ML031130298

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Final - Outlines (Change Summaries - Optional)
ML031130298
Person / Time
Site: Hope Creek PSEG icon.png
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: ML031130298 (18)
v • d • e


Text

BWR LSRO NRC Examination Outline Facility:

Date of Exam:

Exam Level:

Hope Creek 3/10/2003 LSRO K/A Category Points Point Tier GroupKK K K K

K K

A A

A AG Total 1

2 3

4 5

6 1

2 3

4 Emergency & 1 1

Abnormal 2

2 0

3 3

1 5

14 Plant Evolutions Tier 2

1 3

4 3

7 20 Totals

2.

1 0

0 1

0 0

1 1

0 1

0 1

5 Plant Systems 2

1 1

0 3

1 0

0 2

0 0

0 8

3 1

0 0

0 2

0 1

0 0

0 4

Totals 2

1 1

3 1

3 1

3 1

0 1

17

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 5

5 activities and general employee responsibilities l

l 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 +/-5 percent 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 System # IName 295003 IPartial or Complete Loss of A.C. Power 295003 IPartial or Complete Loss of A.C. Power 295006 SCRAM 295007 High Reactor Pressure 295009 lLow Reactor Water Level 295010 High Drywell Pressure 295013 High Suppression Pool Temperature 295014 Inadvertent Reactivity Addition 295014 Inadvertent Reactivity Addition BWR SRO Exami.ion Outline Emergency and Abnormal Evolutions - Tier 1/Group 1 Kl.K2 K3 Al A2IG KA Topic(s)

X 1AA1.01 A.C. electrical distribution system

  • X24.11 Knowledge of abnormal condition procedures.

(

ES-401 -1

Imp. i Pts.

3.8 1 1 3.6 1

I i

r i-i 1 -__

I i_

- tAK2.05 Neutron monitoring system IAA2.01 Reactor power 41 IX __

I A-1 A X

295015 Incomplete SCRAM 295016 Control Room Abandonment 295017 High Off-Site Release Rate 295023 Refueling Accidents 295023 ~Refueling Accidents 1q.4 III l-i i-i X

I

.- I AA2.04 Occurrence of fuel handling accident X 2.4.35 Knowledge of local auxiliary operator tasks during emergency operations including isystem geography and system implications.

4.1 3.5 1

1 295024 High Drywell Pressure 295025 High Reactor Pressure 295026 Suppression Pool High Water Temperature 295027 High Containment Temperature (Mark III Containment Only) 295030 !Low Suppression Pool Water Level I

FIitii I-iII IIII i

i Tuesday, March 04, 2003 2:25:39 PM Page 1

JES-401 BWR SRO Examiution Outline Emergency and Abnormal Evolutions - Tier 1/Group 1 K2 K3 Al 'A2 G KA Topic(s)

(

~System #

295031 295037 Name Reactor Low Water Level SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Kl ES-401 -1 Imp. Pts. I 1

I Ii I

295038 High Off-Site Release Rate 500000 IHigh Containment Hydrogen Concentration I

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Tuesday, March 04, 2003 2:25:39 PM Page 2

ES-401

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System # IName 295001 Partial or Complete Loss of Forced Core Flow 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 IlIl Containment Only) 295012 High Drywell Temperature 295018 Partial or Complete Loss of Component Cooling Water 295018 Partial or Complete Loss of Component Cooling Water 295019 lPartial or Complete Loss of Instrument Air 295020 Inadvertent Containment Isolation 295021 Loss of Shutdown Cooling 295022 iLoss of CRD Pumps 295028 IHigh Drywell Temperature 295029 High Suppression Pool Water Level 295032 High Secondary Containment Area Temperature 295033 High Secondary Containment Area Radiation Levels 295033 High Secondary Containment Area Radiation Levels 295034 Secondary Containment Ventilation High Radiation BWR SRO Examidtion Outline Emergency and Abnormal Evolutions - Tier 1/Group 2 Kl K2 K3 A1 A2 G KA Topic(s)

X AA1.01 Recirculation system

(

ES-401 -1 Imp. Pts.

