ML030420312

Init Exam - 8/ 2002- Final Outlines

ML030420312
Person / Time
Site:	Grand Gulf
Issue date:	08/23/2002
From:	Gody A NRC Region 4
To:	Eaton W Entergy Operations
References
50-416/02-301 50-416/02-301
Download: ML030420312 (39)

Text

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 TOTAL 5

7 7

7 8

2 36 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 TOTAL 7

3 4

5 4

5 4

9 6

2 2

51 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

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 Topic/Subject Description Describe method of evaluation:

1. ONE Administrative JPM, OR

2. TWO Administrative Questions Knowledge

/ Ability IMP Additional K/As ORIGIN NOTES A.1 Plant Drawings JPM GJPM-OP-ADM21 Given a component, determine the fuse(s) to be removed to de-energize the component and how the component will fail.

2.1.25 2.8 BANK NRC 5/2000 Operator Responsibilities (Condition Reports)

JPM GJPM-OP-ADM42 Given a plant discrepancy, initiate a condition report.

2.1.2 3.0 NEW A.2 Protective Tagging JPM GJPM-OP-ADM41 Given an installed equipment clearance, perform the duties of an independent verifier.

2.2.13 3.6 2.1.2: 3.0 MOD NRC 6/01 HPCS changed to LPCS A.3 Radiation Control JPM GJPM-OP-ADM34 Perform required actions to access the Controlled Access Area (CAA), determine requirements to enter a Contamination Area, and exit the CAA.

2.3.1 2.6 MOD Changed to Contam.

Area NRC 6/01 was entry and exit with entry rqmts to high contam area CFR 55.45 (a)9 & 10 A.4 Emergency Plan Assessment JPM GJPM-OP-ADM25 Given initial notification form, perform notification of offsite agencies using the Operational Hotline.

2.4.43 2.8 2.4.39: 3.3 2.4.30: 3.6 BANK NRC 5/2000 CFR 55.45 (a) 11 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 Function Knowledge

/ Ability IMP.

Additional K/As ORIGIN NOTES B.1. CONTROL ROOM SYSTEMS

1. 241000 REACTOR / TURBINE PRESSURE REGULATING SYSTEM (N)(S)(L)(A) 3 A4.06 3.9 A1.07: 3.8 A2.03: 4.1 NEW Operate the Turbine Pressure Control System to Lower Reactor Pressure (Pressure Regulator fault requiring use of Manual Jack)

GJPM-RO-N3202 A3.08: 3.8 CFR 55.45(a)3, 4, 5 & 8

2. 204000 REACTOR WATER CLEANUP SYSTEM (RWCU) (D)(S)(L)(A) 2 A1.01 3.1 A1.04: 2.8 A1.06: 2.8 BANK Failure of G3-F234 Align RWCU for Vessel Level Control (Faulted

- Valve Failure)

GJPM-RO-G3301 A2.01: 3.2 NRC 5/2000 CFR 55.45(a)3

& 8

3. 295029 SUPPRESSION POOL LEVEL HIGH (D)(S) 5 EA1.02 3.1 209002 A3.01: 3.3 BANK CFR 55.45(a)8 Lower Suppression Pool Level with HPCS GJPM-RO-E2209 A4.01: 3.7 A4.02: 3.6 A4.14: 3.0 NRC 3/1998

4. 261000 STANDBY GAS TREATMENT SYSTEM (D)(S)(A) 9 A4.03 3.0 A4.02: 3.1 A4.09: 2.7 BANK CFR 55.45(a)9 Place SBGT Train in Standby with an Auto Start Signal Present (Faulted - High Rad)

GJPM-RO-T4801 NRC 6/2001

5. 295021 LOSS OF SHUTDOWN COOLING (ADHR)

(M)(S)(A) 4 AA1.04 3.7 205000 A4.01: 3.7 MOD CFR 55.45(a)5 Startup Alternate Decay Heat Removal in Reactor to Reactor Mode (E12-F042C fail on stroke)

GJPM-RO-E1231 A4.02: 3.6 A4.03: 3.6 A4.09: 3.1 A2.10: 2.9 Added Valve failure

& 8 NRC 5/2000 REVISION 0 4/29/2002

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 Function Knowledge

/ Ability IMP.

Additional K/As ORIGIN NOTES

6. 262001 AC ELECTRICAL DISTRIBUTION (N)(S) 6 A4.01 3.4 A4.02: 3.4 A4.04: 3.6 NEW CFR 55.45(a) 8 Transfer Electrical Loads from Service Transformer 21 to Service Transformer 11 GJPM-RO-R2730 A4.05: 3.3 2.1.31: 4.2 2.1.30: 3.9

7. 201005 ROD CONTROL AND INFORMATION SYSTEM (RCIS) (D)(C) 7 A2.03 3.2 A2.06: 3.2 295037 BANK CFR 55.45(a) 3 Defeat Rod Control and Information System per EP2 Attachment 20.

GJPM-RO-EP030 EA1.08: 3.6 295015 AA1.04: 3.4 NRC 5/2000

& 6 Also Safety Function 1 B.2. FACILITY WALK-THROUGH

8. 295016 CONTROL ROOM ABANDONMENT (D)(P)(R) 7 AA1.07 4.2 BANK CFR 55.45(a) 8 Align the Remote Shutdown Panel Alternate Shutdown Panels for a Fire in the Control Room GJPM-RO-C6108 NRC 6/2001 9, & 12 Emergency/

Abnormal

9. 201001 CONTROL ROD DRIVE (CRD) HYDRAULIC SYSTEM (D)(P)(R) 1 2.1.30 3.9 2.1.29: 3.4 BANK CFR 55.45(a) 8 Rotate the CRD Drive Water Filters GJPM-NLO-C1102

& 9

10. 286000 FIRE PROTECTION SYSTEM (D)(P)(A) 8 A4.06 3.4 2.1.20: 4.3 2.1.30: 3.9 BANK CFR 55.45(a) 6 Perform Local Start of Fire Pump Diesel after Failure to Start (Faulted)

GJPM-NLO-P6401 NRC 5/2000

• 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

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 Topic/Subject Description Describe method of evaluation:

1. ONE Administrative JPM, OR

2. TWO Administrative Questions Knowledge

/ Ability IMP Additional K/As ORIGIN NOTES A.1 Technical Specifications JPM GJPM-SRO-ADM43 Given a component, determine Limiting Condition for Operations and complete LCO documentation.

2.1.12 4.0 2.2.23: 3.8 BANK Different component Plant Safety Index (EOOS)

JPM GJPM-SRO-ADM46 Given plant conditions, determine the Plant Safety Index (EOOS).

2.1.19 3.0 BANK NRC 6/2001 Different component A.2 Protective Tagging JPM GJPM-SRO-ADM45 Given a clearance, perform the supervisory review of the clearance for adequacy and issue.

2.2.13 3.8 2.1.2: 4.0 NEW A.3 Radiation Control JPM GJPM-SRO-ADM44 Perform required actions to access the Controlled Access Area (CAA), determine requirements to enter a High Radiation Area in an Emergency and authorization required, and exit the CAA.

2.3.1 3.0 2.3.4: 3.1 MOD Changed to Planned Special Exposures CFR 55.45 (a)9 & 10 A.4 Emergency Plan Action Levels JPM GJPM-SRO-A&E40 Given conditions, determine the appropriate emergency classification and complete the required notification form.

2.4.41 4.1 2.4.30: 3.6 NEW CFR 55.45 (a) 11 REVISION 0 4/29/2002

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 Function Knowledge

/ Ability IMP.

Additional K/As ORIGIN NOTES B.1. CONTROL ROOM SYSTEMS

1. 295021 LOSS OF SHUTDOWN COOLING (ADHR)

(M)(S)(A) 4 AA1.04 3.7 205000 A4.01: 3.7 MOD CFR 55.45(a)5 Startup Alternate Decay Heat Removal in Reactor to Reactor Mode (E12-F042C fail on stroke)

GJPM-RO-E1231 A4.02: 3.6 A4.03: 3.6 A4.09: 3.1 A2.10: 2.9 Added Valve failure

& 8 NRC 5/2000

2. 262001 AC ELECTRICAL DISTRIBUTION (N)(S) 6 A4.01 3.7 A4.02: 3.4 A4.04: 3.7 NEW CFR 55.45(a) 8 Transfer Electrical Loads from Service Transformer 21 to Service Transformer 11 GJPM-RO-R2730 A4.05: 3.3 2.1.31: 3.9 2.1.30: 3.4 B.2. FACILITY WALK-THROUGH

3. 295016 CONTROL ROOM ABANDONMENT (D)(P)(R) 7 AA1.07 4.3 BANK CFR 55.45(a) 8 Align the Remote Shutdown Panel Alternate Shutdown Panels for a Fire in the Control Room GJPM-RO-C6108 NRC 6/2001 9, & 12 Emergency/

Abnormal

4. 201001 CONTROL ROD DRIVE (CRD)

HYDRAULIC SYSTEM (D)(P)(R) 1 2.1.30 3.4 2.1.29: 3.3 BANK CFR 55.45(a) 8 Rotate the CRD Drive Water Filters GJPM-NLO-C1102

& 9

5. 286000 FIRE PROTECTION SYSTEM (D)(P)(A) 8 A4.06 3.4 2.1.20: 4.2 2.1.30: 3.4 BANK CFR 55.45(a) 6 Perform Local Start of Fire Pump Diesel after Failure to Start (Faulted)

GJPM-NLO-P6401 NRC 5/2000

• 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

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 TOTAL 6

8 7

11 9

2 43 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 TOTAL 5

2 2

4 2

5 5

7 4

2 2

40 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

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 Topic/Subject Description Describe method of evaluation:

1. ONE Administrative JPM, OR

2. TWO Administrative Questions Knowledge

/ Ability IMP Additional K/As ORIGIN NOTES A.1 Plant Drawings JPM GJPM-OP-ADM21 Given a component, determine the fuse(s) to be removed to de-energize the component and how the component will fail.

