ML030420295

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Init Exam - 8/2002- Draft Outlines
ML030420295
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 02/23/2002
From: Gody A
NRC Region 4
To: Eaton W
Entergy Operations
References
50-416/02-301 50-416/02-301
Download: ML030420295 (41)
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Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 8/26/2002 - 8/30/2002 Examination Level (circle one): RO / SRO Operating Test Number: __1___

Administrative Describe method of evaluation: Knowledge IMP Additional ORIGIN NOTES Topic/Subject 1. ONE Administrative JPM, OR / Ability K/As Description 2. TWO Administrative Questions A.1 Plant Drawings JPM GJPM-OP-ADM21 2.1.25 2.8 BANK NRC Given a component, determine the fuse(s) 5/2000 to be removed to de-energize the component and how the component will fail.

Operator JPM GJPM-OP-ADM42 2.1.2 3.0 NEW Responsibilities (Condition Given a plant discrepancy, initiate a Reports) condition report.

A.2 Protective JPM GJPM-OP-ADM41 2.2.13 3.6 2.1.2: 3.0 MOD NRC 6/01 Tagging HPCS Given an installed equipment clearance, changed perform the duties of an independent to LPCS verifier.

A.3 Radiation JPM GJPM-OP-ADM34 2.3.1 2.6 MOD NRC 6/01 Control Changed was entry Perform required actions to access the to and exit Controlled Access Area (CAA), determine Contam. with requirements to enter a Contamination Area entry Area, and exit the CAA. rqmts to high contam area CFR 55.45 (a)9 & 10 A.4 Emergency Plan JPM GJPM-OP-ADM25 2.4.43 2.8 2.4.39: 3.3 BANK CFR 55.45 Assessment NRC (a) 11 2.4.30: 3.6 Given initial notification form, perform 5/2000 notification of offsite agencies using the Operational Hotline.

REVISION 0 4/29/2002

ES-301 Individual Walk-Through Test Outline Form ES-301-2 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 8/26/2002 - 8/30/2002 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___1___

System / JPM Title / Type Codes* Safety Knowledge IMP. Additional ORIGIN NOTES Function / Ability K/As B.1. CONTROL ROOM SYSTEMS

1. 241000 REACTOR / TURBINE PRESSURE 3 A4.06 3.9 A1.07: 3.8 NEW REGULATING SYSTEM (N)(S)(L)(A) A2.03: 4.1 Operate the Turbine Pressure Control System A3.08: 3.8 CFR to Lower Reactor Pressure (Pressure Regulator 55.45(a)3, fault requiring use of Manual Jack) 4, 5 & 8 GJPM-RO-N3202
2. 204000 REACTOR WATER CLEANUP SYSTEM 2 A1.01 3.1 A1.04: 2.8 BANK Failure of (RWCU) (D)(S)(L)(A) A1.06: 2.8 G3-F234 Align RWCU for Vessel Level Control (Faulted A2.01: 3.2 NRC CFR

- Valve Failure) 5/2000 55.45(a)3

& 8 GJPM-RO-G3301

3. 295029 SUPPRESSION POOL LEVEL HIGH 5 EA1.02 3.1 209002 BANK CFR (D)(S) A3.01: 3.3 55.45(a)8 Lower Suppression Pool Level with HPCS A4.01: 3.7 NRC A4.02: 3.6 3/1998 GJPM-RO-E2209 A4.14: 3.0
4. 261000 STANDBY GAS TREATMENT SYSTEM 9 A4.03 3.0 A4.02: 3.1 BANK CFR (D)(S)(A) A4.09: 2.7 55.45(a)9 Place SBGT Train in Standby with an Auto NRC Start Signal Present (Faulted - High Rad) 6/2001 GJPM-RO-T4801
5. 295021 LOSS OF SHUTDOWN COOLING (ADHR) 4 AA1.04 3.7 205000 MOD CFR (M)(S)(A) A4.01: 3.7 55.45(a)5 Startup Alternate Decay Heat Removal in A4.02: 3.6 Added & 8 Reactor to Reactor Mode (E12-F042C fail on A4.03: 3.6 Valve NRC 5/2000 stroke) A4.09: 3.1 failure A2.10: 2.9 GJPM-RO-E1231 REVISION 0 4/29/2002

ES-301 Individual Walk-Through Test Outline Form ES-301-2 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 8/26/2002 - 8/30/2002 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___1___

System / JPM Title / Type Codes* Safety Knowledge IMP. Additional ORIGIN NOTES Function / Ability K/As

6. 262001 AC ELECTRICAL DISTRIBUTION 6 A4.01 3.4 A4.02: 3.4 NEW CFR (N)(S) A4.04: 3.6 55.45(a) 8 Transfer Electrical Loads from Service A4.05: 3.3 Transformer 21 to Service Transformer 11 2.1.31: 4.2 2.1.30: 3.9 GJPM-RO-R2730
7. 201005 ROD CONTROL AND INFORMATION SYSTEM 7 A2.03 3.2 A2.06: 3.2 BANK CFR (RCIS) (D)(C) 295037 55.45(a) 3 Defeat Rod Control and Information System per EA1.08: 3.6 NRC & 6 EP2 Attachment 20. 295015 5/2000 Also Safety AA1.04: 3.4 Function 1 GJPM-RO-EP030 B.2. FACILITY WALK-THROUGH
8. 295016 CONTROL ROOM ABANDONMENT 7 AA1.07 4.2 BANK CFR (D)(P)(R) 55.45(a) 8 Align the Remote Shutdown Panel Alternate NRC 9, & 12 Shutdown Panels for a Fire in the Control Room 6/2001 Emergency/

Abnormal GJPM-RO-C6108

9. 201001 CONTROL ROD DRIVE (CRD) HYDRAULIC 1 2.1.30 3.9 2.1.29: 3.4 BANK CFR SYSTEM (D)(P)(R) 55.45(a) 8 Rotate the CRD Drive Water Filters & 9 GJPM-NLO-C1102
10. 286000 FIRE PROTECTION SYSTEM 8 A4.06 3.4 2.1.20: 4.3 BANK CFR (D)(P)(A) 2.1.30: 3.9 55.45(a) 6 Perform Local Start of Fire Pump Diesel after NRC Failure to Start (Faulted) 5/2000 GJPM-NLO-P6401
  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (P)lant, (R)CA REVISION 0 4/29/2002

ES-301 Administrative Topics Outline Form ES-301-1 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 8/26/2002 - 8/30/2002 Examination Level (circle one): RO / SRO Operating Test Number: __1___

Administrative Describe method of evaluation: Knowledge IMP Additional ORIGIN NOTES Topic/Subject 1. ONE Administrative JPM, OR / Ability K/As Description 2. TWO Administrative Questions A.1 Technical JPM GJPM-SRO-ADM43 2.1.12 4.0 2.2.23: 3.8 BANK Different Specifications component Given a component, determine Limiting Condition for Operations and complete LCO documentation.

Plant Safety JPM GJPM-SRO-ADM46 2.1.19 3.0 BANK Different Index (EOOS) NRC component Given plant conditions, determine the 6/2001 Plant Safety Index (EOOS).

A.2 Protective JPM GJPM-SRO-ADM45 2.2.13 3.8 2.1.2: 4.0 NEW Tagging Given a clearance, perform the supervisory review of the clearance for adequacy and issue.

A.3 Radiation JPM GJPM-SRO-ADM44 2.3.1 3.0 2.3.4: 3.1 MOD Changed Control to Perform required actions to access the Planned Controlled Access Area (CAA), determine Special requirements to enter a High Radiation Exposures Area in an Emergency and authorization CFR 55.45 required, and exit the CAA. (a)9 & 10 A.4 Emergency Plan JPM GJPM-SRO-A&E40 2.4.41 4.1 2.4.30: 3.6 NEW CFR 55.45 Action Levels (a) 11 Given conditions, determine the appropriate emergency classification and complete the required notification form.

REVISION 0 4/29/2002

ES-301 Individual Walk-Through Test Outline Form ES-301-2 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 8/26/2002 - 8/30/2002 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___1___

System / JPM Title / Type Codes* Safety Knowledge IMP. Additional ORIGIN NOTES Function / Ability K/As B.1. CONTROL ROOM SYSTEMS

1. 295021 LOSS OF SHUTDOWN COOLING (ADHR) 4 AA1.04 3.7 205000 MOD CFR (M)(S)(A) A4.01: 3.7 55.45(a)5 Startup Alternate Decay Heat Removal in A4.02: 3.6 Added & 8 Reactor to Reactor Mode (E12-F042C fail on A4.03: 3.6 Valve NRC 5/2000 stroke) A4.09: 3.1 failure A2.10: 2.9 GJPM-RO-E1231
2. 262001 AC ELECTRICAL DISTRIBUTION 6 A4.01 3.7 A4.02: 3.4 NEW CFR 55.45(a)

(N)(S) A4.04: 3.7 8 Transfer Electrical Loads from Service A4.05: 3.3 Transformer 21 to Service Transformer 11 2.1.31: 3.9 2.1.30: 3.4 GJPM-RO-R2730 B.2. FACILITY WALK-THROUGH

3. 295016 CONTROL ROOM ABANDONMENT 7 AA1.07 4.3 BANK CFR 55.45(a)

(D)(P)(R) 8 Align the Remote Shutdown Panel Alternate NRC 9, & 12 Shutdown Panels for a Fire in the Control 6/2001 Emergency/

Room Abnormal GJPM-RO-C6108

4. 201001 CONTROL ROD DRIVE (CRD) 1 2.1.30 3.4 2.1.29: 3.3 BANK CFR 55.45(a)

HYDRAULIC SYSTEM (D)(P)(R) 8 Rotate the CRD Drive Water Filters & 9 GJPM-NLO-C1102

5. 286000 FIRE PROTECTION SYSTEM 8 A4.06 3.4 2.1.20: 4.2 BANK CFR 55.45(a)

(D)(P)(A) 2.1.30: 3.4 6 Perform Local Start of Fire Pump Diesel after NRC Failure to Start (Faulted) 5/2000 GJPM-NLO-P6401

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (P)lant, (R)CA REVISION 0 4/29/2002

ES-401 FORM ES-401-2 BWR RO EXAMINATION OUTLINE Facility: GRAND GULF NUCLEAR STATION Date of Exam: 23 AUGUST 2002 K/A CATEGORY POINTS TIER GROUP K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G POINT

  • TOTAL
1. 1 2 3 3 2 2 1 13 Emergency &

Abnormal 2 3 4 3 4 4 1 19 Plant Evolutions 3 0 0 1 1 2 0 4 TIER 5 7 7 7 8 2 36 TOTAL 1 5 2 2 3 3 4 1 3 2 1 2 28 2.

Plant 2 1 1 2 2 1 1 2 4 4 1 0 19 Systems 3 1 0 0 0 0 0 1 2 0 0 0 4 TIER 7 3 4 5 4 5 4 9 6 2 2 51 TOTAL CAT 1 CAT 2 CAT 3 CAT 4

3. Generic Knowledge & Abilities 4 2 2 5 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 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.