3.6 1

i I

i I

I i

x AK1.01 Effects on component/system operations LAA2.01 Component temperatures x

3.6 3.4 i e I 3.7 3.7 I

1 I ; X I

i i

I j

I i

IAA1.04 Alternate heat removal methods X AK1.02 Reactivity control

.!i_

1 1

--Ix ix x

iEK3.04 Personnel evacuation 1EA1.01 Area radiation monitoring system EK3.01 Isolating secondary containment ventilation 4.4 1

4.0 1

4.1 1

I Tuesday, March 04, 2003 2:25:40 PM Page 3

ES-401 System # Name 295035 Secondary Containment High Differential Pressure 295035 Secondary Containment High Differential Pressure 295035 Secondary Containment High Differential Pressure 295035 !Secondary Containment High Differential Pressure 295035 Secondary Containment High Differential Pressure 295036 Secondary Containment High Sump/Area Water Level 600000 Plant Fire On Site BWR SRO Exari.adon Outline ES-401-1 Emergency and Abnormal Evolutions - Tier 1/Group 2 Kl K2 K3 Al A2 G KA Topic(s)

Imp. Pts.

X 2.4.30 Knowledge of which events related to system operations/status should be reported 3.6 1

to outside agencies.

X 2.2.20 Knowledge of the process for managing troubleshooting activities.

3.3 1

X 2.1.4 Knowledge of shift staffing requirements.

3.4 1

X 2.1.10 Knowledge of conditions and limitations in the facility license, 3.9 1

X 2.1.1 Knowledge of conduct of operations requirements.

3.8 11 X

EK3.02 Steps called our in the site fire protection plant, fire protection system manual, and 2.8 1

fire zone manual Tuesday, March 04, 2003 2:25:40 PM Page 4

ES-401 (I

System Name 201005 Rod Control and Information System (RCIS) 202002 Recirculation Flow Control System 203000 RHR/LPCI: Injection Mode (Plant Specific) 206000 High Pressure Coolant Injection System 207000 Isolation (Emergency) Condenser 209001 Low Pressure Core Spray System 209002 IHigh Pressure Core Spray System (HPCS) 211000 Standby Liquid Control System 212000 Reactor Protection System 215004 Source Range Monitor (SRM) System BWR SRO Exam(

in Outline Plant Systems - Tier 2/Group 1 K1 K2 K3 K4 K51K61A1 A2 A3 A4 G 'KA Topic(s)

ES-401 -1 Imp. Pts.]

I x -I K3.01 Reactor water level 4.4 1

i I

X 2.2.32 Knowledge of RO duties in the control room during fuel handling such as 3 3.3 1

L alarms from fuel handling area, communication with fuel storage facility,

-systems operated from the control room in support of fueling operations, and supporting instrumentation.

A1.03 Control rod block status 3.6 1

i I

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 Tuesday, March 04, 2003 2:26:03 PM I

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I-Page 1

ES-401

zn Outline Plant Systems - Tier 2/Group 1 K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic(s)

ES-401-1 Imp. Pts i 259002 Reactor Water Level Control System 261000 Standby Gas Treatment System 262001 A.C. Electrical Distribution 264000 Emergency Generators (Diesel/Jet) 290001 Secondary Containment i,

I I

i x I i i

I:

I I

I X I A3.01 System flow i

IK6.09 D.C. power 3.3 1

3.5 1 I Tuesday, March 04, 2003 2:26:04 PM Page 2

ES-401

(

BWR SRO Exam(

Dn Outline Plant Systems - Tier 2/Group 2 System Name KlK21K3 K4 K5 K6 A1A2 A3 A4 G KATopic(s) 201001 jControl Rod Drive Hydraulic System 201002 lReactor Manual Control System 201004 Rod Sequence Control System (Plant Specific) 201006 Rod Worth Minimizer System (RWM)

(Plant Specific)

K 2

c c

202001 Recirculation System X

K4.01 2/3 core coverage: Plant 204000 Reactor Water Cleanup System l

205000 Shutdown Cooling System (RHR i

X i

K4.02 High pressure isolation:

Shutdown Cooling Mode) 205000 Shutdown Cooling System (RHR X

A2.05 System isolation

'Shutdown Cooling Mode) 214000 lRod Position Information System 215002 Rod Block Monitor System I

I 215003 Intermediate Range Monitor (IRM) System X

'K2.01 IRM channels/detectors 219000 lRHR/LPCI: Torus/Suppression Pool

!Cooling Mode 230000 RHR/LPCI: Torus/Suppression Pool Spray Mode 234000 Fuel Handling Equipment X

K5.02 Fuel handling equipmenl 234000 lFuel Handling Equipment X

K1.05 Reactor vessel compone 239003 MSIV Leakage Control System 245000 Main Turbine Generator and Auxiliary Systems 259001 Reactor Feedwater System 262002 Uninterruptable Power Supply (A.C./D.C.)