2.1.25 2.8 BANK NRC 5/2000 Operator Responsibilities (Condition Reports)

JPM GJPM-OP-ADM42 Given a plant discrepancy, initiate a condition report.

2.1.2 3.0 NEW A.2 Protective Tagging JPM GJPM-OP-ADM41 Given an installed equipment clearance, perform the duties of an independent verifier.

2.2.13 3.6 2.1.2: 3.0 MOD NRC 6/01 HPCS changed to LPCS A.3 Radiation Control JPM GJPM-OP-ADM34 Perform required actions to access the Controlled Access Area (CAA), determine requirements to enter a Contamination Area, and exit the CAA.

2.3.1 2.6 MOD Changed to Contam.

Area NRC 6/01 was entry and exit with entry rqmts to high contam area CFR 55.45 (a)9 & 10 A.4 Emergency Plan Assessment JPM GJPM-OP-ADM25 Given initial notification form, perform notification of offsite agencies using the Operational Hotline.

2.4.43 2.8 2.4.39: 3.3 2.4.30: 3.6 BANK NRC 5/2000 CFR 55.45 (a) 11 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 Function Knowledge

/ Ability IMP.

Additional K/As ORIGIN NOTES B.1. CONTROL ROOM SYSTEMS

1. 241000 REACTOR / TURBINE PRESSURE REGULATING SYSTEM (N)(S)(L)(A) 3 A4.06 3.9 A1.07: 3.8 A2.03: 4.1 NEW Operate the Turbine Pressure Control System to Lower Reactor Pressure (Pressure Regulator fault requiring use of Manual Jack)

GJPM-RO-N3202 A3.08: 3.8 CFR 55.45(a)3, 4, 5 & 8

2. 204000 REACTOR WATER CLEANUP SYSTEM (RWCU) (D)(S)(L)(A) 2 A1.01 3.1 A1.04: 2.8 A1.06: 2.8 BANK Failure of G3-F234 Align RWCU for Vessel Level Control (Faulted

- Valve Failure)

GJPM-RO-G3301 A2.01: 3.2 NRC 5/2000 CFR 55.45(a)3

& 8

3. 295029 SUPPRESSION POOL LEVEL HIGH (D)(S) 5 EA1.02 3.1 209002 A3.01: 3.3 BANK CFR 55.45(a)8 Lower Suppression Pool Level with HPCS GJPM-RO-E2209 A4.01: 3.7 A4.02: 3.6 A4.14: 3.0 NRC 3/1998

4. 261000 STANDBY GAS TREATMENT SYSTEM (D)(S)(A) 9 A4.03 3.0 A4.02: 3.1 A4.09: 2.7 BANK CFR 55.45(a)9 Place SBGT Train in Standby with an Auto Start Signal Present (Faulted - High Rad)

GJPM-RO-T4801 NRC 6/2001

5. 295021 LOSS OF SHUTDOWN COOLING (ADHR)

(M)(S)(A) 4 AA1.04 3.7 205000 A4.01: 3.7 MOD CFR 55.45(a)5 Startup Alternate Decay Heat Removal in Reactor to Reactor Mode (E12-F042C fail on stroke)

GJPM-RO-E1231 A4.02: 3.6 A4.03: 3.6 A4.09: 3.1 A2.10: 2.9 Added Valve failure

& 8 NRC 5/2000 REVISION 0 4/29/2002

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 Function Knowledge

/ Ability IMP.

Additional K/As ORIGIN NOTES

6. 262001 AC ELECTRICAL DISTRIBUTION (N)(S) 6 A4.01 3.4 A4.02: 3.4 A4.04: 3.6 NEW CFR 55.45(a) 8 Transfer Electrical Loads from Service Transformer 21 to Service Transformer 11 GJPM-RO-R2730 A4.05: 3.3 2.1.31: 4.2 2.1.30: 3.9

7. 201005 ROD CONTROL AND INFORMATION SYSTEM (RCIS) (D)(C) 7 A2.03 3.2 A2.06: 3.2 295037 BANK CFR 55.45(a) 3 Defeat Rod Control and Information System per EP2 Attachment 20.

GJPM-RO-EP030 EA1.08: 3.6 295015 AA1.04: 3.4 NRC 5/2000

& 6 Also Safety Function 1 B.2. FACILITY WALK-THROUGH

8. 295016 CONTROL ROOM ABANDONMENT (D)(P)(R) 7 AA1.07 4.2 BANK CFR 55.45(a) 8 Align the Remote Shutdown Panel Alternate Shutdown Panels for a Fire in the Control Room GJPM-RO-C6108 NRC 6/2001 9, & 12 Emergency/

Abnormal

9. 201001 CONTROL ROD DRIVE (CRD) HYDRAULIC SYSTEM (D)(P)(R) 1 2.1.30 3.9 2.1.29: 3.4 BANK CFR 55.45(a) 8 Rotate the CRD Drive Water Filters GJPM-NLO-C1102

& 9

10. 286000 FIRE PROTECTION SYSTEM (D)(P)(A) 8 A4.06 3.4 2.1.20: 4.3 2.1.30: 3.9 BANK CFR 55.45(a) 6 Perform Local Start of Fire Pump Diesel after Failure to Start (Faulted)

GJPM-NLO-P6401 NRC 5/2000

• 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

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 Topic/Subject Description Describe method of evaluation:

1. ONE Administrative JPM, OR

2. TWO Administrative Questions Knowledge

/ Ability IMP Additional K/As ORIGIN NOTES A.1 Technical Specifications JPM GJPM-SRO-ADM43 Given a component, determine Limiting Condition for Operations and complete LCO documentation.

2.1.12 4.0 2.2.23: 3.8 BANK Different component Plant Safety Index (EOOS)

JPM GJPM-SRO-ADM46 Given plant conditions, determine the Plant Safety Index (EOOS).

2.1.19 3.0 BANK NRC 6/2001 Different component A.2 Protective Tagging JPM GJPM-SRO-ADM45 Given a clearance, perform the supervisory review of the clearance for adequacy and issue.

2.2.13 3.8 2.1.2: 4.0 NEW A.3 Radiation Control JPM GJPM-SRO-ADM44 Perform required actions to access the Controlled Access Area (CAA), determine requirements to enter a High Radiation Area in an Emergency and authorization required, and exit the CAA.

2.3.1 3.0 2.3.4: 3.1 MOD Changed to Planned Special Exposures CFR 55.45 (a)9 & 10 A.4 Emergency Plan Action Levels JPM GJPM-SRO-A&E40 Given conditions, determine the appropriate emergency classification and complete the required notification form.

2.4.41 4.1 2.4.30: 3.6 NEW CFR 55.45 (a) 11 REVISION 0 4/29/2002

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 Function Knowledge

/ Ability IMP.

Additional K/As ORIGIN NOTES B.1. CONTROL ROOM SYSTEMS

1. 295021 LOSS OF SHUTDOWN COOLING (ADHR)

(M)(S)(A) 4 AA1.04 3.7 205000 A4.01: 3.7 MOD CFR 55.45(a)5 Startup Alternate Decay Heat Removal in Reactor to Reactor Mode (E12-F042C fail on stroke)

GJPM-RO-E1231 A4.02: 3.6 A4.03: 3.6 A4.09: 3.1 A2.10: 2.9 Added Valve failure

& 8 NRC 5/2000

2. 262001 AC ELECTRICAL DISTRIBUTION (N)(S) 6 A4.01 3.7 A4.02: 3.4 A4.04: 3.7 NEW CFR 55.45(a) 8 Transfer Electrical Loads from Service Transformer 21 to Service Transformer 11 GJPM-RO-R2730 A4.05: 3.3 2.1.31: 3.9 2.1.30: 3.4 B.2. FACILITY WALK-THROUGH

3. 295016 CONTROL ROOM ABANDONMENT (D)(P)(R) 7 AA1.07 4.3 BANK CFR 55.45(a) 8 Align the Remote Shutdown Panel Alternate Shutdown Panels for a Fire in the Control Room GJPM-RO-C6108 NRC 6/2001 9, & 12 Emergency/

Abnormal

4. 201001 CONTROL ROD DRIVE (CRD)

HYDRAULIC SYSTEM (D)(P)(R) 1 2.1.30 3.4 2.1.29: 3.3 BANK CFR 55.45(a) 8 Rotate the CRD Drive Water Filters GJPM-NLO-C1102

& 9

5. 286000 FIRE PROTECTION SYSTEM (D)(P)(A) 8 A4.06 3.4 2.1.20: 4.2 2.1.30: 3.4 BANK CFR 55.45(a) 6 Perform Local Start of Fire Pump Diesel after Failure to Start (Faulted)

GJPM-NLO-P6401 NRC 5/2000

• 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

GRAND GULF NUCLEAR STATION AUGUST 2002 BWR RO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 ES-401-2 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO

/BOTH REC RELATED K/A ORIGIN NOTES:

295005 Main Turbine Generator Trip / 3 CFR41.5/41.6 05 Given a spurious Main Turbine Generator trip, determine the initial affect on Reactor Power and Reactor Pressure.