REVISION 0 4/29/2002 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-2 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 1 2 # /BOTH K/A 295005 Main Turbine Generator Trip / 3 05 Given a spurious Main Turbine Generator trip, determine 3.8 553 BOTH AA2.04: 3.7 BANK CFR41.5/41.6 the initial affect on Reactor Power and Reactor Pressure. q 001 AA2.03: 3.1 NRC 6/01 295006 SCRAM / 1 02 Describe the position of control rods following a reactor 4.3 10 BOTH 201005 BANK Rx SCRAM CFR41.6/41.10/43.5 scram and how position is determined. q002 A3.02: 3.5 NRC 6/01 Immediate A4.02: 3.7 Actions 201003 K4.05: 3.2 A3.01: 3.7 A4.02: 3.5 295007 High Reactor Pressure / 3 01 Describe the response of the Turbine Pressure Control 3.5 69 BOTH 241000 BANK CFR41.5 System on an increasing reactor pressure. q003 K4.01: 3.8 NRC 6/01 A2.02: 3.7 295009 Low Reactor Water Level / 2 03 Given plant conditions, identify the status of 3.1 121 BOTH AK1.02: 3.0 BANK CFR41.5/43.5 Recirculation pumps originally operating in fast speed q004 AA1.03: 3.0 NRC 3/98 with a lowering water level.

295010 High Drywell Pressure / 5 04 Given plant parameters and elevated in-leakage into the 3.5 286 BOTH AK3.05: 3.5 BANK GGNS Drywell CFR41.5 Drywell, determine the status of reactor coolant system q005 2.4.21: 3.7 NRC 4/00 air leak integrity.

295014 Inadvertent Reactivity Addition / 1 01 With the reactor in startup conditions such that the reactor 4.1 204 BOTH BANK Susquehanna CFR41.1/41.2/41.6/43.6 has dropped subcritical, what are the operator actions if a q006 NRC 6/01 reactivity event high worth control rod is withdrawn fully. 7/98 295015 Incomplete SCRAM / 1 01 Describe the method to be used to allow the insertion of 3.4 203 BOTH BANK CFR41.6/43.6 control rods using RCIS during an ATWS. q007 NRC 6/01 295024 High Drywell Pressure / 5 14 Given conditions, determine ability to initiate 3.9 601 BOTH NEW High Drywell CFR41.9/41.10/43.5 Containment Spray per EOPs. q008 pressure is EOP entry for CTMT Press Control 295025 High Reactor Pressure / 3 05 State the Reactor Vessel pressure Safety Limit and its 4.4 30 RO EK1.02: 4.1 BANK CFR41.3/43.2 basis. q076 2.2.22: 3.4 NRC 2.2.25: 2.5 12/00 295031 Reactor Low Water Level / 2 01 Given plant conditions and a low reactor water level, 4.6 309 BOTH 2.1.1: 3.7 BANK CFR41.2/41.3/41.10/43.5 determine core cooling mechanism and adequacy. q009 2.4.21: 3.7 NRC 12/00 PAGE 1 TOTAL TIER 1 GROUP 1 2 3 3 0 2 0 PAGE TOTAL # QUESTIONS 10 REVISION 0 4/29/2002 PAGE 1 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE CONT. ES-401-2 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295037 SCRAM Condition Present and Reactor 04 Determine when plant conditions allow the termination of 4.5 216 BOTH EK1.04: 3.4 BANK Power Above APRM Downscale or Unknown / 1 Standby Liquid Control with multiple control rods stuck q010 EK1.05: 3.4 NRC 4/00 CFR41.1/41.2/41.6/43.6 out and what is the basis. EA2.03: 4.3 500000 High Containment Hydrogen Conc. / 5 2. Given applicable SOP, graph, and plant conditions, 4.3 219 BOTH EA1.03: 3.4 BANK moved to CFR41.10/43.5 1. determine the power settings for the Hydrogen q011 2.1.25: 2.8 NRC 4/00 Generics 20 Recombiners and time to full power.

295014 Inadvertent Reactivity Addition / 1 02 Determine effects on the reactor of the fast opening a 3.6 116 BOTH 202002 BANK CFR41.6/43.6 Recirc Flow Control Valve. q012 K1.02: 4.2 NRC 3/98 K3.02: 4.0 PAGE 2 TOTAL TIER 1 GROUP 1 0 0 0 2 0 1 PAGE TOTAL # QUESTIONS 3 PAGE 1 TOTAL TIER 1 GROUP 1 2 3 3 0 2 0 PAGE TOTAL # QUESTIONS 10 K/A CATEGORY TOTALS: 2 3 3 2 2 1 TIER 1 GROUP 1 GROUP POINT TOTAL 13 500000 random selection was EA. Topic covers both EA 1.03 operation of CTMT Atmosphere control system and Generic procedure usage; moved to Generics due to higher importance.

REVISION 0 4/29/2002 PAGE 2 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-2 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295001 Partial or Complete Loss of Forced Core 2. Given plant conditions and the power to flow map, 3.4 602 BOTH 2.4.1: 4.3 MOD moved to Flow Circulation / 1 & 4 4. determine the actions to be taken. q013 AA2.01: 3.5 NRC 4/00 Generics CFR41.5/41.10/43.5 11 295002 Loss of Main Condenser Vacuum / 3 04 Describe the basis for the isolation of the Main Steam 3.3 220 BOTH AK3.05: 3.4 BANK CFR41.4/43.4 Isolation Valves on a loss of condenser vacuum. q015 NRC 4/00 295003 Partial or Complete Loss of AC Power/ 6 03 Given a lockout on BOP Transformer 12B, determine the 3.7 507 BOTH AA1.01: 3.7 BANK CFR41.7 configuration of the AC Distribution System. q014 NRC 6/01 295004 Partial or Complete Loss of DC Power / 6 295008 High Reactor Water Level / 2 01 Identify the affects of a high Reactor Water Level on the 3.0 275 BOTH 245000 BANK CFR41.4/41.5 Main Turbine and Reactor Feed Pump Turbines. q016 A3.01: 3.6 NRC 6/01 259001 K6.07: 3.8 295011 High Containment Temperature / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Water Temp. / 5 01 During a surveillance operating RCIC, determine how 3.8 BOTH AA1.02: 3.9 NEW CFR41.5/41.10/43.2/43.5 often Suppression Pool Temperature is required to be q018 2.1.33: 3.4 monitored and the threshold for alternate actions.

295016 Control Room Abandonment / 7 02 Given parameters from the Remote Shutdown Panel 4.2 603 BOTH 2.1.25: 2.8 MOD CFR41.5/41.10/43.5 indications, determine actual and Narrow Range RPV q017 2.4.11: 3.4 NRC 6/01 level.

295017 High Offsite Release Rate / 9 08 Given plant conditions with MSIV Leakage Control 3.1 BOTH AA1.09: 3.6 NEW CFR41.11/41.13/43.4 System operating, determine the mechanism for q019 monitoring radiological release to locations outside the Auxiliary Building.

PAGE 1 TOTAL TIER 1 GROUP 2 1 1 0 2 2 1 PAGE TOTAL # QUESTIONS 7 295001 random selection was AA2. Topic moved to Generics that are in addition to the random selection.

REVISION 0 4/29/2002 PAGE 3 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE CONT. ES-401-2 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295018 Partial or Complete Loss of CCW / 8 03 Given a partial loss of Component Cooling Water, 3.1 314 BOTH BANK CFR41.4/41.10/43.5 determine actions to be taken and their basis. q020 NRC 12/00 295019 Partial or Complete Loss of Inst. Air / 8 14 Given a reduction in Instrument Air Header pressure, 3.2 BOTH NEW New Air CFR41.4/41.10/43.5 determine a possible cause. q021 Dryers moved to AK2 295020 Inadvertent Cont. Isolation / 5 & 7 03 Given a loss of Instrument Air System pressure to the 3.1 BOTH AK3.03: 3.2 NEW CFR41.4/41.7/41.9 Auxiliary Building, determine the effects on the ability to q022 remove heat from the Containment.

295022 Loss of CRD Pumps / 1 04 Describe the effects on reactor water level during a 2.5 55 BOTH AK2.04: 2.5 BANK CFR41.5 reactor startup with minimal decay heat and a loss of q023 NRC 6/01 CRD Pumps. (RWCU is lined up to blowdown to the main condenser to compensate for CRD flow.)

295026 Suppression Pool High Water Temp. / 5 01 Given plant conditions, determine when Suppression Pool 3.9 301 BOTH BANK CFR41.7/41.9/41.10/41.14/43.5 Cooling is no longer effective and alternate actions are q024 NRC 12/00 required.

295027 High Containment Temperature / 5 03 Given plant conditions, determine the Technical 3.7 512 BOTH 2.2.25: 2.5 BANK CFR41.9/41.10/43.2 Specification Bases for shutting down the Reactor due to q025 NRC 6/01 a high Containment Temperature.

295028 High Drywell Temperature / 5 03 Given rising temperatures in the Reactor lower cavity 3.9 BOTH EK2.04: 3.6 NEW CFR41.4/41.7 area, determine the response of the Drywell Cooling q26 EK3.04: 3.6 system.

295029 High Suppression Pool Water Level / 5 01 Identify the bases for Emergency RPV Depressurization 3.4 513 BOTH BANK CFR41.9/41.10 when Suppression Pool Level cannot be maintained q027 NRC 6/01 below 24.4 feet.

295030 Low Suppression Pool Water Level / 5 02 Evaluate Suppression Pool Temperature, with a Low 3.9 8 BOTH BANK Caution 2 CFR41.9/41.10/43.5 Suppression Pool Level. q028 NRC 6/01 EOP-2 PAGE 2 TOTAL TIER 1 GROUP 2 1 3 2 2 1 0 PAGE TOTAL # QUESTIONS 9 295019 random selection AK1 has NONE. Moved selection to AK2 to support question concerning new Instrument Air Dryer System failures.

REVISION 0 4/29/2002 PAGE 4 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE CONT. ES-401-2 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295033 High Secondary Containment Area 03 Given operations during a Refueling outage, determine 3.7 BOTH NEW Fuel Pool HX in Radiation Levels / 9 the cause of elevated radiation levels in the area of the q029 an open area of CFR41.12/43.4 Fuel Pool Cooling Heat Exchangers. Sec CTMT 295034 Secondary Containment Ventilation High Radiation / 9 295038 High Offsite Release Rate / 9 02 Given plant conditions and procedures, determine the 4.2 BOTH BANK CFR41.10/41.12/43.4/43.5 protective action recommendations to be recommended to q030 the state and local officials in an emergency.