263000 D.C. Electrical Distribution Tuesday, March 04, 2003 2:26:04 PM Page 3

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ES-4011.

Imp.

Pts. l

-Specific 3.9 1

I Plant-Specific 3.8 1

3.7 1

2.7 1

I interlocks ents: Plant-Specific 3.7 1

3.7 1

i i 3.3 1

ES-401 System iName 271000 Offgas System 272000 Radiation Monitoring System 286000 Fire Protection System 290003 !Control Room HVAC 300000 Instrument Air System (lAS) 400000 Component Cooling Water Syste (CCWS)

BWR SRO ExarT{

on Outline Plant Systems - Tier 2/Group 2 K1 K2 K3 K4 K51K6 A1 A2 A3 A4 G KA Topic(s)

X A2.01 Fuel e t

m K4.01 Autorr L1 1

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ES-401 -1 Imp.

Pts.-I I

lement failure 4.1 i 1 1 i

natic start of standby pump i.

3.9 1

Tuesday, March 04, 2003 2:26:04 PM Page 4

ES-401 BWR SRO Exam/

on Outline Plant Systems - Tier 2/Group 3 System Name Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic(s) 201003 Control Rod and Drive Mechanism X

K1.01 Control rod drive hydraulic system 215001 Traversing In-Core Probe X

K6.04 Primary containment isolation system: Mark-I&lI(Not-BWR1) 233000 Fuel Pool Cooling and Clean-up X

A2.02 Low pool level 239001 Main and Reheat Steam System 256000 Reactor Condensate System 268000 Radwaste 288000 Plant Ventilation Systems 290002 Reactor Vessel Internals X

K6.05 SBLC

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ES-401 1 [

Imp.

Pts.'^

-.- q 3.3 1

3.4 1

3.3 1

3 I

3.4 1

Tuesday, March 04, 2003 2:26:04 PM Page 5

I ES-701 BWR LSRO NRC Examination Outline Facility:

Date of Exam:

Exam Level:

Hope Creek 3/10/03 LSRO Category K/A#

Topic Imp.

Points 292001 KI.02 Reactor Theory - Neutrons - Define prompt and delayed 3.1 1

neutrons.

292002 K1.08 Reactor Theory - Neutron Life Cycle - Define effective 2.8 1

multiplication factor and discuss its relationship to the state of d reactor.

292003 K1.07 Reactor Theory - Reactor Kinetics and Neutron Sources -

3.3 1

3. Reactor and fuel Explain prompt critical, prompt jump, and prompt drop.

l characteristics and 292004 K1.05 Reactor Theory - Reactivity Coefficients - Define the Doppler 2.9 1

physical aspects of coefficient of reactivity.

core construction 292008 K1.30 Reactor Theory - Reactor Operational Physics - Explain the 3.5 1

important to fuel relationship between decay heat generation and: a) power level handling or history, b) power production, and c) time since reactor shutdown activities shutdown.

293006 K1.13 Thermodynamics Theory - Fluid Statics - Explain the results of 2.7 1

putting centrifugal pumps in parallel or series combinations 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

G 2.3.4 Knowledge of radiation exposure limits and contamination 3.1 1

4. Health Physics and control, including permissible levels in excess of those Radiation Protection authorized.

for fuel handling G 2.3.5 Knowledge of use and function of personnel monitoring 2.5 1

activities and equipment.

general employee G 2.3.10 Ability to perform procedures to reduce excessive levels of 3.3 1

responsibilities radiation and guard against personnel exposure.

Total I

_I__

5

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ot ES-401 Record of Rejected K/As Form ES-401 -i0 Tier/Group Randomly Selected K/A Reason for Rejection Reduce oversampling in RHR Shutdown Cooling following NRC Outline review.

1/2 295021 K3.02 Replaced 295021K3.02 with 261000A3.01 SBGTS/FRVS system flow. Randomly selected from 261000 system KA's.

Reduce oversampling in RHR Shutdown Cooling following NRC Outline review.

1/2 295021 G 2.4.35 Replaced KA 29502fG2.4.35 with KA Generic 2.3.5 by random selection of remaining Health Physics KA s.

Reduce oversampling in CRD following NRC Outline review.