3.8 553 q 001 BOTH AA2.04: 3.7 AA2.03: 3.1 BANK NRC 6/01 295006 SCRAM / 1 CFR41.6/41.10/43.5 02 Describe the position of control rods following a reactor scram and how position is determined.

4.3 10 q002 BOTH 201005 A3.02: 3.5 A4.02: 3.7 201003 K4.05: 3.2 A3.01: 3.7 A4.02: 3.5 BANK NRC 6/01 Rx SCRAM Immediate Actions 295007 High Reactor Pressure / 3 CFR41.5 01 Describe the response of the Turbine Pressure Control System on an increasing reactor pressure.

3.5 69 q003 BOTH 241000 K4.01: 3.8 A2.02: 3.7 BANK NRC 6/01 295009 Low Reactor Water Level / 2 CFR41.5/43.5 03 Given plant conditions, identify the status of Recirculation pumps originally operating in fast speed with a lowering water level.

3.1 121 q004 BOTH AK1.02: 3.0 AA1.03: 3.0 BANK NRC 3/98 295010 High Drywell Pressure / 5 CFR41.5 04 Given plant parameters and elevated in-leakage into the Drywell, determine the status of reactor coolant system integrity.

3.5 286 q005 BOTH AK3.05: 3.5 2.4.21: 3.7 BANK NRC 4/00 GGNS Drywell air leak 295014 Inadvertent Reactivity Addition / 1 CFR41.1/41.2/41.6/43.6 01 With the reactor in startup conditions such that the reactor has dropped subcritical, what are the operator actions if a high worth control rod is withdrawn fully.

4.1 204 q006 BOTH BANK NRC 6/01 Susquehanna reactivity event 7/98 295015 Incomplete SCRAM / 1 CFR41.6/43.6 01 Describe the method to be used to allow the insertion of control rods using RCIS during an ATWS.

3.4 203 q007 BOTH BANK NRC 6/01 295024 High Drywell Pressure / 5 CFR41.9/41.10/43.5 14 Given conditions, determine ability to initiate Containment Spray per EOPs.

3.9 601 q008 BOTH NEW High Drywell pressure is EOP entry for CTMT Press Control 295025 High Reactor Pressure / 3 CFR41.3/43.2 05 State the Reactor Vessel pressure Safety Limit and its basis.

4.4 30 q076 RO EK1.02: 4.1 2.2.22: 3.4 2.2.25: 2.5 BANK NRC 12/00 295031 Reactor Low Water Level / 2 CFR41.2/41.3/41.10/43.5 01 Given plant conditions and a low reactor water level, determine core cooling mechanism and adequacy.

4.6 309 q009 BOTH 2.1.1: 3.7 2.4.21: 3.7 BANK 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 CONT.

ES-401-2 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO/

BOTH REC RELATED K/A ORIGIN NOTES:

295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 CFR41.1/41.2/41.6/43.6 04 Determine when plant conditions allow the termination of Standby Liquid Control with multiple control rods stuck out and what is the basis.

4.5 216 q010 BOTH EK1.04: 3.4 EK1.05: 3.4 EA2.03: 4.3 BANK NRC 4/00 500000 High Containment Hydrogen Conc. / 5 CFR41.10/43.5 2.

1.

20 Given applicable SOP, graph, and plant conditions, determine the power settings for the Hydrogen Recombiners and time to full power.

4.3 219 q011 BOTH EA1.03: 3.4 2.1.25: 2.8 BANK NRC 4/00 moved to Generics 295014 Inadvertent Reactivity Addition / 1 CFR41.6/43.6 02 Determine effects on the reactor of the fast opening a Recirc Flow Control Valve.

3.6 116 q012 BOTH 202002 K1.02: 4.2 K3.02: 4.0 BANK NRC 3/98 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 ES-401-2 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO/

BOTH REC RELATED K/A ORIGIN NOTES:

295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 CFR41.5/41.10/43.5 2.

4.

11 Given plant conditions and the power to flow map, determine the actions to be taken.

3.4 602 q013 BOTH 2.4.1: 4.3 AA2.01: 3.5 MOD NRC 4/00 moved to Generics 295002 Loss of Main Condenser Vacuum / 3 CFR41.4/43.4 04 Describe the basis for the isolation of the Main Steam Isolation Valves on a loss of condenser vacuum.

3.3 220 q015 BOTH AK3.05: 3.4 BANK NRC 4/00 295003 Partial or Complete Loss of AC Power/ 6 CFR41.7 03 Given a lockout on BOP Transformer 12B, determine the configuration of the AC Distribution System.

3.7 507 q014 BOTH AA1.01: 3.7 BANK NRC 6/01 295004 Partial or Complete Loss of DC Power / 6 295008 High Reactor Water Level / 2 CFR41.4/41.5 01 Identify the affects of a high Reactor Water Level on the Main Turbine and Reactor Feed Pump Turbines.

3.0 275 q016 BOTH 245000 A3.01: 3.6 259001 K6.07: 3.8 BANK NRC 6/01 295011 High Containment Temperature / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Water Temp. / 5 CFR41.5/41.10/43.2/43.5 01 During a surveillance operating RCIC, determine how often Suppression Pool Temperature is required to be monitored and the threshold for alternate actions.

3.8 q018 BOTH AA1.02: 3.9 2.1.33: 3.4 NEW 295016 Control Room Abandonment / 7 CFR41.5/41.10/43.5 02 Given parameters from the Remote Shutdown Panel indications, determine actual and Narrow Range RPV level.

4.2 603 q017 BOTH 2.1.25: 2.8 2.4.11: 3.4 MOD NRC 6/01 295017 High Offsite Release Rate / 9 CFR41.11/41.13/43.4 08 Given plant conditions with MSIV Leakage Control System operating, determine the mechanism for monitoring radiological release to locations outside the Auxiliary Building.

3.1 q019 BOTH AA1.09: 3.6 NEW 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 CONT.

ES-401-2 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO/

BOTH REC RELATED K/A ORIGIN NOTES:

295018 Partial or Complete Loss of CCW / 8 CFR41.4/41.10/43.5 03 Given a partial loss of Component Cooling Water, determine actions to be taken and their basis.

3.1 314 q020 BOTH BANK NRC 12/00 295019 Partial or Complete Loss of Inst. Air / 8 CFR41.4/41.10/43.5 14 Given a reduction in Instrument Air Header pressure, determine a possible cause.

3.2 q021 BOTH NEW New Air Dryers moved to AK2 295020 Inadvertent Cont. Isolation / 5 & 7 CFR41.4/41.7/41.9 03 Given a loss of Instrument Air System pressure to the Auxiliary Building, determine the effects on the ability to remove heat from the Containment.

3.1 q022 BOTH AK3.03: 3.2 NEW 295022 Loss of CRD Pumps / 1 CFR41.5 04 Describe the effects on reactor water level during a reactor startup with minimal decay heat and a loss of CRD Pumps. (RWCU is lined up to blowdown to the main condenser to compensate for CRD flow.)

2.5 55 q023 BOTH AK2.04: 2.5 BANK NRC 6/01 295026 Suppression Pool High Water Temp. / 5 CFR41.7/41.9/41.10/41.14/43.5 01 Given plant conditions, determine when Suppression Pool Cooling is no longer effective and alternate actions are required.

3.9 301 q024 BOTH BANK NRC 12/00 295027 High Containment Temperature / 5 CFR41.9/41.10/43.2 03 Given plant conditions, determine the Technical Specification Bases for shutting down the Reactor due to a high Containment Temperature.

3.7 512 q025 BOTH 2.2.25: 2.5 BANK NRC 6/01 295028 High Drywell Temperature / 5 CFR41.4/41.7 03 Given rising temperatures in the Reactor lower cavity area, determine the response of the Drywell Cooling system.

3.9 q26 BOTH EK2.04: 3.6 EK3.04: 3.6 NEW 295029 High Suppression Pool Water Level / 5 CFR41.9/41.10 01 Identify the bases for Emergency RPV Depressurization when Suppression Pool Level cannot be maintained below 24.4 feet.

3.4 513 q027 BOTH BANK NRC 6/01 295030 Low Suppression Pool Water Level / 5 CFR41.9/41.10/43.5 02 Evaluate Suppression Pool Temperature, with a Low Suppression Pool Level.

3.9 8

q028 BOTH BANK NRC 6/01 Caution 2 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 CONT.