600000 Plant Fire On Site / 8 04 Given a fire at the Hydrogen Bulk Storage Facility 2.8 BOTH 2.4.25: 2.9 NEW moved to CFR41.10/43.5 describe the actions to be taken to combat the fire. q031 2.1.32: 3.4 AK3 PAGE 3 TOTAL TIER 1 GROUP 2 1 0 1 0 1 0 PAGE TOTAL # QUESTIONS 3 PAGE 1 TOTAL TIER 1 GROUP 2 1 1 0 2 2 1 PAGE TOTAL # QUESTIONS 7 PAGE 2 TOTAL TIER 1 GROUP 2 1 3 2 2 1 0 PAGE TOTAL # QUESTIONS 9 K/A CATEGORY TOTALS: 3 4 3 4 4 1 TIER 1 GROUP 2 GROUP POINT TOTAL 19 600000 random selection was AA1. Topic was moved to AK3 due to limited discriminatory value of AA1 for Licensed Operators. AK3 allows testing of precaution.

REVISION 0 4/29/2002 PAGE 5 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-2 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 3 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 1 2 # /BOTH K/A 295021 Loss of Shutdown Cooling / 4 01 Given plant parameters and the graphs from the 3.5 604 BOTH AK1.01: 3.6 MOD Multiple graphs CFR41.5/41.10/43.5 Inadequate Decay Heat Removal ONEP, determine time q032 NRC 3/98 for various to boil. conditions.

295023 Refueling Accidents / 8 295032 High Secondary Containment Area 03 Determine the systems affected by high temperatures in 3.8 48 BOTH EK3.07: 3.6 BANK Steam leak at Temperature / 5 the RHR A Pump Room. q033 219000 NRC 3/98 GGNS in Steam CFR41.4/41.10/43.5 A1.08: 3.7 Condensing A2.14: 4.1 piping 295035 Secondary Containment High Differential 02 Describe the operation of the Standby Gas Treatment 3.8 BOTH BANK Pressure / 5 System with regard to Auxiliary Building and Enclosure q034 CFR41.4/41.7 Building Pressures.

295036 Secondary Containment High Sump/Area 03 Given a rising water level in an ECCS Pump Room, 3.4 58 BOTH EK3.03: 3.5 BANK EOP-4 Water Level / 5 determine appropriate actions to be taken with regard to q035 EA2.02: 3.1 NRC 3/98 CFR41.4/41.10/43.5 the overall plant operation.

K/A CATEGORY TOTALS: 0 0 1 1 2 0 TIER 1 GROUP 3 GROUP POINT TOTAL 4 REVISION 0 4/29/2002 PAGE 6 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-2 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 1 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 201001 CRD Hydraulic 02 During a reactor scram with scram signals not 2.6 BOTH MOD CFR41.5/41.6 reset, determine the CRD System flow rates. q036 201005 RCIS 01 Given a failure of the Turbine First Stage 3.2 BOTH NEW CFR41.6/43.6 pressure signal to RCIS, determine the mode q037 of control rod movement.

202002 Recirculation Flow Control 01 Given plant conditions following an actuation 3.4 BOTH 202001 NEW EOC-RPT CFR41.6 of EOC-RPT, determine the Recirculation q038 A2.15: 3.7 changes Pump circuit breaker configuration. 2002 203000 RHR/LPCI: Injection Mode 17 State the basis for monitoring reactor pressure 4.0 60 BOTH K4.01: 4.2 BANK Where is CFR41.8 when aligning the RHR system for injection q039 K4.02: 3.3 NRC 6/01 pressure into the vessel for the LPCI mode. A3.01: 3.8 sensed on A3.08: 4.1 LPCI for A4.08: 4.3 operation of the injection valve 209001 LPCS 02 Describe the method of operation of the ADS 3.8 BOTH 218000 K6.01: NEW moved to CFR41.7 logic given LPCS and LPCI A out of service q040 3.9 K6.02: 4.1 K3 during a LOCA.

209002 HPCS 01 Given a spurious initiation of the HPCS 3.9 519 BOTH 259002 BANK CFR41.7/41.8 system, determine the effect on Reactor water q041 A2.08: 4.5 NRC 6/01 level.

211000 SLC 04 Given the initiation of Standby Liquid Control 4.5 BOTH A4.01: 3.9 NEW CFR41.1/41.6/41.7/43.6 in an ATWS, discern the parameters q042 A4.03: 4.1 indicating injection to the reactor. A4.06: 3.9 A4.07: 3.6 212000 RPS 01 Describe the response of the RPS Power 3.6 102 BOTH K1.04: 3.4 BANK CFR41.6 System upon an ESF inverter loss. q043 NRC 3/98 215003 IRM 01 Identify the power supply to the Intermediate 2.5 BOTH NEW CFR41.6 Range Nuclear Instrumentation. q044 PAGE 1 TOTAL TIER 2 GROUP 1 1 1 2 0 1 2 0 1 0 1 0 PAGE TOTAL # QUESTIONS 9 209001 random selection K5. Topic moved to K3 due to low discriminatory value and importance values of Topic K5.

REVISION 0 4/29/2002 PAGE 7 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-2 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 215004 Source Range Monitor 2. Determine the conditions that would allow the 3.4 BOTH A4.04: 3.2 moved to CFR41.5/41.6 1. operation of Source Range Monitor detector q046 generics 32 drives.

215005 APRM / LPRM 03 Determine the affects on APRMs with reduced 3.1 326 BOTH BANK CFR41.6/41.7 LPRM inputs. q045 NRC 12/00 216000 Nuclear Boiler Instrumentation 23 Determine the ability of the Recirculation 3.3 122 BOTH 202001 BANK CFR41.5 Pumps to start based on Reactor Temperature. q047 A4.01: 3.7 NRC 3/98 217000 RCIC 01 Given plant conditions, determine the 3.7 328 BOTH BANK CFR41.5/41.7/41.10 operation of RCIC and indications of q048 NRC 12/00 injection.

218000 ADS 03 Given plant conditions determine automatic 3.8 BOTH NEW CFR41.7 operation of Automatic Depressurization. q049 223001 Primary CTMT and Auxiliaries 08 Given plant conditions and electrical busses 2.7 528 BOTH K2.09: 2.7 BANK CFR41.7/41.8 that are unavailable, determine which q050 K2.10: 2.7 NRC 6/01 components are available.

223002 PCIS / Nuclear Steam Supply 09 Given plant conditions, evaluate the systems 3.6 385 BOTH 2.4.21: 3.7 BANK Shutoff that should actuate or isolate. q051 2.4.4: 4.0 NRC 12/00 CFR41.7/41.9 239002 SRVs 05 Describe the operation of the Safety Relief 3.1 337 BOTH BANK CFR41.3 Valves in different modes of operation. q052 NRC 12/00 (system air pressure or reactor pressure) 241000 Reactor / Turbine Pressure 11 Describe the response of the plant with a 3.4 244 BOTH A1.01: 3.9 BANK Regulator failure of the Main Stop and Control Valves q053 A1.02: 4.1 NRC 4/00 CFR41.5 closed with the reactor at power. A1.07: 3.8 259001 Reactor Feedwater 03 Given a set of plant conditions, determine the 2.8 548 RO BANK CFR41.5/41.10/43.5 operational limitations for the Reactor Feed q077 NRC 6/01 Pump Turbines 259002 Reactor Water Level Control 10 Describe the operation of the Feedwater Level 3.4 273 RO K6.04: 3.1 BANK CFR41.5 Control System on a failure of the Feed Flow q078 NRC 4/00 input signal.

PAGE 2 TOTALS TIER 2 GROUP 1 2 1 0 2 1 2 1 1 0 0 1 PAGE 2 TOTAL # QUESTIONS 11 215004 random section A4. Topic moved to generic that covers both the random selection A4.04 and generic 2.1.32.

REVISION 0 4/29/2002 PAGE 8 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-2 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 261000 SGTS 01 Given conditions determine the automatic 3.7 BOTH NEW CFR41.7/41.11 operation of Standby Gas Treatment. q054 264000 EDGs 04 Identify the signal that will automatically start 3.2 RO NEW CFR41.4/41.7 the associated Standby Service Water System q079 and align it to the Diesel Generator.

217000 RCIC 01 Determine the indications of RCIC following 3.5 676 RO A3.02: 3.6 MOD CFR41.7 overspeed trip and attempting to open the trip q080 Aud 6/01 throttle valve.

202002 Recirculation Flow Control 01 Given plant conditions, identify the plant 3.6 235 BOTH BANK Recirc Pumps CFR41.6 response to a Recirc Flow Control Runback. q055 NRC 4/00 in abnormal pump configuration 239002 SRVs 06 Describe the purpose of the SRV tailpipe 2.7 RO 2.1.28: 3.2 NEW CFR41.7 vacuum breakers. q081 259001 Reactor Feedwater 01 Given a failure of the RFPT Lube Oil System, 3.7 BOTH K1.11: 2.7 BANK CFR41.4 identify the Reactor Feedwater System q056 K4.06: 2.5 response. K6.09: 2.8 259002 Reactor Water Level Control 03 Describe the effects on the Reactor Water 3.8 BOTH BANK CFR41.4/41.7 Level Control System from a failure of the q057 Narrow Range Reactor Water level signal.

264000 EDGs 2. During degraded grid conditions, determine 4.0 11 BOTH K4.05: 3.2 BANK CFR41.8 4. the response of the diesel generators. q058 A3.05: 3.4 NRC 6/01 4

PAGE 3 TOTALS TIER 2 GROUP 1 2 0 0 1 1 0 0 1 2 0 1 PAGE TOTAL # QUESTIONS 8 PAGE 1 TOTALS TIER 2 GROUP 1 1 1 2 0 1 2 0 1 0 1 0 PAGE TOTAL # QUESTIONS 9 PAGE 2 TOTALS TIER 2 GROUP 1 2 1 0 2 1 2 1 1 0 0 1 PAGE TOTAL # QUESTIONS 11 K/A CATEGORY TOTALS: 5 2 2 3 3 4 1 3 2 1 2 TIER 2 GROUP 1 GROUP POINT TOTAL 28 REVISION 0 4/29/2002 PAGE 9 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-2 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 2 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # / BOTH K/A 201003 Control Rod and Drive Mechanism 08 Given conditions of the HCU, determine its 3.8 341 RO BANK CFR41.6/41.10/43.5 status. q082 NRC 12/00 202001 Recirculation 10 Given plant conditions in regard to Reactor 3.5 540 BOTH A1.09: 3.3 BANK CFR41.3/41.5 Recirc. Pump seals, determine the failed q059 A1.10: 2.6 NRC 6/01 mechanism.

204000 RWCU 02 During a reactor startup and heatup, determine 3.1 RO NEW CFR41.4 the effects on reactor water level with a loss of q083 Reactor Water Cleanup operation.

205000 Shutdown Cooling 03 Given plant conditions and configuration 3.4 541 BOTH K3.03: 3.8 BANK CFR41.2/41.3/41.4/41.5 lineup, determine valid method for q060 A1.03: 3.3 NRC 6/01 determining Reactor coolant temperature. A1.06: 3.7 A1.08: 3.1 219000 RHR /LPCI Suppression Pool 04 Describe the method used to control 2.9 BOTH A4.12: 4.1 NEW moved to Cooling Mode Suppression Pool Temperature and cooldown q061 K5 CFR41.7 rate.

