1_ _ i .I Wednesday, December 18, 2002 8:10:03 AM Page 5

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ES-701 BWRH LSRO NRC Examination Outline Facility: Date of Exam: Exam Level:

Hope Creek 3/10/03 LSRO -

Category K/A# Topic Imp. Points 292001 K1.02 Reactor Theory - Neutrons - Define prompt and delayed 3.1 1 neutrons.

292003 K1.07 Reactor Theory - Reactor Kinetics and Neutron Sources - 3.3 1 l__ _ Explain prompt critical, prompt jump, and prompt drop.

292004 K1.05 Reactor Theory - Reactivity Coefficients - Define the Doppler 2.9 1 coefficient of reactivity.

3. Reactor and fuel 292005 K1.01 Reactor Theory - Reactor Control Rods - Relate notch and rod 3.3 1 characteristics and position.

physical aspects of 293008 K1.30 Reactor Theory - Reactor Operational Physics - Explain the 3.5 1 core construction relationship between decay heat generation and: a) power level important to fuel history, b) power production, and c) time since reactor handling or shutdown.

shutdown activities 293006 KM .13 Thermodynamics Theory - Fluid Statics - Explain the results of 2.7 1 putting centrifugal pumps in parallel or series combinaions 293008 K1.06 Thermodynamics Theory - Thermal Hydraulics - Define natural 2.6 1 convection heat transfer.

293010 K1.01 Thermodynamics Theory - Brittle Fracture and Vessel Thermal 2.8 1 Stress - State the brittle fracture mode of failure.

Total 8 G 2.3.1 Knowledge of 10CFR20 and related facility radiation control 3.0 1 requirements.

G 2.3.2 Knowledge of facility ALARA program. 2.9 1

4. Health Physics and G 2.3.4 Knowledge of radiation exposure limits and contamination 3.1 1 Radiation Protection control, including permissible levels in excess of those fradilatindPotciong authorized.

fuel vforhandling G 2.3.10 Ability to perform procedures to reduce excessive levels of 3.3 1 general employee radiation and guard against personnel exposure.

responsibilities G 2.1.4 Knowledge of shift staffing requirements. 3.4 1 r G 2.1.10 Knowledge of conditions and limitations of the facility license. 3.9 1 G 2.4.30 Knowledge of which events related to system operations/status 3.6 1 should be reported to outside agencies l7 Total 7

ES-301 Administrative Topics Outline Form-ES-301-1 Facility: Hope Creek Date of Examination: 3/17/03 Examination Level: SRO(L) Operating Test Number: 1 Administrative Describe method of evaluation:

Topic/Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A. 1 Conduct of 2.1.5 (3.4) - Ability to locate and use procedures and directives related to shift staffing Operations and activities JPM JPM: Apply working hour limitations for LSRO and platform operator Conduct of 2.1.24 (3.1) - Ability to obtain and interpret station electrical and mechanical drawings Operations JPM JPM: Demonstrate flowpath for Alternate Decay Heat Removal using D RHR Loop using P+lD's A.2 Equipment 2.2.26 (3.7) - Knowledge of refueling administrative requirements Control JPM JPM: Verify HC.OP-DL.ZZ-0026 log requirements for resuming Core Alterations 2.3. 5 (2.5) - Knowledge of use of personnel monitoring equipment QUESTION: Personnel contamination response A.3 RadiationControl 2.3. 10 (3.3) -Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure QUESTION: Requirements for removal of tools or equipment from the Spent Fuel Pool 2.4.40 (4.0) - Knowledge of SRO responsibilities in emergency plan implementation QUESTION: EP Event requiring Accountability of plant personnel A.4 Emergency Plan Questions 2.4.41 (4.1) - Knowledge of emergency action level thresholds and classifications QUESTION: EP Event classification for fuel handling event 03HCLSRONRCADMOUTL Revised-12/18/02, 9:17 AM NUREG-1021, Revision 8

ES-301 Control Room Systems and Facility Walk-Through Test Outline Form-301-2 Facility: Hope Creek Date of Examination: 3/17/03 Exam Level: SRO(L Operating Test No.: 1 B.1: Control Room Systems System JPMDecrptonType JPM Description Code* Safety Function S.1 205000 Shutdown Restoration of NSSSS Isolation Capability to SDC M IC/

Cooling System Valves IAW HC.OP-GP.SM-0001 DHR (RHR Shutdown Cooling Mode)

S.2 204000 Align RWCU for Alternate Heat Removal N, R, A AUX/

RWCU DHR S.3 234000 Perform Main Fuel Hoist checks lAW HC.OP- D, R, A FHE Fuel Handling ST.KE-0001 5.4.4.F through 5.4.4. I Systems S.4 234000 Perform Monorail Aux Hoist Controls Functional N, R FHE Fuel Handling Test HC.OP-FT.KE-0001 Section 5.4.1 through Systems 5.4.15 S.5 234000 Perform Fuel Grapple Interlocks Test lAW HC.OP- D, R FHE Fuel Handling ST.KE-0001 5.1 through 5.1.10 Systems B.2: Facility Walk-Through (Same as RO In-Plant Walkthrough)