ES-401-2 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO/

BOTH REC RELATED K/A ORIGIN NOTES:

295033 High Secondary Containment Area Radiation Levels / 9 CFR41.12/43.4 03 Given operations during a Refueling outage, determine the cause of elevated radiation levels in the area of the Fuel Pool Cooling Heat Exchangers.

3.7 q029 BOTH NEW Fuel Pool HX in an open area of Sec CTMT 295034 Secondary Containment Ventilation High Radiation / 9 295038 High Offsite Release Rate / 9 CFR41.10/41.12/43.4/43.5 02 Given plant conditions and procedures, determine the protective action recommendations to be recommended to the state and local officials in an emergency.

4.2 q030 BOTH BANK 600000 Plant Fire On Site / 8 CFR41.10/43.5 04 Given a fire at the Hydrogen Bulk Storage Facility describe the actions to be taken to combat the fire.

2.8 q031 BOTH 2.4.25: 2.9 2.1.32: 3.4 NEW moved to 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 3 ES-401-2 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO

/BOTH REC RELATED K/A ORIGIN NOTES:

295021 Loss of Shutdown Cooling / 4 CFR41.5/41.10/43.5 01 Given plant parameters and the graphs from the Inadequate Decay Heat Removal ONEP, determine time to boil.

3.5 604 q032 BOTH AK1.01: 3.6 MOD NRC 3/98 Multiple graphs for various conditions.

295023 Refueling Accidents / 8 295032 High Secondary Containment Area Temperature / 5 CFR41.4/41.10/43.5 03 Determine the systems affected by high temperatures in the RHR A Pump Room.

3.8 48 q033 BOTH EK3.07: 3.6 219000 A1.08: 3.7 A2.14: 4.1 BANK NRC 3/98 Steam leak at GGNS in Steam Condensing piping 295035 Secondary Containment High Differential Pressure / 5 CFR41.4/41.7 02 Describe the operation of the Standby Gas Treatment System with regard to Auxiliary Building and Enclosure Building Pressures.

3.8 q034 BOTH BANK 295036 Secondary Containment High Sump/Area Water Level / 5 CFR41.4/41.10/43.5 03 Given a rising water level in an ECCS Pump Room, determine appropriate actions to be taken with regard to the overall plant operation.

3.4 58 q035 BOTH EK3.03: 3.5 EA2.02: 3.1 BANK NRC 3/98 EOP-4 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 1 ES-401-2 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

201001 CRD Hydraulic CFR41.5/41.6 02 During a reactor scram with scram signals not reset, determine the CRD System flow rates.

2.6 q036 BOTH MOD 201005 RCIS CFR41.6/43.6 01 Given a failure of the Turbine First Stage pressure signal to RCIS, determine the mode of control rod movement.

3.2 q037 BOTH NEW 202002 Recirculation Flow Control CFR41.6 01 Given plant conditions following an actuation of EOC-RPT, determine the Recirculation Pump circuit breaker configuration.

3.4 q038 BOTH 202001 A2.15: 3.7 NEW EOC-RPT changes 2002 203000 RHR/LPCI: Injection Mode CFR41.8 17 State the basis for monitoring reactor pressure when aligning the RHR system for injection into the vessel for the LPCI mode.

4.0 60 q039 BOTH K4.01: 4.2 K4.02: 3.3 A3.01: 3.8 A3.08: 4.1 A4.08: 4.3 BANK NRC 6/01 Where is pressure sensed on LPCI for operation of the injection valve 209001 LPCS CFR41.7 02 Describe the method of operation of the ADS logic given LPCS and LPCI A out of service during a LOCA.

3.8 q040 BOTH 218000 K6.01:

3.9 K6.02: 4.1 NEW moved to K3 209002 HPCS CFR41.7/41.8 01 Given a spurious initiation of the HPCS system, determine the effect on Reactor water level.

3.9 519 q041 BOTH 259002 A2.08: 4.5 BANK NRC 6/01 211000 SLC CFR41.1/41.6/41.7/43.6 04 Given the initiation of Standby Liquid Control in an ATWS, discern the parameters indicating injection to the reactor.

4.5 q042 BOTH A4.01: 3.9 A4.03: 4.1 A4.06: 3.9 A4.07: 3.6 NEW 212000 RPS CFR41.6 01 Describe the response of the RPS Power System upon an ESF inverter loss.

3.6 102 q043 BOTH K1.04: 3.4 BANK NRC 3/98 215003 IRM CFR41.6 01 Identify the power supply to the Intermediate Range Nuclear Instrumentation.

2.5 q044 BOTH NEW 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

ES-401-2 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

215004 Source Range Monitor CFR41.5/41.6 2.

1.

32 Determine the conditions that would allow the operation of Source Range Monitor detector drives.

3.4 q046 BOTH A4.04: 3.2 moved to generics 215005 APRM / LPRM CFR41.6/41.7 03 Determine the affects on APRMs with reduced LPRM inputs.

3.1 326 q045 BOTH BANK NRC 12/00 216000 Nuclear Boiler Instrumentation CFR41.5 23 Determine the ability of the Recirculation Pumps to start based on Reactor Temperature.

3.3 122 q047 BOTH 202001 A4.01: 3.7 BANK NRC 3/98 217000 RCIC CFR41.5/41.7/41.10 01 Given plant conditions, determine the operation of RCIC and indications of injection.

3.7 328 q048 BOTH BANK NRC 12/00 218000 ADS CFR41.7 03 Given plant conditions determine automatic operation of Automatic Depressurization.

3.8 q049 BOTH NEW 223001 Primary CTMT and Auxiliaries CFR41.7/41.8 08 Given plant conditions and electrical busses that are unavailable, determine which components are available.

2.7 528 q050 BOTH K2.09: 2.7 K2.10: 2.7 BANK NRC 6/01 223002 PCIS / Nuclear Steam Supply Shutoff CFR41.7/41.9 09 Given plant conditions, evaluate the systems that should actuate or isolate.

3.6 385 q051 BOTH 2.4.21: 3.7 2.4.4: 4.0 BANK NRC 12/00 239002 SRVs CFR41.3 05 Describe the operation of the Safety Relief Valves in different modes of operation.

(system air pressure or reactor pressure) 3.1 337 q052 BOTH BANK NRC 12/00 241000 Reactor / Turbine Pressure Regulator CFR41.5 11 Describe the response of the plant with a failure of the Main Stop and Control Valves closed with the reactor at power.

3.4 244 q053 BOTH A1.01: 3.9 A1.02: 4.1 A1.07: 3.8 BANK NRC 4/00 259001 Reactor Feedwater CFR41.5/41.10/43.5 03 Given a set of plant conditions, determine the operational limitations for the Reactor Feed Pump Turbines 2.8 548 q077 RO BANK NRC 6/01 259002 Reactor Water Level Control CFR41.5 10 Describe the operation of the Feedwater Level Control System on a failure of the Feed Flow input signal.

3.4 273 q078 RO K6.04: 3.1 BANK NRC 4/00 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

ES-401-2 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

261000 SGTS CFR41.7/41.11 01 Given conditions determine the automatic operation of Standby Gas Treatment.

3.7 q054 BOTH NEW 264000 EDGs CFR41.4/41.7 04 Identify the signal that will automatically start the associated Standby Service Water System and align it to the Diesel Generator.

3.2 q079 RO NEW 217000 RCIC CFR41.7 01 Determine the indications of RCIC following overspeed trip and attempting to open the trip throttle valve.

3.5 676 q080 RO A3.02: 3.6 MOD Aud 6/01 202002 Recirculation Flow Control CFR41.6 01 Given plant conditions, identify the plant response to a Recirc Flow Control Runback.

3.6 235 q055 BOTH BANK NRC 4/00 Recirc Pumps in abnormal pump configuration 239002 SRVs CFR41.7 06 Describe the purpose of the SRV tailpipe vacuum breakers.

2.7 q081 RO 2.1.28: 3.2 NEW 259001 Reactor Feedwater CFR41.4 01 Given a failure of the RFPT Lube Oil System, identify the Reactor Feedwater System response.

3.7 q056 BOTH K1.11: 2.7 K4.06: 2.5 K6.09: 2.8 BANK 259002 Reactor Water Level Control CFR41.4/41.7 03 Describe the effects on the Reactor Water Level Control System from a failure of the Narrow Range Reactor Water level signal.

3.8 q057 BOTH BANK 264000 EDGs CFR41.8 2.

4.

4 During degraded grid conditions, determine the response of the diesel generators.

4.0 11 q058 BOTH K4.05: 3.2 A3.05: 3.4 BANK NRC 6/01 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 2 ES-401-2 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO

/ BOTH RELATED K/A ORIGIN NOTES:

201003 Control Rod and Drive Mechanism CFR41.6/41.10/43.5 08 Given conditions of the HCU, determine its status.

3.8 341 q082 RO BANK NRC 12/00 202001 Recirculation CFR41.3/41.5 10 Given plant conditions in regard to Reactor Recirc. Pump seals, determine the failed mechanism.

3.5 540 q059 BOTH A1.09: 3.3 A1.10: 2.6 BANK NRC 6/01 204000 RWCU CFR41.4 02 During a reactor startup and heatup, determine the effects on reactor water level with a loss of Reactor Water Cleanup operation.