226001 RHR/LPCI: CTMT Spray Mode 07 Determine conditions that would result in 3.5 BOTH A3.01: 3.0 BANK moved to CFR41.7/41.8 automatic initiation of RHR Containment q062 A3 Spray mode.

239001 Main and Reheat Steam 03 Following an isolation of the Main Steam 3.2 RO K1.22: 3.1 NEW CFR41.4/41.14 Isolation Valves, describe the operation of the q084 Condensate and Feedwater Systems.

245000 Main Turbine Gen. and Auxiliaries 10 Discuss the basis for the limit on Main 2.5 44 BOTH K4.06: 2.7 BANK GGNS CFR41.4/41.10/43.5 Generator Reactive loading at GGNS. q063 A4.14: 2.5 NRC 3/98 generator has lower limits due to reverse power PAGE 1 TOTAL TIER 2 GROUP 2 1 0 2 0 1 0 0 2 2 0 0 PAGE TOTAL # QUESTIONS 8 219000 random selection K2. Topic moved to K5 due to low discriminatory value of K2 topic for RHR Suppression Pool Cooling.

226001 random selection K5. Topic moved to A3 due to low discriminatory value of K5 topic for RHR Containment Spray Mode.

REVISION 0 4/29/2002 PAGE 10 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE CONT. ES-401-2 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 2 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 256000 Reactor Condensate 08 Describe the effects of Low Pressure 3.1 294 BOTH A3.01: 2.7 BANK CFR41.4 Feedwater Heater isolation on the Condensate q064 A3.04: 3.0 NRC 4/00 System operation and plant operations. A3.07: 2.9 262001 AC Electrical Distribution 04 Given a loss of offsite power and plant 3.4 BOTH MOD CFR41.4/41.7 conditions, determine the type of load q065 sequencing to occur on ESF busses.

262002 UPS (AC/DC) 01 Concerning the ESF Static Inverters, identify 2.8 544 BOTH A2.01: 2.6 BANK CFR41.7/41.10/43.5 the correct response of the inverter on a loss of q066 A1.01: 2.4 NRC 6/01 normal power supply.

263000 DC Electrical Distribution 01 Identify the power supply to the DC Turbine 3.1 RO NEW moved to CFR41.4 Building Cooling Water Pump and its start q085 K2 Plant signal. modification 271000 Offgas 09 Determine the startup sequence of the Offgas 3.3 RO 2.1.32: 3.4 NEW CFR41.7/41.13 system to prevent explosive mixtures from q086 forming.

272000 Radiation Monitoring 01 Identify the normal radiation monitoring 3.2 BOTH NEW CFR41.10/41.11/43.4/43.5 alarms received on a reactor down power from q067 full power and their cause. (Hydrogen Water Chemistry) 286000 Fire Protection 08 Discern the response of the Diesel Driven Fire 3.2 3 BOTH K5.05: 3.0 BANK CFR41.4 Pumps on an auto start signal with a failure to q068 K4.07: 3.3 NRC 3/98 start. A3.01: 3.4 A4.06: 3.4 290001 Secondary CTMT 09 Describe the ability of Auxiliary Building Fire 3.4 264 BOTH A2.06: 3.7 BANK CFR41.9 Protection system to be restored following q069 286000 NRC 4/00 Auxiliary Building isolation in conjunction A2.09: 2.7 with a loss of AC power.

PAGE 2 TOTAL TIER 2 GROUP 2 0 1 0 0 0 1 1 2 2 1 0 PAGE TOTAL # QUESTIONS 8 263000 random selection A3. Topic moved to K2 due to low discriminatory value of A3 and plant modification supports question from K2.

REVISION 0 4/29/2002 PAGE 11 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE CONT. ES-401-2 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 2 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 290003 Control Room HVAC 05 Identify the response of the Control Room 3.2 BOTH NEW CFR41.7/41.11/43.4 HVAC system upon receipt of radiation q070 alarms.

300000 Instrument Air 02 Describe the crosstie between Service and 3.0 BOTH BANK moved to CFR41.4/41.10/43.5 Instrument Air and when it occurs. q071 K3 400000 Component Cooling Water 01 Discuss the status of the standby feature of the 3.4 105 BOTH BANK CFR41.4 ESF powered CCW pump during a LOCA. q072 NRC 3/98 PAGE 3 TOTALS 0 0 0 2 0 0 1 0 0 0 0 PAGE 3 TOTAL # QUESTIONS 3 PAGE 1 TOTALS 1 0 2 0 1 0 0 2 2 0 0 PAGE 1 TOTAL # QUESTIONS 8 PAGE 2 TOTALS 0 1 0 0 0 1 1 2 2 1 0 PAGE 2 TOTAL # QUESTIONS 8 K/A CATEGORY TOTALS: 1 1 2 2 1 1 2 4 4 1 0 TIER 2 GROUP 2 GROUP POINT TOTAL 19 300000 random selection A4. Topic moved to K3 due to low discriminatory value of single subject in A4.

REVISION 0 4/29/2002 PAGE 12 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-2 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 3 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # / BOTH K/A 215001 Traversing In-core Probe 233000 Fuel Pool Cooling and Cleanup 07 Identify the Tech Spec Limits on Spent Fuel 3.0 RO NEW CFR41.7 Pool Temperature. q087 234000 Fuel Handling Equipment 239003 MSIV Leakage Control 268000 Radwaste 02 Given conditions and a Liquid Radwaste 2.6 110 BOTH 272000 BANK Process Rad CFR41.13/43.4 discharge in progress, determine whether a q073 A3.03: 3.0 NRC 3/98 Monitor release will continue. isolation 288000 Plant Ventilation 05 Describe the response of the Auxiliary 3.3 BOTH K1.02: 3.4 NEW CFR41.7 Building Ventilation System on High q074 Radiation.

290002 Reactor Vessel Internals 05 Identify the consequence of exceeding a 3.7 BOTH K5.01: 3.5 NEW moved to CFR41.3/41.14/43.2 thermal limit. q075 A2 K/A CATEGORY TOTALS: 1 0 0 0 0 0 1 2 0 0 0 TIER 2 GROUP 3 GROUP POINT TOTAL 4 290002 random selection A1. Topic moved to A2 due to no subjects in A1.

REVISION 0 4/29/2002 PAGE 13 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR RO EXAMINATION OUTLINE ES-401-5 AUGUST 2002 GENERIC KNOWLEDGE AND ABILITIES TIER 3 CATEGORY C1 C2 C3 C4 TOPIC(S) IMP REC # SRO/RO RELATED ORIGIN NOTES:

/BOTH K/A CONDUCT OF OPERATIONS - Mode of Operation 2.1.22 Given plant conditions determine the Tech Spec 2.8 585 RO BANK CFR41.10/43.1 Operational Mode of the plant. q088 NRC 6/01 CONDUCT OF OPERATIONS - Conduct of 2.1.1 Given components, determine components that are 3.7 RO NEW Operations Requirements CONTROLS of the facility. q089 CFR41.10/43.5 CONDUCT OF OPERATIONS - Shift Turnover 2.1.3 Given a change of operators at other than normal 3.0 RO NEW CFR41.10/43.5 shift turnover, identify requirements for relief. q090 CONDUCT OF OPERATIONS - Component 2.1.28 Identify the purpose for injecting Oxygen into the 3.2 RO NEW purpose Condensate System. q091 CFR41.4 EQUIPMENT CONTROL - Configuration Control 2.2.11 Given a component temporarily out of normal 2.5 387 RO BANK SOER 98-1 Safety CFR41.10/43.5 alignment per system operating instructions, q092 NRC 12/00 System Status determine the tracking mechanism to be employed. Control EQUIPMENT CONTROL - Protective Tagging 2.2.13 Identify the proper method of tagging an air- 3.6 RO NEW CFR41.10/43.5 operated component for system isolation. q093 RADIATION CONTROL - Radiation Reduction 2.3.10 Identify methods of reducing radiation levels in 2.9 RO NEW CFR41.12/43.4 work areas. q094 RADIATION CONTROL - ALARA 2.3.2 Describe the personnel hazards when operating 2.5 287 RO 2.1.32: 3.4 BANK Internal Hazard to CFR41.10/43.4 RHR in Suppression Pool Cooling mode. (ALARA) q095 NRC 6/01 personnel EMERGENCY PROCEDURES / PLAN - EOP/SAP 2.4.4 Given plan conditions, identify the Emergency 4.0 298 RO BANK Entry into SAP CFR41.10/43.5 Operating Procedures (SAP) requiring q096 NRC 4/00 implementation.

EMERGENCY PROCEDURES / PLAN - Local 2.4.35 Describe the response of Non-Licensed Operators 3.3 577 RO BANK E-Plan changes Operator Emergency Response CFR41.10/43.5 during implementation of the Emergency Plan q097 NRC 6/01 2/2001 EMERGENCY PROCEDURES / PLAN - AOP 2.4.49 Given plant conditions, identify the immediate 4.0 RO NEW AOP Immediate Immediate Operator Actions operator actions to be taken. q098 actions CFR41.10/43.5 EMERGENCY PROCEDURES / PLAN - 2.4.43 Identify methods and times for notification of 2.8 RO NEW Communications Offsite agencies in the event of Emergency Plan q099 CFR41.10/43.5 activation.

EMERGENCY PROCEDURES / PLAN - Fire 2.4.25 Describe the Control Room Operator Actions for a 2.9 578 RO BANK Protection actions CFR41.10/43.5 fire in Division I Diesel Generator Room. q100 NRC 6/01 K/A CATEGORY TOTALS: 4 2 2 5 TIER 3 GROUP POINT TOTAL 13 REVISION 0 4/29/2002 PAGE 14 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1

ES-401 FORM ES-401-1 BWR SRO EXAMINATION OUTLINE Facility: GRAND GULF NUCLEAR STATION Date of Exam: 23 AUGUST 2002 K/A CATEGORY POINTS TIER GROUP K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G POINT

  • TOTAL
1. 1 4 5 4 7 5 1 26 Emergency &

Abnormal 2 2 3 3 4 4 1 17 Plant Evolutions TIER 6 8 7 11 9 2 43 TOTAL 1 3 1 2 2 0 5 1 2 3 2 2 23 2.

Plant 2 1 1 0 2 2 0 3 3 1 0 0 13 Systems 3 1 0 0 0 0 0 1 2 0 0 0 4 TIER 5 2 2 4 2 5 5 7 4 2 2 40 TOTAL CAT 1 CAT 2 CAT 3 CAT 4

3. Generic Knowledge & Abilities 5 4 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 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.