1/2 295022 K3.02 Replaced KK 295027 K3.02 with KA 203000 K3.01 RHR LPCI Reactor water level. Randomly selected from 203000 System KA's.

Replaced KA 201001 K5.03 with KA 215005A1.03 APRM / LPRM Control Rod Block status due to 2/2 201001 K5.03 oversampling in CRD following NRC Outline review. Randomly selected from the 215005 system KA's.

3 2Replaced Generic 2.1.10 with KA 295035 G 2.1.10 to move from Category 4 to Procedures section 3G 2.1.10 due to NRC request. IGNORE 295035 K/A Title due to software limitations.

3 G 2.4.30 Replaced Generic KA G 2.4.30 with KA 295035 G 2.4.30 when moved from Category 4 tol Procedures section due to NRC request. IGNORE 295035 K/A Title due to software limitations.

3 2Replaced Generic KA G 2.1.4 with KA 295035 G 2.1.4 when moved from Category 4 to Procedures 3G 2.1.4 section due to NRC request. IGNORE 295035 K/A Title due to software limitations.

2/2 204000 K5.04 Replaced KA from 204000 K5.04 with 234000 K5.02 Fuel Handling Equipment - interlocks per NRC Lead Examiner request due to oversampling. RWCU KA too close to JPM.

2/2 205000 K6.03 295001A101. Randomly selected from 205000 K categories.

1/2 600000 K3.04 Replaced KA 600000 K3.04 with 600000 K3.02 based on original KA outside LSRO responsibilities.

.KA Selected randomly from 6000001<3 category.

1/1 295023 G 2.1.20 Replaced KA 295023 G 2.1.20 with 295023 G 2.4.35 due to being too similar to 272000A201.

Randomly selected due to oversampling.

2/3 290002 K6.02 Replaced KA 290002 K6.02 with 290002 K6.05 due to oversampling of CRD. Randomly selected from 290002K6 category.

NUREG 1021, Revision 8. Supplement 1

(

ES-401 Record of Rejected K/As Form ES-401-,iu 1/2 295001 A2.04 Replaced 295001 A2.04 with 295035 G 2.2.20 due to oversampling of SDC. Ignore 295035 K/A Title due to software limitations.

Replaced KA 295034 A1.02 with Generic KA 295035 G 2.1.1 due to oversampling of High radiation in 1/2 295034 A1.02 the secondary containment. Randomly selected from Generic topics. Ignore 295035 K/A Title due to software limitations.

295003 G 2.4.32 replaced with 295003 G 2.4.11. Could not make valid question from KA topic that 1/1 295003 G2.4.32 was not either a direct lookup, or outside scope of LSRO. Replaced KA with KA not already used from same Generic category. Change does not impact Outline KA distribution.

GFE 292005 K1.01 Replaced 292005 K1.01 with 292002 K1.08 due to low disriminatory value IAW Lead Examiner following NRC review. KA selected from remaining unused RX Theory KAs.

NUREG 1021, Revision 8. Supplement 1

ES-301 Administrative Topics Outline Form-ES-301-1 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 2.1.18 (3.0) - Ability to make accurate / clear and concise logs / records / status boards Operations

/ and reports.

JPM JPM: Verify HC.OP-DL.ZZ-0026 log requirements for resuming Core Alterations A.2 Equipment 2.2.26 (3.7) - Knowledge of refueling administrative requirements Control JPM JPM: Verification of Minor changes to the Fuel Movement Sheet IAW HC.RE-FR.ZZ-0001 (Q) Attachment 4 2.3. 5 (2.5) - Knowledge of use of personnel monitoring equipment QUESTION: Personnel contamination response Radiation Control A.3 Questions 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 Emergency Plan Questions 2.4.41 (4.1) - Knowledge of emergency action level thresholds and classifications QUESTION: EP Event classification for fuel handling event I ______________________________________________

03HCLSRONRCADMOUTL.do C

Revised-01/16/03,'6:07 PM 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 SytmJPV ecrpinType Safety Syste m JPM Description Code*

Function S.1 215004 G2.1.23 (4.0) Ability to perform specific system and integrated L

IC Source Range plant procedures during different modes of plant Monitor operation.

SRM/IRM Rod Block Bypassing during refueling operations IAW HC.OP-SO.SE-0001 Section 5.4.

_Perform independent verification of installed jumpers.

S.2 204000 G2.1.20 (4.2) Ability to execute procedure steps.