3.1 q083 RO NEW 205000 Shutdown Cooling CFR41.2/41.3/41.4/41.5 03 Given plant conditions and configuration lineup, determine valid method for determining Reactor coolant temperature.

3.4 541 q060 BOTH K3.03: 3.8 A1.03: 3.3 A1.06: 3.7 A1.08: 3.1 BANK NRC 6/01 219000 RHR /LPCI Suppression Pool Cooling Mode CFR41.7 04 Describe the method used to control Suppression Pool Temperature and cooldown rate.

2.9 q061 BOTH A4.12: 4.1 NEW moved to K5 226001 RHR/LPCI: CTMT Spray Mode CFR41.7/41.8 07 Determine conditions that would result in automatic initiation of RHR Containment Spray mode.

3.5 q062 BOTH A3.01: 3.0 BANK moved to A3 239001 Main and Reheat Steam CFR41.4/41.14 03 Following an isolation of the Main Steam Isolation Valves, describe the operation of the Condensate and Feedwater Systems.

3.2 q084 RO K1.22: 3.1 NEW 245000 Main Turbine Gen. and Auxiliaries CFR41.4/41.10/43.5 10 Discuss the basis for the limit on Main Generator Reactive loading at GGNS.

2.5 44 q063 BOTH K4.06: 2.7 A4.14: 2.5 BANK NRC 3/98 GGNS 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 2 CONT.

ES-401-2 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

256000 Reactor Condensate CFR41.4 08 Describe the effects of Low Pressure Feedwater Heater isolation on the Condensate System operation and plant operations.

3.1 294 q064 BOTH A3.01: 2.7 A3.04: 3.0 A3.07: 2.9 BANK NRC 4/00 262001 AC Electrical Distribution CFR41.4/41.7 04 Given a loss of offsite power and plant conditions, determine the type of load sequencing to occur on ESF busses.

3.4 q065 BOTH MOD 262002 UPS (AC/DC)

CFR41.7/41.10/43.5 01 Concerning the ESF Static Inverters, identify the correct response of the inverter on a loss of normal power supply.

2.8 544 q066 BOTH A2.01: 2.6 A1.01: 2.4 BANK NRC 6/01 263000 DC Electrical Distribution CFR41.4 01 Identify the power supply to the DC Turbine Building Cooling Water Pump and its start signal.

3.1 q085 RO NEW moved to K2 Plant modification 271000 Offgas CFR41.7/41.13 09 Determine the startup sequence of the Offgas system to prevent explosive mixtures from forming.

3.3 q086 RO 2.1.32: 3.4 NEW 272000 Radiation Monitoring CFR41.10/41.11/43.4/43.5 01 Identify the normal radiation monitoring alarms received on a reactor down power from full power and their cause. (Hydrogen Water Chemistry) 3.2 q067 BOTH NEW 286000 Fire Protection CFR41.4 08 Discern the response of the Diesel Driven Fire Pumps on an auto start signal with a failure to start.

3.2 3

q068 BOTH K5.05: 3.0 K4.07: 3.3 A3.01: 3.4 A4.06: 3.4 BANK NRC 3/98 290001 Secondary CTMT CFR41.9 09 Describe the ability of Auxiliary Building Fire Protection system to be restored following Auxiliary Building isolation in conjunction with a loss of AC power.

3.4 264 q069 BOTH A2.06: 3.7 286000 A2.09: 2.7 BANK NRC 4/00 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 2 CONT.

ES-401-2 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

290003 Control Room HVAC CFR41.7/41.11/43.4 05 Identify the response of the Control Room HVAC system upon receipt of radiation alarms.

3.2 NEW q070 BOTH 300000 Instrument Air CFR41.4/41.10/43.5 02 Describe the crosstie between Service and Instrument Air and when it occurs.

3.0 q071 BOTH BANK moved to K3 400000 Component Cooling Water CFR41.4 01 Discuss the status of the standby feature of the ESF powered CCW pump during a LOCA.

3.4 105 q072 BOTH BANK 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 AUGUST 2002 BWR RO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 3 ES-401-2 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO

/ BOTH RELATED K/A ORIGIN NOTES:

215001 Traversing In-core Probe 233000 Fuel Pool Cooling and Cleanup CFR41.7 07 Identify the Tech Spec Limits on Spent Fuel Pool Temperature.

3.0 q087 RO NEW 234000 Fuel Handling Equipment 239003 MSIV Leakage Control 268000 Radwaste CFR41.13/43.4 02 Given conditions and a Liquid Radwaste discharge in progress, determine whether a release will continue.

2.6 110 q073 BOTH 272000 A3.03: 3.0 BANK NRC 3/98 Process Rad Monitor isolation 288000 Plant Ventilation CFR41.7 05 Describe the response of the Auxiliary Building Ventilation System on High Radiation.

3.3 q074 BOTH K1.02: 3.4 NEW 290002 Reactor Vessel Internals CFR41.3/41.14/43.2 05 Identify the consequence of exceeding a thermal limit.

3.7 q075 BOTH K5.01: 3.5 NEW moved to 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

REVISION 0 4/29/2002 PAGE 14 OF 14 NUREG 1021, REVISION 8 SUPPLEMENT 1 GRAND GULF NUCLEAR STATION AUGUST 2002 BWR RO EXAMINATION OUTLINE GENERIC KNOWLEDGE AND ABILITIES TIER 3 ES-401-5 CATEGORY C1 C2 C3 C4 TOPIC(S)

IMP REC #

SRO/RO

/BOTH RELATED K/A ORIGIN NOTES:

CONDUCT OF OPERATIONS - Mode of Operation CFR41.10/43.1 2.1.22 Given plant conditions determine the Tech Spec Operational Mode of the plant.

2.8 585 q088 RO BANK NRC 6/01 CONDUCT OF OPERATIONS - Conduct of Operations Requirements CFR41.10/43.5 2.1.1 Given components, determine components that are CONTROLS of the facility.

3.7 q089 RO NEW CONDUCT OF OPERATIONS - Shift Turnover CFR41.10/43.5 2.1.3 Given a change of operators at other than normal shift turnover, identify requirements for relief.

3.0 q090 RO NEW CONDUCT OF OPERATIONS - Component purpose CFR41.4 2.1.28 Identify the purpose for injecting Oxygen into the Condensate System.

3.2 q091 RO NEW EQUIPMENT CONTROL - Configuration Control CFR41.10/43.5 2.2.11 Given a component temporarily out of normal alignment per system operating instructions, determine the tracking mechanism to be employed.

2.5 387 q092 RO BANK NRC 12/00 SOER 98-1 Safety System Status Control EQUIPMENT CONTROL - Protective Tagging CFR41.10/43.5 2.2.13 Identify the proper method of tagging an air-operated component for system isolation.

3.6 q093 RO NEW RADIATION CONTROL - Radiation Reduction CFR41.12/43.4 2.3.10 Identify methods of reducing radiation levels in work areas.

2.9 q094 RO NEW RADIATION CONTROL - ALARA CFR41.10/43.4 2.3.2 Describe the personnel hazards when operating RHR in Suppression Pool Cooling mode. (ALARA) 2.5 287 q095 RO 2.1.32: 3.4 BANK NRC 6/01 Internal Hazard to personnel EMERGENCY PROCEDURES / PLAN - EOP/SAP CFR41.10/43.5 2.4.4 Given plan conditions, identify the Emergency Operating Procedures (SAP) requiring implementation.

4.0 298 q096 RO BANK NRC 4/00 Entry into SAP EMERGENCY PROCEDURES / PLAN - Local Operator Emergency Response CFR41.10/43.5 2.4.35 Describe the response of Non-Licensed Operators during implementation of the Emergency Plan 3.3 577 q097 RO BANK NRC 6/01 E-Plan changes 2/2001 EMERGENCY PROCEDURES / PLAN - AOP Immediate Operator Actions CFR41.10/43.5 2.4.49 Given plant conditions, identify the immediate operator actions to be taken.

4.0 q098 RO NEW AOP Immediate actions EMERGENCY PROCEDURES / PLAN -

Communications CFR41.10/43.5 2.4.43 Identify methods and times for notification of Offsite agencies in the event of Emergency Plan activation.

2.8 q099 RO NEW EMERGENCY PROCEDURES / PLAN - Fire Protection actions CFR41.10/43.5 2.4.25 Describe the Control Room Operator Actions for a fire in Division I Diesel Generator Room.

2.9 578 q100 RO BANK NRC 6/01 K/A CATEGORY TOTALS:

4 2

2 5

TIER 3 GROUP POINT TOTAL 13

GRAND GULF NUCLEAR STATION AUGUST 2002 BWR SRO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 ES-401-1 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO

/BOTH REC RELATED K/A ORIGIN NOTES:

295003 Partial or Complete Loss of AC Power/ 6 CFR41.7 01 Given a lockout on BOP Transformer 12B, determine the configuration of the AC Distribution System.

3.8 507 q014 BOTH AK2.03: 3.9 BANK NRC 6/01 295006 SCRAM / 1 CFR41.6/41.10/43.5 02 Describe the position of control rods following a reactor scram and how position is determined.