REVISION 0 4/29/2002 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 1 2 # /BOTH K/A 295003 Partial or Complete Loss of AC Power/ 6 01 Given a lockout on BOP Transformer 12B, determine the 3.8 507 BOTH AK2.03: 3.9 BANK CFR41.7 configuration of the AC Distribution System. q014 NRC 6/01 295006 SCRAM / 1 02 Describe the position of control rods following a reactor 4.4 10 BOTH 201005 BANK Rx SCRAM CFR41.6/41.10/43.5 scram and how position is determined. q002 A3.02: 3.5 NRC 6/01 Immediate A4.02: 3.7 Actions 201003 K4.05: 3.3 A3.01: 3.6 A4.03: 3.5 295007 High Reactor Pressure / 3 01 Describe the response of the Turbine Pressure Control 3.7 69 BOTH 241000 BANK CFR41.5 System on an increasing reactor pressure. q003 K4.01: 3.8 NRC 6/01 A2.02: 3.7 295009 Low Reactor Water Level / 2 02 Given plant conditions, identify the status of 3.1 121 BOTH AK2.03: 3.2 BANK CFR41.5/43.5 Recirculation pumps originally operating in fast speed q004 AA1.03: 3.1 NRC 3/98 with a lowering water level.

295010 High Drywell Pressure / 5 04 Given plant parameters and elevated in-leakage into the 3.8 286 BOTH AK3.05: 3.4 BANK GGNS Drywell CFR41.5 Drywell, determine the status of reactor coolant system q005 2.4.21: 4.3 NRC 4/00 air leak integrity.

295013 High Suppression Pool Water Temp. / 5 01 During a surveillance operating RCIC, determine how 4.0 BOTH AA1.02: 3.9 NEW CFR41.5/41.10/43.2/43.5 often Suppression Pool Temperature is required to be q018 2.1.33: 4.0 monitored and the threshold for alternate actions.

295014 Inadvertent Reactivity Addition / 1 01 With the reactor in startup conditions such that the reactor 4.1 204 BOTH BANK Susquehanna CFR41.1/41.2/41.6/43.6 has dropped subcritical, what are the operator actions if a q006 NRC 6/01 reactivity event high worth control rod is withdrawn fully. 7/98 295015 Incomplete SCRAM / 1 01 Describe the method to be used to allow the insertion of 3.7 203 BOTH BANK CFR41.6/43.6 control rods using RCIS during an ATWS. q007 NRC 6/01 295016 Control Room Abandonment / 7 04 Determine the impact of operation of circuit breakers 3.2 701 SRO 239001 MOD CFR41.7/41.10/43.5 during operations from the Remote Shutdown Panel. q076 K2.01: 3.3 NRC 3/98 295017 High Offsite Release Rate / 9 08 Given plant conditions with MSIV Leakage Control 3.1 BOTH AA1.09: 3.6 NEW CFR41.11/41.13/43.4 System operating, determine the mechanism for q019 monitoring radiological release to locations outside the Auxiliary Building.

PAGE 1 TOTAL TIER 1 GROUP 1 0 2 3 3 2 0 PAGE TOTAL # QUESTIONS 10 REVISION 0 4/29/2002 PAGE 1 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-1 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 1 2 # /BOTH K/A 295023 Refueling Accidents / 8 03 Describe the requirements to ensure proper positioning of 3.6 187 SRO BANK Damaged Fuel CFR41.2/41.6/43.6/43.7 spent fuel bundles being moved on the Refueling q077 NRC 6/01 at RBS Platform Grapple.

295024 High Drywell Pressure / 5 14 Given conditions determine ability to initiate 3.9 601 BOTH NEW High Drywell CFR41.9/41.10/43.5 Containment Spray per EOPs. q008 Pressure is EP entry for CTMT Press. Control 295025 High Reactor Pressure / 3 05 State the Reactor Vessel pressure Safety Limit and its 4.7 SRO 2.2.22: 4.1 MOD CFR41.3/43.2 basis and actions to be taken if violated. q080 2.2.25: 3.7 EK1.02: 4.2 295026 Suppression Pool High Water Temp. / 5 01 Given plant conditions, determine when Suppression Pool 4.0 301 BOTH BANK CFR41.7/41.9/41.10/41.14/43.5 Cooling is no longer effective and alternate actions are q024 NRC required. 12/00 295027 High Containment Temperature / 5 03 Given plant conditions, determine the Technical 3.7 512 BOTH 2.2.25: 3.7 BANK CFR41.9/41.10/43.2 Specification Bases for shutting down the Reactor due to q025 NRC 6/01 a high Containment Temperature.

295030 Low Suppression Pool Water Level / 5 04 Given plant conditions, determine the method to use to 4.0 SRO NEW CFR41.9/41.10/43.5 makeup inventory to the Suppression Pool. q079 295031 Reactor Low Water Level / 2 01 Given plant conditions and a low reactor water level, 4.7 309 BOTH 2.1.1: 3.8 BANK CFR41.2/41.3/41.10/43.5 determine core cooling mechanism and adequacy. q009 2.4.21: 4.3 NRC 12/00 295037 SCRAM Condition Present and Reactor 04 Determine when plant conditions allow the termination of 4.5 216 BOTH EK1.04: 3.6 BANK Power Above APRM Downscale or Unknown / 1 Standby Liquid Control with multiple control rods stuck q010 EK1.05: 3.6 NRC 4/00 CFR41.1/41.2/41.6/43.6 out and what is the basis. EA2.03: 4.4 295038 High Offsite Release Rate / 9 02 Given plant conditions and procedures, determine the 4.4 BOTH 2.4.44: 4.0 BANK CFR41.10/41.12/43.4/43.5 protective action recommendations to be recommended to q030 the state and local officials in an emergency.

500000 High Containment Hydrogen Conc. / 5 01 Determine the bases for the Hydrogen Control leg of 3.9 505 SRO 2.4.18: 3.6 BANK CFR41.10/43.5 EP 3. q078 NRC 6/01 295009 Low Reactor Water Level / 2 02 Given a steam flow / feed flow mismatch and plant 3.7 SRO NEW moved to CFR41.7/43.5 conditions, determine the reactor water level response and q081 AA2 response of Reactor Water Level control.

PAGE 2 TOTAL TIER 1 GROUP 1 4 2 1 3 1 0 PAGE TOTAL # QUESTIONS 11 295009 random selection AK3. Topic is adequately covered on other questions, moved to topic AA2.

REVISION 0 4/29/2002 PAGE 2 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-1 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295003 Partial or Complete Loss of AC Power / 6 03 During a station blackout with DG 13 available, 3.9 SRO AK1.06: 4.0 NEW CFR41.4/43.5 determine the alignment to supply power from 17AC to q082 AA1.01: 3.8 16AB.

295016 Control Room Abandonment / 7 02 Given parameters from the Remote Shutdown Panel 4.3 603 BOTH 2.1.25: 3.1 MOD CFR41.5/41.10/43.5 indications, determine actual and Narrow Range RPV q017 2.4.4: 3.6 NRC 6/01 level.

295014 Inadvertent Reactivity Addition / 1 02 Determine effects on the reactor of the fast opening a 3.8 116 BOTH 202002 BANK CFR41.6/43.6 Recirc Flow Control Valve. q012 K1.02: 4.2 NRC 3/98 K3.02: 4.0 295030 Low Suppression Pool Water Level / 5 02 Evaluate Suppression Pool Temperature, with a Low 3.9 8 BOTH BANK Caution 2 CFR41.7/41.9 Suppression Pool Level. q028 NRC 6/01 EOP-2 500000 High Containment Hydrogen Conc. / 5 2. Given applicable SOP, graph, and plant conditions, 4.2 219 BOTH EA1.03: 3.2 BANK moved to CFR41.10/43.5 1. determine the power settings for the Hydrogen q011 2.1.25:3.1 NRC 4/00 Generics 20 Recombiners and time to full power.

PAGE 3 TOTAL TIER 1 GROUP 1 0 1 0 1 2 1 PAGE TOTAL # QUESTIONS 5 PAGE 1 TOTAL TIER 1 GROUP 1 0 2 3 3 2 0 PAGE TOTAL # QUESTIONS 10 PAGE 2 TOTAL TIER 1 GROUP 1 4 2 1 3 1 0 PAGE TOTAL # QUESTIONS 11 K/A CATEGORY TOTALS: 4 5 4 7 5 1 TIER 1 GROUP 1 GROUP POINT TOTAL 26 500000 random selection was EA. Topic covers both EA 1.03 operation of CTMT Atmosphere control system and Generic procedure usage; moved to Generics due to higher importance.

REVISION 0 4/29/2002 PAGE 3 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295001 Partial or Complete Loss of Forced Core 2. Given plant conditions and the power to flow map, 3.6 602 BOTH 2.4.1: 4.6 MOD moved to Flow Circulation / 1 & 4 4. determine the actions to be taken. q013 AA2.01: 3.8 NRC 4/00 Generics CFR41.5/41.10/43.5 11 2.1.25: 3.1 295002 Loss of Main Condenser Vacuum / 3 04 Describe the basis for the isolation of the Main Steam 3.4 220 BOTH AK3.05: 3.4 BANK CFR41.4/43.4 Isolation Valves on a loss of condenser vacuum. q015 NRC 4/00 295004 Partial or Complete Loss of DC Power / 6 295005 Main Turbine Generator Trip / 3 05 Given a spurious Main Turbine Generator trip, determine 3.9 553 BOTH AA2.04: 3.8 BANK CFR41.5/41.6 the initial effect on Reactor Power and Reactor Pressure. q001 AA2.03: 3.1 NRC 6/01 295008 High Reactor Water Level / 2 01 Identify the effects of a High Reactor Water Level on the 3.2 275 BOTH 245000 BANK CFR41.4/41.5 Main Turbine and Reactor Feed Pump Turbines. q016 A3.01: 3.6 NRC 6/01 259001 K6.07: 3.8 295011 High Containment Temperature / 5 295012 High Drywell Temperature / 5 295018 Partial or Complete Loss of CCW / 8 03 Given a partial loss of Component Cooling Water, 3.3 314 BOTH BANK CFR41.4/41.10/43.5 determine actions to be taken and their basis. q020 NRC 12/00 295019 Partial or Complete Loss of Inst. Air / 8 14 Given a reduction in Instrument Air Header pressure, 3.2 BOTH NEW New Air Dryers CFR41.4/41.10/43.5 determine a possible cause. q021 moved to AK2 295020 Inadvertent Cont. Isolation / 5 & 7 03 Given a loss of Instrument Air System pressure to the 3.3 BOTH AK3.03: 3.2 NEW CFR41.4/41.7/41.9 Auxiliary Building, determine the effects on the ability to q022 remove heat from the Containment.

295021 Loss of Shutdown Cooling / 4 01 Given plant parameters and the graphs from the 3.6 604 BOTH AK1.01: 3.8 MOD Multiple graphs CFR41.5/41.10/43.5 Inadequate Decay Heat Removal ONEP, determine time q032 NRC 3/98 for various to boil. conditions.