N, R, E, AUX/

RWCU L, A DHR Align RWCU for Alternate Heat Removal Alternate path for bypassing RHX for additional cooling.

S.3 234000 G2.2.28 (3.5) Knowledge of new and spent fuel movement N, R, A FHE Fuel Handling procedures.

Systems Manual transfer of dummy bundle within Spent Storage Pool. Unexpected Slack Cable I Bent Mast IAW HC.OP-SO.KE-0001 Attachment 2 (perform or simulate)

(JPM-KE-014 Modified for Alternate path due to unexpected Slack Cable.)

S.4 234000 A3.02 (3.7)

Interlock operation N, R FHE Fuel Handling Systems Perform Monorail Aux Hoist Controls Functional Test HC.OP-FT.KE-0001 Section 5.4.1 through 5.4.15 (perform or simulate) l S.5 234000 A3.01 (3.6)

Crane/refuel bridge movement.

N, R FHE Fuel Handling Systems Semi-Automatic dummy bundle transfer in the Spent Fuel Pool (perform actual movement).

B.2: Facility Walk-Through (Same as RO In-Plant Walkthrough)

NANA NA 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 GILS RO/03 H C LS R ON RCJ PM OUTL. doc 03/07/03, 5:21 PM NUREG-1021, Revision 8

Appe~ndix D Scenario Outline Form ES-D-1 In nI...cea.oO.ln.Fr E--

Facility: Hope Creek Scenario No.:1 Op Test No.: 1 Examiners:

Candidates:

LSRO LS RO LS RO Obiectives: Evaluate applicants' response to an SRM failure. Applicant determines requirements for CRB removal are not met. 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. The Reactor Mode Switch is Operable and locked in Refuel position. All Control Rods are inserted except rod 30-31 for friction testing.

The CRDM for 14-23 has been replaced after rebuild. All SRMs are operable. Shutdown Margin requirements are met. The Dominion Engineering Inc. (DEI) FSP tool with grid guide is attached to the Frame Mounted Aux Hoist. The Control Rod Grapple is on the Monorail Hoist.

Turnover: You are the Refueling SRO. All fuel is in the vessel except the 4 bundles of the 14-23 cell.

Control Rod Blade 14-23 needs to be removed and replaced. The Control Rod Blade 14-23 is fully withdrawn with the CRDM uncoupled from under-vessel. The double blade guide was just removed from cell 14-23 and is on the Main Hoist. You are at Step 5.3 of HC.RE-FR.ZZ-0002.

Event l Malf.

l Event Event Description Evaluator Guide No.

i No.

Type 1

1 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 Determines rod does not meet Blade.

Tech Spec 3.9.10.1 requirements for a single control rod removal. Rod 30-30 must be fully inserted.

Discusses Restricted Core Operations Form (RCOF) to continue.

Remove fuel support piece.

Uses CRB Grapple on Monorail Hoist to remove CRB.

HC LSRO ES 1 Rev. 2/14/03 Page 1 of 2

3 NA M

Under vessel crew reports water pouring out 14-23 CRDM flange.

They are unable to stop the leak.

Recognizes cavity level would be lowering and takes actions of HC.OP-AB.COOL-0004 Fuel 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. 2/14/03 Page 2 of 2

Scepnario Outline Form ES-D-1 A

tin--eI, L

t AypinlulA U

Facility: Hope Creek Scenario No.: 2 Op Test No.: I Examiners:_

Candidates:

LSRO LSRO LSRO LS RO Obmectives: To evaluate the applicants' ability to recognize and address problems with control rod support. 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. A move sheet and core map is provided.

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

Event Malf.

Event Event Evaluator No No.

Type*

Description Guide N/A N

Double blade guide removed lAW Double Blade Guide is grappled to HC.OP-SO.KE-0001 be removed from the core location according to procedure.

2 1

N Provide adequate support for Determines inadequate support Control Rod 30-27.

and initiates corrective action.

3 N/A C

A fuel bundle fails in a location Notify Control Room.

causing high radiation conditions on the Refueling Floor and in the Implement actions of HC.OP-Drywell.

AB.CONT-0005 IRRADIATED FUEL DAMAGE.

Suspends all refueling operations.

Recognizes radiological effects on the Drywell.

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.

5 N/A M

Classify the event.

Classifies the event as an ALERT lAW ECG 6.4.2.a

' (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

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