4.4 10 q002 BOTH 201005 A3.02: 3.5 A4.02: 3.7 201003 K4.05: 3.3 A3.01: 3.6 A4.03: 3.5 BANK NRC 6/01 Rx SCRAM Immediate Actions 295007 High Reactor Pressure / 3 CFR41.5 01 Describe the response of the Turbine Pressure Control System on an increasing reactor pressure.

3.7 69 q003 BOTH 241000 K4.01: 3.8 A2.02: 3.7 BANK NRC 6/01 295009 Low Reactor Water Level / 2 CFR41.5/43.5 02 Given plant conditions, identify the status of Recirculation pumps originally operating in fast speed with a lowering water level.

3.1 121 q004 BOTH AK2.03: 3.2 AA1.03: 3.1 BANK NRC 3/98 295010 High Drywell Pressure / 5 CFR41.5 04 Given plant parameters and elevated in-leakage into the Drywell, determine the status of reactor coolant system integrity.

3.8 286 q005 BOTH AK3.05: 3.4 2.4.21: 4.3 BANK NRC 4/00 GGNS Drywell air leak 295013 High Suppression Pool Water Temp. / 5 CFR41.5/41.10/43.2/43.5 01 During a surveillance operating RCIC, determine how often Suppression Pool Temperature is required to be monitored and the threshold for alternate actions.

4.0 q018 BOTH AA1.02: 3.9 2.1.33: 4.0 NEW 295014 Inadvertent Reactivity Addition / 1 CFR41.1/41.2/41.6/43.6 01 With the reactor in startup conditions such that the reactor has dropped subcritical, what are the operator actions if a high worth control rod is withdrawn fully.

4.1 204 q006 BOTH BANK NRC 6/01 Susquehanna reactivity event 7/98 295015 Incomplete SCRAM / 1 CFR41.6/43.6 01 Describe the method to be used to allow the insertion of control rods using RCIS during an ATWS.

3.7 203 q007 BOTH BANK NRC 6/01 295016 Control Room Abandonment / 7 CFR41.7/41.10/43.5 04 Determine the impact of operation of circuit breakers during operations from the Remote Shutdown Panel.

3.2 701 q076 SRO 239001 K2.01: 3.3 MOD NRC 3/98 295017 High Offsite Release Rate / 9 CFR41.11/41.13/43.4 08 Given plant conditions with MSIV Leakage Control System operating, determine the mechanism for monitoring radiological release to locations outside the Auxiliary Building.

3.1 q019 BOTH AA1.09: 3.6 NEW 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 CONT.

ES-401-1 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO

/BOTH REC RELATED K/A ORIGIN NOTES:

295023 Refueling Accidents / 8 CFR41.2/41.6/43.6/43.7 03 Describe the requirements to ensure proper positioning of spent fuel bundles being moved on the Refueling Platform Grapple.

3.6 187 q077 SRO BANK NRC 6/01 Damaged Fuel at RBS 295024 High Drywell Pressure / 5 CFR41.9/41.10/43.5 14 Given conditions determine ability to initiate Containment Spray per EOPs.

3.9 601 q008 BOTH NEW High Drywell Pressure is EP entry for CTMT Press. Control 295025 High Reactor Pressure / 3 CFR41.3/43.2 05 State the Reactor Vessel pressure Safety Limit and its basis and actions to be taken if violated.

4.7 q080 SRO 2.2.22: 4.1 2.2.25: 3.7 EK1.02: 4.2 MOD 295026 Suppression Pool High Water Temp. / 5 CFR41.7/41.9/41.10/41.14/43.5 01 Given plant conditions, determine when Suppression Pool Cooling is no longer effective and alternate actions are required.

4.0 301 q024 BOTH BANK NRC 12/00 295027 High Containment Temperature / 5 CFR41.9/41.10/43.2 03 Given plant conditions, determine the Technical Specification Bases for shutting down the Reactor due to a high Containment Temperature.

3.7 512 q025 BOTH 2.2.25: 3.7 BANK NRC 6/01 295030 Low Suppression Pool Water Level / 5 CFR41.9/41.10/43.5 04 Given plant conditions, determine the method to use to makeup inventory to the Suppression Pool.

4.0 q079 SRO NEW 295031 Reactor Low Water Level / 2 CFR41.2/41.3/41.10/43.5 01 Given plant conditions and a low reactor water level, determine core cooling mechanism and adequacy.

4.7 309 q009 BOTH 2.1.1: 3.8 2.4.21: 4.3 BANK NRC 12/00 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 CFR41.1/41.2/41.6/43.6 04 Determine when plant conditions allow the termination of Standby Liquid Control with multiple control rods stuck out and what is the basis.

4.5 216 q010 BOTH EK1.04: 3.6 EK1.05: 3.6 EA2.03: 4.4 BANK NRC 4/00 295038 High Offsite Release Rate / 9 CFR41.10/41.12/43.4/43.5 02 Given plant conditions and procedures, determine the protective action recommendations to be recommended to the state and local officials in an emergency.

4.4 q030 BOTH 2.4.44: 4.0 BANK 500000 High Containment Hydrogen Conc. / 5 CFR41.10/43.5 01 Determine the bases for the Hydrogen Control leg of EP 3.

3.9 505 q078 SRO 2.4.18: 3.6 BANK NRC 6/01 295009 Low Reactor Water Level / 2 CFR41.7/43.5 02 Given a steam flow / feed flow mismatch and plant conditions, determine the reactor water level response and response of Reactor Water Level control.

3.7 q081 SRO NEW moved to AA2 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1 CONT.

ES-401-1 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO/

BOTH REC RELATED K/A ORIGIN NOTES:

295003 Partial or Complete Loss of AC Power / 6 CFR41.4/43.5 03 During a station blackout with DG 13 available, determine the alignment to supply power from 17AC to 16AB.

3.9 q082 SRO AK1.06: 4.0 AA1.01: 3.8 NEW 295016 Control Room Abandonment / 7 CFR41.5/41.10/43.5 02 Given parameters from the Remote Shutdown Panel indications, determine actual and Narrow Range RPV level.

4.3 603 q017 BOTH 2.1.25: 3.1 2.4.4: 3.6 MOD NRC 6/01 295014 Inadvertent Reactivity Addition / 1 CFR41.6/43.6 02 Determine effects on the reactor of the fast opening a Recirc Flow Control Valve.

3.8 116 q012 BOTH 202002 K1.02: 4.2 K3.02: 4.0 BANK NRC 3/98 295030 Low Suppression Pool Water Level / 5 CFR41.7/41.9 02 Evaluate Suppression Pool Temperature, with a Low Suppression Pool Level.

3.9 8

q028 BOTH BANK NRC 6/01 Caution 2 EOP-2 500000 High Containment Hydrogen Conc. / 5 CFR41.10/43.5 2.

1.

20 Given applicable SOP, graph, and plant conditions, determine the power settings for the Hydrogen Recombiners and time to full power.

4.2 219 q011 BOTH EA1.03: 3.2 2.1.25:3.1 BANK NRC 4/00 moved to Generics 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 ES-401-1 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO/

BOTH REC RELATED K/A ORIGIN NOTES:

295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 CFR41.5/41.10/43.5 2.

4.

11 Given plant conditions and the power to flow map, determine the actions to be taken.

3.6 602 q013 BOTH 2.4.1: 4.6 AA2.01: 3.8 2.1.25: 3.1 MOD NRC 4/00 moved to Generics 295002 Loss of Main Condenser Vacuum / 3 CFR41.4/43.4 04 Describe the basis for the isolation of the Main Steam Isolation Valves on a loss of condenser vacuum.

3.4 220 q015 BOTH AK3.05: 3.4 BANK NRC 4/00 295004 Partial or Complete Loss of DC Power / 6 295005 Main Turbine Generator Trip / 3 CFR41.5/41.6 05 Given a spurious Main Turbine Generator trip, determine the initial effect on Reactor Power and Reactor Pressure.

3.9 553 q001 BOTH AA2.04: 3.8 AA2.03: 3.1 BANK NRC 6/01 295008 High Reactor Water Level / 2 CFR41.4/41.5 01 Identify the effects of a High Reactor Water Level on the Main Turbine and Reactor Feed Pump Turbines.

3.2 275 q016 BOTH 245000 A3.01: 3.6 259001 K6.07: 3.8 BANK NRC 6/01 295011 High Containment Temperature / 5 295012 High Drywell Temperature / 5 295018 Partial or Complete Loss of CCW / 8 CFR41.4/41.10/43.5 03 Given a partial loss of Component Cooling Water, determine actions to be taken and their basis.

3.3 314 q020 BOTH BANK NRC 12/00 295019 Partial or Complete Loss of Inst. Air / 8 CFR41.4/41.10/43.5 14 Given a reduction in Instrument Air Header pressure, determine a possible cause.

3.2 q021 BOTH NEW New Air Dryers moved to AK2 295020 Inadvertent Cont. Isolation / 5 & 7 CFR41.4/41.7/41.9 03 Given a loss of Instrument Air System pressure to the Auxiliary Building, determine the effects on the ability to remove heat from the Containment.

3.3 q022 BOTH AK3.03: 3.2 NEW 295021 Loss of Shutdown Cooling / 4 CFR41.5/41.10/43.5 01 Given plant parameters and the graphs from the Inadequate Decay Heat Removal ONEP, determine time to boil.