PAGE 1 TOTAL TIER 1 GROUP 2 1 2 1 1 2 1 PAGE TOTAL # QUESTIONS 8 295001 random selection was AA2. Topic moved to Generics that are in addition to the random selection.

295019 random selection AK1 has NONE. Moved selection to AK2 to support question concerning new Instrument Air Dryer System failures.

REVISION 0 4/29/2002 PAGE 4 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-1 AUGUST 2002 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 E/APE #/NAME/SAFETY FUNCTION K K K A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 1 2 # BOTH K/A 295022 Loss of CRD Pumps / 1 04 Describe the affects on reactor water level during a 2.6 55 BOTH AK2.04: 2.7 BANK CFR41.5 reactor startup with minimal decay heat and a loss of q023 AK2.05: 2.5 NRC 6/01 CRD Pumps. (RWCU is lined up to blowdown to the main condenser to compensate for CRD flow.)

295028 High Drywell Temperature / 5 03 Given rising temperatures in the Reactor lower cavity 3.9 BOTH EK2.04: 3.6 NEW CFR41.4/41.7 area, determine the response of the Drywell Cooling q026 EK3.04: 3.8 system.

295029 High Suppression Pool Water Level / 5 01 Identify the bases for Emergency RPV Depressurization 3.7 513 BOTH BANK CFR41.9/41.10 when Suppression Pool Level cannot be maintained q027 NRC 6/01 below 24.4 feet.

295032 High Secondary Containment Area 03 Determine the systems affected by high temperatures in 3.9 48 BOTH EK2.07: 3.8 BANK Steam leak at Temperature / 5 the RHR A Pump Room. q033 219000 NRC 3/98 GGNS in Steam CFR41.4/41.10/43.5 A1.08: 3.6 Condensing A2.14: 4.3 piping 295033 High Secondary Containment Area 03 Given operations during a Refueling outage, determine 3.7 BOTH NEW Fuel Pool HX in Radiation Levels / 9 the cause of elevated radiation levels in the area of the q029 an open area of CFR41.12/43.4 Fuel Pool Cooling Heat Exchangers. Sec CTMT 295034 Secondary Containment Ventilation High 05 Given plant conditions, determine the configuration of the 3.7 SRO NEW Radiation / 9 Fuel Handling area Ventilation system. q083 CFR41.4/41.10/41.13/43.4 295035 Secondary Containment High Differential 02 Describe the operation of the Standby Gas Treatment 3.8 BOTH BANK Pressure / 5 System with regard to Auxiliary Building and Enclosure q034 CFR41.4/41.7 Building Pressures.

295036 Secondary Containment High Sump/Area 03 Given a rising water level in an ECCS Pump Room, 3.8 58 BOTH EK3.03: 3.6 BANK EOP-4 Water Level / 5 determine appropriate actions to be taken with regard to q035 EA2.02: 3.1 NRC 3/98 CFR41.4/41.10/43.5 the overall plant operation.

600000 Plant Fire On Site / 8 04 Given a fire at the Hydrogen Bulk Storage Facility 3.4 BOTH 2.4.25: 3.4 NEW moved to CFR41.10/43.5 describe the actions to be taken to combat the fire. q031 2.1.32: 3.8 AK3 PAGE 2 TOTAL TIER 1 GROUP 2 1 1 2 3 2 0 PAGE TOTAL # QUESTIONS 9 PAGE 1 TOTAL TIER 1 GROUP 2 1 2 1 1 2 1 PAGE TOTAL # QUESTIONS 8 K/A CATEGORY TOTALS: 2 3 3 4 4 1 TIER 1 GROUP 2 GROUP POINT TOTAL 17 600000 random selection was AA1. Topic was moved to AK3 due to limited discriminatory value of AA1 for Licensed Operators. AK3 allows testing of precaution.

REVISION 0 4/29/2002 PAGE 5 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 1 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 201005 RCIS 01 Given a failure of the Turbine First Stage 3.2 BOTH NEW CFR41.6/43.6 pressure signal to RCIS, determine the mode q037 of control rod movement.

202002 Recirculation Flow Control 01 Given plant conditions following an actuation 3.4 BOTH 202001 NEW EOC-RPT CFR41.6 of EOC-RPT, determine the Recirculation q038 A2.15: 3.9 changes Pump circuit breaker configuration. 2002 203000 RHR/LPCI: Injection Mode 08 State the basis for monitoring reactor pressure 4.3 60 BOTH K4.01: 4.2 BANK Where is CFR41.8 when aligning the RHR system for injection q039 K4.02: 3.3 NRC 6/01 pressure into the vessel for the LPCI mode. A3.01: 3.8 sensed on A3.08: 4.1 LPCI for K1.17: 4.0 operation of the injection valve 209001 LPCS 02 Describe the method of operation of the ADS 3.9 BOTH 218000 NEW moved to CFR41.7 logic given LPCS and LPCI A out of service q040 K6.01: 4.1 K3 during a LOCA. K6.02: 4.1 209002 HPCS 01 Given a spurious initiation of the HPCS 3.9 519 BOTH 259002 BANK CFR41.7/41.8 system, determine the effect on Reactor water q041 A2.08: 4.5 NRC 6/01 level.

211000 SLC 04 Given the initiation of Standby Liquid Control 4.6 BOTH A4.01: 3.9 NEW CFR41.1/41.6/41.7/43.6 in an ATWS, discern the parameters q042 A4.03: 4.1 indicating injection to the reactor. A4.06: 3.9 A4.07: 3.6 212000 RPS 01 Describe the response of the RPS Power 3.8 102 BOTH K1.04: 3.6 BANK CFR41.6 System upon an ESF inverter loss. q043 NRC 3/98 215004 Source Range Monitor 2. Determine the conditions that would allow the 3.4 BOTH A4.04: 3.2 moved to CFR41.6/41.5 1. operation of Source Range Monitor detector q046 generics 32 drives.

PAGE 1 TOTAL TIER 2 GROUP 1 0 0 2 0 0 2 0 1 0 2 1 PAGE TOTAL # QUESTIONS 8 209001 random selection K5. Topic moved to K3 due to low discriminatory value and importance values of Topic K5.

215004 random section A4. Topic moved to generic that covers both the random selection A4.04 and generic 2.1.32.

REVISION 0 4/29/2002 PAGE 6 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 215005 APRM / LPRM 03 Determine the affects on APRMs with reduced 3.3 326 BOTH BANK CFR41.6/41.7 LPRM inputs. q045 NRC 12/00 216000 Nuclear Boiler Instrumentation 23 Determine the ability of the Recirculation 3.4 122 BOTH 202001 BANK CFR41.5 Pumps to start based on Reactor Temperature. q047 A4.01: 3.7 NRC 3/98 217000 RCIC 01 Given plant conditions, determine the 3.7 328 BOTH BANK CFR41.5/41.7/41.10 operation of RCIC and indications of q048 NRC injection. 12/00 218000 ADS 03 Given plant conditions determine automatic 4.0 BOTH NEW CFR41.7 operation of Automatic Depressurization. q049 223001 Primary CTMT and Auxiliaries 08 Given plant conditions and electrical busses 3.0 528 BOTH K2.09: 2.9 BANK CFR41.7/41.8 that are unavailable, determine which q050 K2.10: 2.9 NRC 6/01 components are available.

223002 PCIS / Nuclear Steam Supply 09 Given plant conditions, evaluate the systems 3.7 385 BOTH 2.4.21: 4.3 BANK Shutoff that should actuate or isolate. q051 2.4.4: 4.3 NRC CFR41.7/41.9 12/00 226001 RHR/LPCI: CTMT Spray Mode 07 Determine conditions that would result in 3.5 BOTH A3.01: 3.0 BANK moved to CFR41.7/41.8 automatic initiation of RHR Containment q062 A3 Spray mode.

239002 SRVs 05 Describe the operation of the Safety Relief 3.3 337 BOTH BANK CFR41.3 Valves in different modes of operation. q052 NRC (system air pressure or reactor pressure) 12/00 241000 Reactor / Turbine Pressure 11 Describe the response of the plant with a 3.4 244 BOTH A1.01: 3.8 BANK Regulator failure of the Main Stop and Control Valves q053 A1.02: 3.9 NRC 4/00 CFR41.5 closed with the reactor at power. A1.07: 3.7 PAGE 2 TOTALS TIER 2 GROUP 1 2 1 0 1 0 2 1 1 1 0 0 PAGE 2 TOTAL # QUESTIONS 9 226001 random selection K5. Topic moved to A3 due to low discriminatory value of K5 topic for RHR Containment Spray Mode.

REVISION 0 4/29/2002 PAGE 7 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 259002 Reactor Water Level Control 03 Describe the effects on the Reactor Water 3.8 BOTH BANK CFR41.4/41.7 Level Control System from a failure of the q057 Narrow Range Reactor Water level signal.

261000 SGTS 01 Given conditions determine the automatic 3.8 BOTH NEW CFR41.7/41.11 operation of Standby Gas Treatment. q054 262001 AC Electrical Distribution 04 Given a loss of offsite power and plant 3.6 BOTH MOD CFR41.1/41.7 conditions, determine the type of load q065 sequencing to occur on ESF busses.

264000 EDGs 2. During degraded grid conditions, determine 4.3 11 BOTH K4.05: 3.2 BANK CFR41.8 4. the response of the diesel generators. q058 A3.05: 3.4 NRC 6/01 4

290001 Secondary CTMT 09 Describe the ability of Auxiliary Building Fire 3.6 264 BOTH A2.06: 4.0 BANK CFR41.9 Protection system to be restored following q069 286000 NRC 4/00 Auxiliary Building isolation in conjunction A2.09: 2.8 with a loss of AC power.

202002 Recirculation Flow Control 01 Given plant conditions, identify the plant 3.4 235 BOTH BANK Recirc Pumps CFR41.6 response to a Recirc Flow Control Runback. q055 NRC 4/00 in abnormal pump configuration PAGE 3 TOTALS TIER 2 GROUP 1 1 0 0 1 0 1 0 0 2 0 1 PAGE TOTAL # QUESTIONS 6 PAGE 1 TOTALS TIER 2 GROUP 1 0 0 2 0 0 2 0 1 0 2 1 PAGE TOTAL # QUESTIONS 8 PAGE 2 TOTALS TIER 2 GROUP 1 2 1 0 1 0 2 1 1 1 0 0 PAGE TOTAL # QUESTIONS 9 K/A CATEGORY TOTALS: 3 1 2 2 0 5 1 2 3 2 2 TIER 2 GROUP 1 GROUP POINT TOTAL 23 REVISION 0 4/29/2002 PAGE 8 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 2 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # / BOTH K/A 201001 CRD Hydraulic 02 During a reactor scram with scram signals not 2.6 BOTH MOD CFR41.5/41.6 reset, determine the CRD System flow rates. q036 202001 Recirculation 10 Given plant conditions in regard to Reactor 3.9 540 BOTH A1.09: 3.3 BANK CFR41.3/41.5 Recirc. Pump seals, determine the failed q059 A1.10: 2.7 NRC 6/01 mechanism.