P.1 NA NA NA NA P.2 NA NA NA NA NA NA P.3 NA NA

• 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, (E)OP/AB G:/LSRO/03HCLSRONRCJPMOUTL 12/18/02, 9:16 AM NUREG-1021, Revision 8

Appendix D Scenario Outline Form ES-D-1 Facility: Hope Creek Scenario No.: 2 Op Test No.: 1 Examiners: Candidates: LSRO

_LSRO LSRO LSRO Objectives: To evaluate the applicants' ability to recognize and address problems with refueling bridge interlocks. Recognize HC.OP-AB.CONT-0005 IRRADIATED FUEL DAMAGE entry and take required actions.

Initial Conditions: Core Alterations are in progress. A fuel bundle is in the Fuel Prep Machine, being re-assembled.

Turnover: You are the Refueling SRO. Double blade guide 31-28/29-26 is about to be removed from the core.

Event Malf. Event Event Evaluator No. No. Type* Description Guide 1 N/A N Double blade guide removed lAW Grapple is open with proper bail HC.OP-SO.KE-0001 alignment.

Grapple is centered over bail handle.

Grapple is lowered, Slack Cable light comes on.

Hoist position indication is consistent with seated blade guide.

Verifies proper location and orientation then engages grapple.

Grapple Engaged light is lit.

Raises double blade guide.

2 1 C Hoist Jam light comes on. Hoist Lowers hoist until Hoist Jam light movement stops. clears. Stops refueling operation until problem is resolved.

3 N/A M A fuel bundle in the fuel prep Notify Control Room.

machine is being re-assembled.

Several pins that were removed fall Implement actions of HC.OP-into the spent fuel storage rack and AB.CONT-0005 IRRADIATED rupture. FUEL DAMAGE.

Suspends all refueling operations.

HC LSRO ESG 2 Rev. 12/17/02 Page 1 of 2

4 N/A M Refuel Floor Exhaust Hi-Hi Evacuate the refuel floor.

Radiation alarms Recognizes FRVS auto start and Reactor Building Ventilation isolation setpoints.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor HC LSRO ESG 2 Rev. 12/17/02 Page 2 of 2

Appendix D Scenario Outline Form ES-D-1 Appendix 0 Scenario Outline Form ES-D-1 Facility: Hope Creek Scenario No.: 1 Op Test No.: I Examiners: Candidates: LSRO ILSRO

_LSRO

_L.RO Objectives: Evaluate applicants' response to a CRD Mechanism leak. Discuss the effects of lowering fuel pool level. Demonstrate knowledge of method to stop CRDM leak from above with CRB Initial Conditions: Operational Condition 5, core alterations in progress. All Control Rods are inserted except rod 30-31. The CRDM for 14-23 has been replaced after rebuild. The Reactor Mode Switch is Operable and locked in Refuel position. All SRMs are operable. Shutdown Margin requirements are met.

Turnover: You are the Refueling SRO. Control Rod Blade 14-23 needs to be removed and replaced from the reactor core using the Frame Mounted Aux Hoist. All fuel has been removed from the cell. A double blade guide is installed. The Control Rod Blade 14-23 is full out with the CRDM uncoupled from under-vessel.

Event No.

1 Malf.

No.

1 Event Type.

Event Description Evaluator Guide SRM A fails to zero (0) cps Reviews Tech Spec 3.9.2 for SRM Operability.

Determines core alterations may continue for 14-23.

2 N/A N Removal of the Control Rod Meets Tech Spec 3.9.10.1 Blade. requirements for a single control rod removal.

Discusses Restricted Core Operations Form (RCOF) to continue.

- Remove double blade guide.

- Remove fuel support piece.

- Uses CRB Grapple on Frame Mounted Aux Hoist to remove CRB.

HC LSRO ES 1 Rev. 12/17102 Page I of 2

I 3 NA M Under vessel crew reports water Recognizes cavity level would pouring out 14-23 CRDM flange. be lowering and takes actions They are unable to stop the of HC.OP-AB.COOL-0004 Fuel leak. Pool Cooling.

-Evacuates the Refuel Floor

- Notifies Control Room.

- Notifies Reactor Engineer.

- Notifies Radiation Protection.

Recognizes that the CRB needs to be placed back into the guide tube to bottom in order to stop leak. (Not required for full credit) 4 N/A M Reactor Engineer and OS Puts CRB back into guide tube concurs with placing CRB back and lowers to the bottom to stop into guide tube. the leak.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor HC LSRO ES 1 Rev. 12/17/02 Page 2 of 2