3.6 604 q032 BOTH AK1.01: 3.8 MOD NRC 3/98 Multiple graphs for various 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 2 CONT.

ES-401-1 E/APE #/NAME/SAFETY FUNCTION K

1 K

2 K

3 A

1 A

2 G

TOPIC(S)

IMP SRO/RO/

BOTH REC RELATED K/A ORIGIN NOTES:

295022 Loss of CRD Pumps / 1 CFR41.5 04 Describe the affects on reactor water level during a reactor startup with minimal decay heat and a loss of CRD Pumps. (RWCU is lined up to blowdown to the main condenser to compensate for CRD flow.)

2.6 55 q023 BOTH AK2.04: 2.7 AK2.05: 2.5 BANK NRC 6/01 295028 High Drywell Temperature / 5 CFR41.4/41.7 03 Given rising temperatures in the Reactor lower cavity area, determine the response of the Drywell Cooling system.

3.9 q026 BOTH EK2.04: 3.6 EK3.04: 3.8 NEW 295029 High Suppression Pool Water Level / 5 CFR41.9/41.10 01 Identify the bases for Emergency RPV Depressurization when Suppression Pool Level cannot be maintained below 24.4 feet.

3.7 513 q027 BOTH BANK NRC 6/01 295032 High Secondary Containment Area Temperature / 5 CFR41.4/41.10/43.5 03 Determine the systems affected by high temperatures in the RHR A Pump Room.

3.9 48 q033 BOTH EK2.07: 3.8 219000 A1.08: 3.6 A2.14: 4.3 BANK NRC 3/98 Steam leak at GGNS in Steam Condensing piping 295033 High Secondary Containment Area Radiation Levels / 9 CFR41.12/43.4 03 Given operations during a Refueling outage, determine the cause of elevated radiation levels in the area of the Fuel Pool Cooling Heat Exchangers.

3.7 q029 BOTH NEW Fuel Pool HX in an open area of Sec CTMT 295034 Secondary Containment Ventilation High Radiation / 9 CFR41.4/41.10/41.13/43.4 05 Given plant conditions, determine the configuration of the Fuel Handling area Ventilation system.

3.7 q083 SRO NEW 295035 Secondary Containment High Differential Pressure / 5 CFR41.4/41.7 02 Describe the operation of the Standby Gas Treatment System with regard to Auxiliary Building and Enclosure Building Pressures.

3.8 q034 BOTH BANK 295036 Secondary Containment High Sump/Area Water Level / 5 CFR41.4/41.10/43.5 03 Given a rising water level in an ECCS Pump Room, determine appropriate actions to be taken with regard to the overall plant operation.

3.8 58 q035 BOTH EK3.03: 3.6 EA2.02: 3.1 BANK NRC 3/98 EOP-4 600000 Plant Fire On Site / 8 CFR41.10/43.5 04 Given a fire at the Hydrogen Bulk Storage Facility describe the actions to be taken to combat the fire.

3.4 q031 BOTH 2.4.25: 3.4 2.1.32: 3.8 NEW moved to 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 1 ES-401-1 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

201005 RCIS CFR41.6/43.6 01 Given a failure of the Turbine First Stage pressure signal to RCIS, determine the mode of control rod movement.

3.2 q037 BOTH NEW 202002 Recirculation Flow Control CFR41.6 01 Given plant conditions following an actuation of EOC-RPT, determine the Recirculation Pump circuit breaker configuration.

3.4 q038 BOTH 202001 A2.15: 3.9 NEW EOC-RPT changes 2002 203000 RHR/LPCI: Injection Mode CFR41.8 08 State the basis for monitoring reactor pressure when aligning the RHR system for injection into the vessel for the LPCI mode.

4.3 60 q039 BOTH K4.01: 4.2 K4.02: 3.3 A3.01: 3.8 A3.08: 4.1 K1.17: 4.0 BANK NRC 6/01 Where is pressure sensed on LPCI for operation of the injection valve 209001 LPCS CFR41.7 02 Describe the method of operation of the ADS logic given LPCS and LPCI A out of service during a LOCA.

3.9 q040 BOTH 218000 K6.01: 4.1 K6.02: 4.1 NEW moved to K3 209002 HPCS CFR41.7/41.8 01 Given a spurious initiation of the HPCS system, determine the effect on Reactor water level.

3.9 519 q041 BOTH 259002 A2.08: 4.5 BANK NRC 6/01 211000 SLC CFR41.1/41.6/41.7/43.6 04 Given the initiation of Standby Liquid Control in an ATWS, discern the parameters indicating injection to the reactor.

4.6 q042 BOTH A4.01: 3.9 A4.03: 4.1 A4.06: 3.9 A4.07: 3.6 NEW 212000 RPS CFR41.6 01 Describe the response of the RPS Power System upon an ESF inverter loss.

3.8 102 q043 BOTH K1.04: 3.6 BANK NRC 3/98 215004 Source Range Monitor CFR41.6/41.5 2.

1.

32 Determine the conditions that would allow the operation of Source Range Monitor detector drives.

3.4 q046 BOTH A4.04: 3.2 moved to generics 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

ES-401-1 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

215005 APRM / LPRM CFR41.6/41.7 03 Determine the affects on APRMs with reduced LPRM inputs.

3.3 326 q045 BOTH BANK NRC 12/00 216000 Nuclear Boiler Instrumentation CFR41.5 23 Determine the ability of the Recirculation Pumps to start based on Reactor Temperature.

3.4 122 q047 BOTH 202001 A4.01: 3.7 BANK NRC 3/98 217000 RCIC CFR41.5/41.7/41.10 01 Given plant conditions, determine the operation of RCIC and indications of injection.

3.7 328 q048 BOTH BANK NRC 12/00 218000 ADS CFR41.7 03 Given plant conditions determine automatic operation of Automatic Depressurization.

4.0 q049 BOTH NEW 223001 Primary CTMT and Auxiliaries CFR41.7/41.8 08 Given plant conditions and electrical busses that are unavailable, determine which components are available.

3.0 528 q050 BOTH K2.09: 2.9 K2.10: 2.9 BANK NRC 6/01 223002 PCIS / Nuclear Steam Supply Shutoff CFR41.7/41.9 09 Given plant conditions, evaluate the systems that should actuate or isolate.

3.7 385 q051 BOTH 2.4.21: 4.3 2.4.4: 4.3 BANK NRC 12/00 226001 RHR/LPCI: CTMT Spray Mode CFR41.7/41.8 07 Determine conditions that would result in automatic initiation of RHR Containment Spray mode.

3.5 q062 BOTH A3.01: 3.0 BANK moved to A3 239002 SRVs CFR41.3 05 Describe the operation of the Safety Relief Valves in different modes of operation.

(system air pressure or reactor pressure) 3.3 337 q052 BOTH BANK NRC 12/00 241000 Reactor / Turbine Pressure Regulator CFR41.5 11 Describe the response of the plant with a failure of the Main Stop and Control Valves closed with the reactor at power.

3.4 244 q053 BOTH A1.01: 3.8 A1.02: 3.9 A1.07: 3.7 BANK NRC 4/00 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 1 CONT.

ES-401-1 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

259002 Reactor Water Level Control CFR41.4/41.7 03 Describe the effects on the Reactor Water Level Control System from a failure of the Narrow Range Reactor Water level signal.

3.8 q057 BOTH BANK 261000 SGTS CFR41.7/41.11 01 Given conditions determine the automatic operation of Standby Gas Treatment.

3.8 q054 BOTH NEW 262001 AC Electrical Distribution CFR41.1/41.7 04 Given a loss of offsite power and plant conditions, determine the type of load sequencing to occur on ESF busses.

3.6 q065 BOTH MOD 264000 EDGs CFR41.8 2.

4.

4 During degraded grid conditions, determine the response of the diesel generators.

4.3 11 q058 BOTH K4.05: 3.2 A3.05: 3.4 BANK NRC 6/01 290001 Secondary CTMT CFR41.9 09 Describe the ability of Auxiliary Building Fire Protection system to be restored following Auxiliary Building isolation in conjunction with a loss of AC power.

3.6 264 q069 BOTH A2.06: 4.0 286000 A2.09: 2.8 BANK NRC 4/00 202002 Recirculation Flow Control CFR41.6 01 Given plant conditions, identify the plant response to a Recirc Flow Control Runback.

3.4 235 q055 BOTH BANK NRC 4/00 Recirc Pumps 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 2 ES-401-1 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO

/ BOTH RELATED K/A ORIGIN NOTES:

201001 CRD Hydraulic CFR41.5/41.6 02 During a reactor scram with scram signals not reset, determine the CRD System flow rates.

2.6 q036 BOTH MOD 202001 Recirculation CFR41.3/41.5 10 Given plant conditions in regard to Reactor Recirc. Pump seals, determine the failed mechanism.

3.9 540 q059 BOTH A1.09: 3.3 A1.10: 2.7 BANK NRC 6/01 204000 RWCU 205000 Shutdown Cooling CFR41.2/41.3/41.4/41.5 03 Given plant conditions and configuration lineup, determine valid method for determining Reactor coolant temperature.