204000 RWCU 205000 Shutdown Cooling 03 Given plant conditions and configuration 3.5 541 BOTH K3.03: 3.9 BANK CFR41.2/41.3/41.4/41.5 lineup, determine valid method for q060 A1.03: 3.3 NRC 6/01 determining Reactor coolant temperature. A1.06: 3.7 A1.08: 2.9 215003 IRM 01 Identify the power supply to the Intermediate 2.7 BOTH NEW CFR41.6 Range Nuclear Instrumentation. q044 219000 RHR /LPCI Suppression Pool 04 Describe the method used to control 2.9 BOTH A4.12: 4.1 NEW moved to Cooling Mode Suppression Pool Temperature and cooldown q061 K5 CFR41.7 rate.

234000 Fuel Handling Equipment 239003 MSIV Leakage Control 245000 Main Turbine Gen., and Auxiliaries 10 Discuss the basis for the limit on Main 2.6 44 BOTH K4.06: 2.8 BANK GGNS CFR41.4/41.10/43.5 Generator Reactive loading at GGNS. q063 A4.14: 2.5 NRC 3/98 generator has lower limits due to reverse power.

259001 Reactor Feedwater 01 Given a failure of the RFPT Lube Oil System, 3.7 BOTH K1.11: 2.7 BANK CFR41.4 identify the Reactor Feedwater System q056 K4.06: 2.6 response. K6.09: 2.9 PAGE 1 TOTAL TIER 2 GROUP 2 1 1 0 0 2 0 0 2 1 0 0 PAGE TOTAL # QUESTIONS 7 219000 random selection K2. Topic moved to K5 due to low discriminatory value of K2 topic for RHR Suppression Pool Cooling.

REVISION 0 4/29/2002 PAGE 9 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-1 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 2 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO/ RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # BOTH K/A 262002 UPS (AC/DC) 01 Concerning the ESF Static Inverters, identify 2.6 544 BOTH A2.01: 2.8 BANK CFR41.7/41.10/43.5 the correct response of the inverter on a loss of q066 A3.01: 3.1 NRC 6/01 normal power supply.

263000 DC Electrical Distribution 271000 Offgas 272000 Radiation Monitoring 01 Identify the normal radiation monitoring 3.2 BOTH NEW CFR41.10/41.11/43.4/43.5 alarms received on a reactor down power from q067 full power and their cause. (Hydrogen Water Chemistry) 286000 Fire Protection 08 Discern the response of the Diesel Driven Fire 3.3 3 BOTH K5.05: 3.1 BANK CFR41.4 Pumps on an auto start signal with a failure to q068 K4.07: 3.3 NRC 3/98 start. A3.01: 3.4 A4.06: 3.4 290003 Control Room HVAC 05 Identify the response of the Control Room 3.2 BOTH NEW CFR41.7/41.11/43.4 HVAC system upon receipt of radiation q070 alarms.

300000 Instrument Air 02 Describe the crosstie between Service and 3.0 BOTH BANK moved to CFR41.4/41.10/43.5 Instrument Air and when it occurs. q071 K3 400000 Component Cooling Water 01 Discuss the status of the standby feature of the 3.9 105 BOTH BANK CFR41.4 ESF powered CCW pump during a LOCA. q072 NRC 3/98 PAGE 2 TOTALS 0 0 0 2 0 0 3 1 0 0 0 PAGE 3 TOTAL # QUESTIONS 6 PAGE 1 TOTALS 1 1 0 0 2 0 0 2 1 0 0 PAGE 1 TOTAL # QUESTIONS 7 K/A CATEGORY TOTALS: 1 1 0 2 2 0 3 3 1 0 0 TIER 2 GROUP 2 GROUP POINT TOTAL 13 300000 random selection A4. Topic moved to K3 due to low discriminatory value of single subject in A4.

REVISION 0 4/29/2002 PAGE 10 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-1 AUGUST 2002 PLANT SYSTEMS - TIER 2 GROUP 3 SYSTEM #/NAME K K K K K K A A A A G TOPIC(S) IMP REC SRO/RO RELATED ORIGIN NOTES:

1 2 3 4 5 6 1 2 3 4 # / BOTH K/A 201003 Control Rod and Drive Mechanism 215001 Traversing In-core Probe 233000 Fuel Pool Cooling and Cleanup 239001 Main and Reheat Steam 256000 Reactor Condensate 08 Describe the effects of Low Pressure 3.1 294 BOTH A3.01: 2.7 BANK CFR41.4 Feedwater Heater isolation on the Condensate q064 A3.04: 3.0 NRC 4/00 System operation and plant operations. A3.07: 2.9 268000 Radwaste 02 Given conditions and a Liquid Radwaste 3.6 110 BOTH 272000 BANK Process Rad CFR41.13/43.4 discharge in progress, determine whether a q073 A3.03: 3.5 NRC 3/98 Monitor release will continue. isolation 288000 Plant Ventilation 05 Describe the response of the Auxiliary 3.6 BOTH K1.02: 3.4 NEW CFR41.7 Building Ventilation System on High q074 Radiation.

290002 Reactor Vessel Internals 05 Identify the consequence of exceeding a 4.2 BOTH K5.01: 3.9 NEW moved to CFR41.3/41.14/43.2 thermal limit. q075 A2 K/A CATEGORY TOTALS: 1 0 0 0 0 0 1 2 0 0 0 TIER 2 GROUP 3 GROUP POINT TOTAL 4 290002 random selection A1. Topic moved to A2 due to no subjects in A1.

REVISION 0 4/29/2002 PAGE 11 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE ES-401-5 AUGUST 2002 GENERIC KNOWLEDGE AND ABILITIES TIER 3 CATEGORY C1 C2 C3 C4 TOPIC(S) IMP REC # SRO/RO RELATED ORIGIN NOTES:

/BOTH K/A CONDUCT OF OPERATIONS - Deviations from 2.1.1 Given plant conditions that require a deviation from 3.8 135 SRO 2.1.2: 4.0 BANK 50.54x Approved documents in emergency conditions procedures, describe the process and when this action q084 NRC 12/00 CFR41.10/43.3/43.5 can be taken.

CONDUCT OF OPERATIONS - Shift Manning 2.1.4 Determine requirements for fire brigade and 3.4 486 SRO 2.4.26: 3.3 BANK CFR41.10/43.1/43.2/43.5 conditions when the fire brigade may be less than q085 NRC 6/01 required.

CONDUCT OF OPERATIONS - Chemistry Limits 2.1.34 Given plant conditions, procedures, and coolant 2.9 197 SRO BANK Use of EPRI CFR41.10/43.2/43.5 samples, determine actions to be taken based on q086 NRC 6/01 Guidelines vs Tech chemistry results. Specs.

CONDUCT OF OPERATIONS - Facility License 2.1.10 Given reactor thermal power, determine acceptability 3.9 SRO NEW CFR43.1 with regard to the Station Operating License. q087 CONDUCT OF OPERATIONS - Station Graphs 2.1.25 Given plant parameters and Tech Specs determine 3.1 SRO 2.1.33: 4.0 NEW SLC concentration CFR41.6/41.10/43.2/43.5 operability of Standby Liquid Control. q088 2.2.22: 4.1 graphs EQUIPMENT CONTROL - Refueling Procedures & 2.2.26 Describe the permissions required for all lifting on 3.7 702 SRO 2.2.27: 3.5 NEW Limitations the Containment Refuel Floor during Core q089 CFR41.10/43.6/43.7 Alterations.

EQUIPMENT CONTROL - Refueling 2.2.28 Given an evolution to be performed in Containment 3.5 411 SRO BANK CFR41.9/41.10/43.2/43.4/43.6/43.7 during fuel handling operations, evaluate the q090 NRC 12/00 allowances of the evolution.

EQUIPMENT CONTROL - Troubleshooting 2.2.20 Given a situation in the plant, apply the rules 3.3 492 SRO BANK CFR41.10/43.5 concerning MAIs vs allowed troubleshooting. q091 NRC 6/01 EQUIPMENT CONTROL - Core Alterations 2.2.32 Given a situation during refueling operations, apply 3.3 491 SRO 2.2.28: 3.5 BANK CFR43.6/43.7 the Criticality Rules. q092 NRC 6/01 RADIATION CONTROL - ALARA 2.3.2 Given a situation requiring independent verification 2.9 127 SRO 2.1.29: 3.3 BANK CFR41.12/43.4 in a high radiation area determine allowances for q093 2.2.11: 3.4 NRC 6/01 waiving verification based on ALARA and 2.2.13: 3.8 requirements for meeting verification.

RADIATION CONTROL - Radiation Work Permits 2.3.7 Given conditions and procedures, determine 3.3 SRO NEW CFR41.10/41.12/43.4/43.5 applicability of radiation work permits. q094 PAGE 1 TOTAL TIER 3 5 4 2 0 PAGE TOTAL # QUESTIONS 11 REVISION 0 4/29/2002 PAGE 12 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION BWR SRO EXAMINATION OUTLINE CONT. ES-401-5 AUGUST 2002 GENERIC KNOWLEDGE AND ABILITIES TIER 3 CATEGORY C1 C2 C3 C4 TOPIC(S) IMP REC # SRO/RO RELATED ORIGIN NOTES:

/BOTH K/A EMERGENCY PROCEDURES / PLAN - EOPs 2.4.20 Given conditions delineated in Caution 1 of the 4.0 415 SRO 2.4.18: 3.6 BANK EOP usage and usage EOPs, determine when EOP transition to q095 2.4.22: 4.0 NRC 12/00 CFR41.10/43.5 contingencies is required. 2.4.23: 3.8 EMERGENCY PROCEDURES / PLAN - SRO 2.4.40 Given a security threat taking over the Main Control 4.0 192 SRO 2.1.2: 4.0 BANK Control Room Responsibilities Security Threat Room, determine actions to be taken. q096 2.4.49: 4.0 NRC 4/00 abandonment CFR41.10/43.5 2.4.11: 3.6 (terrorist attack) 2.4.28: 3.3 EMERGENCY PROCEDURES / PLAN - EOPs 2.4.7 Given plant conditions and EOPs, determine a course 3.8 SRO 2.4.14: 3.9 NEW CFR41.10/43.5 of action to be taken. q097 EMERGENCY PROCEDURES / PLAN - 2.4.30 Given plant conditions, determine the reportability of 3.6 SRO NEW Reportability plant conditions and time requirements. q098 CFR41.10/43.5 EMERGENCY PROCEDURES / PLAN - SAPs 2.4.16 Given conditions requiring transition from EOPs to 4.0 SRO 2.4.8: 3.8 NEW EOP/SAP CFR41.10/43.5 SAPs, identify implementation and responsibilities q099 2.4.14: 3.9 for implementation.