3.5 541 q060 BOTH K3.03: 3.9 A1.03: 3.3 A1.06: 3.7 A1.08: 2.9 BANK NRC 6/01 215003 IRM CFR41.6 01 Identify the power supply to the Intermediate Range Nuclear Instrumentation.

2.7 q044 BOTH NEW 219000 RHR /LPCI Suppression Pool Cooling Mode CFR41.7 04 Describe the method used to control Suppression Pool Temperature and cooldown rate.

2.9 q061 BOTH A4.12: 4.1 NEW moved to K5 234000 Fuel Handling Equipment 239003 MSIV Leakage Control 245000 Main Turbine Gen., and Auxiliaries CFR41.4/41.10/43.5 10 Discuss the basis for the limit on Main Generator Reactive loading at GGNS.

2.6 44 q063 BOTH K4.06: 2.8 A4.14: 2.5 BANK NRC 3/98 GGNS generator has lower limits due to reverse power.

259001 Reactor Feedwater CFR41.4 01 Given a failure of the RFPT Lube Oil System, identify the Reactor Feedwater System response.

3.7 q056 BOTH K1.11: 2.7 K4.06: 2.6 K6.09: 2.9 BANK 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 2 CONT.

ES-401-1 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP REC SRO/RO/

BOTH RELATED K/A ORIGIN NOTES:

262002 UPS (AC/DC)

CFR41.7/41.10/43.5 01 Concerning the ESF Static Inverters, identify the correct response of the inverter on a loss of normal power supply.

2.6 544 q066 BOTH A2.01: 2.8 A3.01: 3.1 BANK NRC 6/01 263000 DC Electrical Distribution 271000 Offgas 272000 Radiation Monitoring CFR41.10/41.11/43.4/43.5 01 Identify the normal radiation monitoring alarms received on a reactor down power from full power and their cause. (Hydrogen Water Chemistry) 3.2 q067 BOTH NEW 286000 Fire Protection CFR41.4 08 Discern the response of the Diesel Driven Fire Pumps on an auto start signal with a failure to start.

3.3 3

q068 BOTH K5.05: 3.1 K4.07: 3.3 A3.01: 3.4 A4.06: 3.4 BANK NRC 3/98 290003 Control Room HVAC CFR41.7/41.11/43.4 05 Identify the response of the Control Room HVAC system upon receipt of radiation alarms.

3.2 q070 BOTH NEW 300000 Instrument Air CFR41.4/41.10/43.5 02 Describe the crosstie between Service and Instrument Air and when it occurs.

3.0 q071 BOTH BANK moved to K3 400000 Component Cooling Water CFR41.4 01 Discuss the status of the standby feature of the ESF powered CCW pump during a LOCA.

3.9 105 q072 BOTH BANK 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE PLANT SYSTEMS - TIER 2 GROUP 3 ES-401-1 SYSTEM #/NAME K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

TOPIC(S)

IMP RELATED K/A REC SRO/RO

/ BOTH ORIGIN NOTES:

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 CFR41.4 08 Describe the effects of Low Pressure Feedwater Heater isolation on the Condensate System operation and plant operations.

3.1 294 q064 BOTH A3.01: 2.7 A3.04: 3.0 A3.07: 2.9 BANK NRC 4/00 268000 Radwaste CFR41.13/43.4 02 Given conditions and a Liquid Radwaste discharge in progress, determine whether a release will continue.

3.6 110 q073 BOTH 272000 A3.03: 3.5 BANK NRC 3/98 Process Rad Monitor isolation 288000 Plant Ventilation CFR41.7 05 Describe the response of the Auxiliary Building Ventilation System on High Radiation.

3.6 q074 BOTH K1.02: 3.4 NEW 290002 Reactor Vessel Internals CFR41.3/41.14/43.2 05 Identify the consequence of exceeding a thermal limit.

4.2 q075 BOTH K5.01: 3.9 NEW moved to 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 AUGUST 2002 BWR SRO EXAMINATION OUTLINE GENERIC KNOWLEDGE AND ABILITIES TIER 3 ES-401-5 CATEGORY C1 C2 C3 C4 TOPIC(S)

IMP REC #

SRO/RO

/BOTH RELATED K/A ORIGIN NOTES:

CONDUCT OF OPERATIONS - Deviations from Approved documents in emergency conditions CFR41.10/43.3/43.5 2.1.1 Given plant conditions that require a deviation from procedures, describe the process and when this action can be taken.

3.8 135 q084 SRO 2.1.2: 4.0 BANK NRC 12/00 50.54x CONDUCT OF OPERATIONS - Shift Manning CFR41.10/43.1/43.2/43.5 2.1.4 Determine requirements for fire brigade and conditions when the fire brigade may be less than required.

3.4 486 q085 SRO 2.4.26: 3.3 BANK NRC 6/01 CONDUCT OF OPERATIONS - Chemistry Limits CFR41.10/43.2/43.5 2.1.34 Given plant conditions, procedures, and coolant samples, determine actions to be taken based on chemistry results.

2.9 197 q086 SRO BANK NRC 6/01 Use of EPRI Guidelines vs Tech Specs.

CONDUCT OF OPERATIONS - Facility License CFR43.1 2.1.10 Given reactor thermal power, determine acceptability with regard to the Station Operating License.

3.9 q087 SRO NEW CONDUCT OF OPERATIONS - Station Graphs CFR41.6/41.10/43.2/43.5 2.1.25 Given plant parameters and Tech Specs determine operability of Standby Liquid Control.

3.1 q088 SRO 2.1.33: 4.0 2.2.22: 4.1 NEW SLC concentration graphs EQUIPMENT CONTROL - Refueling Procedures &

Limitations CFR41.10/43.6/43.7 2.2.26 Describe the permissions required for all lifting on the Containment Refuel Floor during Core Alterations.

3.7 702 q089 SRO 2.2.27: 3.5 NEW EQUIPMENT CONTROL - Refueling CFR41.9/41.10/43.2/43.4/43.6/43.7 2.2.28 Given an evolution to be performed in Containment during fuel handling operations, evaluate the allowances of the evolution.

3.5 411 q090 SRO BANK NRC 12/00 EQUIPMENT CONTROL - Troubleshooting CFR41.10/43.5 2.2.20 Given a situation in the plant, apply the rules concerning MAIs vs allowed troubleshooting.

3.3 492 q091 SRO BANK NRC 6/01 EQUIPMENT CONTROL - Core Alterations CFR43.6/43.7 2.2.32 Given a situation during refueling operations, apply the Criticality Rules.

3.3 491 q092 SRO 2.2.28: 3.5 BANK NRC 6/01 RADIATION CONTROL - ALARA CFR41.12/43.4 2.3.2 Given a situation requiring independent verification in a high radiation area determine allowances for waiving verification based on ALARA and requirements for meeting verification.

2.9 127 q093 SRO 2.1.29: 3.3 2.2.11: 3.4 2.2.13: 3.8 BANK NRC 6/01 RADIATION CONTROL - Radiation Work Permits CFR41.10/41.12/43.4/43.5 2.3.7 Given conditions and procedures, determine applicability of radiation work permits.

3.3 q094 SRO NEW 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

REVISION 0 4/29/2002 PAGE 13 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1 GRAND GULF NUCLEAR STATION AUGUST 2002 BWR SRO EXAMINATION OUTLINE GENERIC KNOWLEDGE AND ABILITIES TIER 3 CONT.

ES-401-5 CATEGORY C1 C2 C3 C4 TOPIC(S)

IMP REC #

SRO/RO

/BOTH RELATED K/A ORIGIN NOTES:

EMERGENCY PROCEDURES / PLAN - EOPs and usage CFR41.10/43.5 2.4.20 Given conditions delineated in Caution 1 of the EOPs, determine when EOP transition to contingencies is required.

4.0 415 q095 SRO 2.4.18: 3.6 2.4.22: 4.0 2.4.23: 3.8 BANK NRC 12/00 EOP usage EMERGENCY PROCEDURES / PLAN - SRO Responsibilities Security Threat CFR41.10/43.5 2.4.40 Given a security threat taking over the Main Control Room, determine actions to be taken.

4.0 192 q096 SRO 2.1.2: 4.0 2.4.49: 4.0 2.4.11: 3.6 2.4.28: 3.3 BANK NRC 4/00 Control Room abandonment (terrorist attack)

EMERGENCY PROCEDURES / PLAN - EOPs CFR41.10/43.5 2.4.7 Given plant conditions and EOPs, determine a course of action to be taken.

3.8 q097 SRO 2.4.14: 3.9 NEW EMERGENCY PROCEDURES / PLAN -

Reportability CFR41.10/43.5 2.4.30 Given plant conditions, determine the reportability of plant conditions and time requirements.

3.6 q098 SRO NEW EMERGENCY PROCEDURES / PLAN - SAPs CFR41.10/43.5 2.4.16 Given conditions requiring transition from EOPs to SAPs, identify implementation and responsibilities for implementation.

4.0 q099 SRO 2.4.8: 3.8 2.4.14: 3.9 NEW EOP/SAP EMERGENCY PROCEDURES / PLAN -

Emergency Response Facilities CFR41.10/43.5 2.4.42 Given plant conditions, determine what emergency facilities are to be activated.

3.7 q100 SRO NEW E-Plan 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