EMERGENCY PROCEDURES / PLAN - 2.4.42 Given plant conditions, determine what emergency 3.7 SRO NEW E-Plan Emergency Response Facilities facilities are to be activated. q100 CFR41.10/43.5 PAGE 2 TOTAL TIER 3 0 0 0 6 PAGE TOTAL # QUESTIONS 6 PAGE 1 TOTAL TIER 3 5 4 2 0 PAGE TOTAL # QUESTIONS 11 K/A CATEGORY TOTALS: 5 4 2 6 TIER 3 GROUP POINT TOTAL 17 REVISION 0 4/29/2002 PAGE 13 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

Appendix D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 1 Op-Test No.: Day 1 Examiners: _________________________ Operators:__________________________

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Operate Standby Service Water B for chemical addition through all loads.
2. Respond to a failure of APRM A upscale.
3. Take actions in response to a Low Pressure Feedwater Heater 3B Tube leak. Complete actions of the Loss of Feedwater Heating ONEP.
4. Analyze the affects of a reduction of Main Condenser Vacuum on plant operations and take required actions.
5. Take actions per the EOPs in response to an ATWS and mitigate the consequences of the ATWS with no Main Steam Bypass Valves.
6. Respond to a failure of Division I ECCS to manually initiate via the Manual Initiation pushbutton.
7. Take actions for a failure of Standby Liquid Control to inject to the Reactor during an ATWS.

Initial Conditions: Reactor Power is at 100 %.

INOPERABLE Equipment APRM H is INOP due to a failed power supply card RHR C Pump is tagged out of service for motor oil replacement TBCW Pump C is tagged out of service for pump seal replacement Appropriate clearances and LCOs are written.

Turnover: The plant is operating at 100% power. Chemistry has requested SSW B be operated through all loads for a chemical addition.

There are scattered thundershowers reported in the Tensas Parish area.

REVISION 0 4/29/2002

Appendix D Scenario Outline Form ES-D-1 Scenario 1 Day 1 (Continued)

Event K/A Event Event 10CFR No. Type* Description 55.45(a) 1 4, 5, 6 2.1.30 N (BOP) Start SSW B and operate through all loads (SOI 04-1-01-P41-1 section 4.3) 2 3, 5 215005 A2.02 I (RO) Respond to APRM A failure upscale. Complete Technical Specification 2.1.12; 2.1.33 determinations.

3 2, 3, 4, 5, 2.4.49 R/C (RO, Respond to a tube failure in LP FW Heater 3B. Perform actions per ONEP 05-1 6 295014 BOP) V-5. Lower Reactor power with Recirc flow.

AA1.07; AA2.03 4 3, 4, 5, 6 2.4.49 C(RO, Recognize and respond to a loss of Main Condenser vacuum. Take actions per 295002 BOP) ONEP 05-1-02-V-8.

AA1.02; AA1.05; AA2.01 2, 3, 4, 7 2.4.4; 2.4.49 When required initiate a manual Reactor Scram.

295006 AA1.01; AA1.05; AA1.07 5 6, 8, 12, 295037 EA1.0; M (ALL) Upon Reactor Scram recognize the failure of all control rods to fully insert and take 13 EA2.0 actions per EOPs for ATWS.

203000 A3.08 241000 A4.06 209001 A4.05; I (BOP) Upon orders to initiate and override Low Pressure ECCS, recognize the failure of 3, 5 A3.01; A3.02 203000 A4.05; Division I to initiate via Manual Initiation pushbutton. Take actions upon automatic A3.01; A3.02; ; initiation to override Division I Low Pressure ECCS.

A2.14A3.08 3, 4, 8 295037 C (BOP) Recognize the failure of Standby Liquid Control to meet the parameters to inject into EA1.04; the Reactor when initiated and actions taken for Alternate Boron Injection.

EA1.10 211000 A1.0; A2.04; A3.0 All evolutions test 55.45(a)12 & 13.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor REVISION 0 4/29/2002

Critical Tasks

- Inject Standby Liquid Control prior to Suppression Pool Temperature reaching 110 °F.

- Identify the need for Alternate Standby Liquid Control injection.

- Terminate and prevent injection from Feedwater and ECCS when conditions require entry into Level/Power Control.

- Commence injection into the reactor using Feedwater or RHR A or B through Shutdown Cooling when reactor level reaches -192.

- Insert Control Rods in response to ATWS conditions.

REVISION 0 4/29/2002

Appendix D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 2 Op-Test No.: Day 2 Examiners: _________________________ Operators:__________________________

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Secure Diesel Generator 12 from diesel run.
2. Raise Reactor Power by withdrawing control rods.
3. Perform operator actions for a stuck control rod per ONEP.
4. Analyze a failure of Recirculation Pump B Seal # 2 failure.
5. Respond to a loss of Bus 12HE and trip of Recirculation Pump B per ONEPs.
6. Respond to a failure of Feedwater Line in the Drywell, initiate a reactor scram based on rising Drywell Pressure per EOPs.
7. Respond to a failure of Division 1 ECCS failure to initiate.
8. With a failure of Feedwater Line in the Drywell and reduced injection systems maintain reactor level per the EOPs.

Initial Conditions: Reactor Power is at 35 % bringing the plant up following an outage; Reactor Recirculation pumps are in Fast Slow Speed at 60 % core flow; a single Reactor Feed Pump in single element Master Level Control. Diesel Generator 12 operating at 2000KW load.

INOPERABLE Equipment APRM H is INOP due to a failed power supply card RHR C is tagged out of service for motor oil replacement TBCW Pump C is tagged out of service for pump seal replacement Appropriate clearances and LCOs are written.

Turnover: Secure Diesel Generator 12 from service. Leave Standby Service Water B in operation for chemistry. Then continue to bring the plant to full power per IOI-2. There are scattered thundershowers reported in the Tensas Parish area.

REVISION 0 4/29/2002

Appendix D Scenario Outline Form ES-D-1 Scenario 2 Day 2 (Continued)

Event K/A Event Event 10CFR No. Type* Description 55.45(a) 1 264000 A4.0; N Secure Diesel Generator 12 from operation 3, 4, 5 (BOP) (SOI 04-1-01-P75-1) 201005 A3.01; A3.02; 2 1, 2, 4, 5 A3.03; A4.01 R(RO) Withdraw control rods to raise power.

2.2.2 (Control Rod Pull Sheet & IOI 03-1-01-2) 3 1, 2, 3, 201001 C (RO, Control Rod 32-09 is stuck, un-stick control rod per ONEP. (ONEP 05-1-02-IV-1)

A4.04 5, 6, 8 2.4.4; 2.4.11; BOP) 2.4.48 4 202001 C (RO) Respond to a failure Seal # 2 of Recirculation Pump B. (Tech Specs) 3, 4, 7 A2.10; A4.10; A4.11 3, 4, 5, 6 202001 5 A2.03 C (RO, Respond to Overcurrent lockout on bus 12HE and trip of Recirculation Pump B.

BOP) (SOI 04-1-01-R21-12 & 05-1-02-III-3) 6 3, 4, 5, 295031 M (ALL) Feedwater Line B ruptures in the Drywell with leakage from the reactor.

EA1.0 6, 7, 13 203000 A3.08 241000 A4.06 3, 4, 7 2.4.4 I (BOP) Failure of Division 1 ECCS to automatically initiate on High Drywell Pressure 295024 EA1.0 3, 4, 5, 6 209002 C HPCS injection valve failure to open on initiation A2.03; A3.01; A4.03 (BOP)

All evolutions test 55.45(a) 12 & 13.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical Tasks

- Recognize failure of Division 1 to initiate and manually initiate Division 1

- Isolate Feedwater line B and reestablish feed through Feedwater line A or Lower reactor pressure to allow injection from low pressure systems REVISION 0 4/29/2002

Appendix D Scenario Outline Form ES-D-1 Facility: GRAND GULF NUCLEAR STATION Scenario No.: 3 Op-Test No.: BACKUP Examiners: _________________________ Operators:__________________________

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Raise Reactor Power using Recirculation Flow.
2. Start 3rd Condensate and Condensate Booster Pumps.
3. Respond to a trip of RPS Motor Generator B.
4. Determine the source and respond to a leak on the suction valve of RHR Pump B, EOP entry.
5. Respond to a steam leak in the Auxiliary Building Steam Tunnel and a failure of Group 1 to isolate.
6. Take actions per the EOPs in response to two stuck control rods following a Reactor Scram.
7. Take actions per EOPs to control RPV parameters with a failure the MSIVs to isolate the steam leak.

Initial Conditions: Reactor Power is at 83 % continuing power ascension to rated conditions.

INOPERABLE Equipment APRM H is INOP due to a failed power supply card RHR Pump C is tagged out of service for motor oil replacement TBCW Pump C is tagged out of service for pump seal replacement Appropriate clearances and LCOs are written.

Turnover: Continue power ascension. Radwaste is prepared for full Condensate and Feedwater operation. There are scattered thundershowers reported in the Tensas Parish area.

REVISION 0 4/29/2002

Scenario 3 BACKUP (Continued)

Event 10CFR K/A Event Event 55.45(a) Description No. Type*

1 1, 2, 4, 5, 202001 A4.04 R (RO) Raise Total Core Flow to >12.5 Mlbm/hr Feedwater Flow.

202002 A4.08 6, 8 2.2.2 (IOI 03-1-01-2) 2 2, 4, 5, 6 256000 A3.02; N (RO) Start 3rd Condensate and Condensate Booster Pump.

A4.01 (SOI 04-1-01-N19-1) 3 3, 5, 6 212000 A1.11; C (RO, Respond to trip of RPS Motor Generator B.

A2.01; A4.07 BOP) (ONEP 05-1-02-III-2) 4 3, 4, 5, 6 295036 EA1.02 C Determine the source and respond to a packing leak on E12-F004B RHR B Suction (BOP) Valve, with the valve failure determine unisolable and take actions per EOP - 3 & 4.

5 3, 4, 6, 2.4.46; 2.4.47; M (ALL) Recognize and respond to a steam leak in the Auxiliary Building Steam Tunnel.

2.4.48; 2.4.49 13 3, 4, 6, 2.4.46; 2.4.47; I (BOP) Recognize the failure of Group 1 to automatically isolate and take actions to isolate the 2.4.48; 2.4.49 13 290001 A2.06; Main Steam Lines (ONEP 05-1-01-III-5)

A4.04 3, 4, 6, 2.4.46; 2.4.47; Recognize the failure of a single Main Steam line to isolate and take actions for mitigation 2.4.48; 2.4.49 13 290001 A2.06; of the leak.

A4.04 4, 6, 12, 295037 EA1.0; C (RO) Recognize the failure of two control rods to fully insert on the Reactor Scram and take EA2.0 13 212000 A4.17 actions as necessary per procedures to insert the control rods.

All evolutions test 55.45(a) 12 & 13.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical Tasks

- Manually scram the reactor.

- Isolate the main steam lines.

REVISION 0 4/29/2002

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