ML093620008

Final Outlines (Folder 3)

ML093620008
Person / Time
Site:	Peach Bottom
Issue date:	12/07/2009
From:	NRC Region 1
To:
Hansell S
Shared Package
ML091380320	List: ML092090318 ML093340500 ML093370113 ML093620008 ML093620012 ML093620016 ML093620021 ML100500538 ML100810008
References
TAC U01792
Download: ML093620008 (21)

Text

ES-401 BWR Examination Outline FORM ES-401-1 Facility Name: Peach Bottom Date of Exam: 12/07/2009 Tier Group RO KJA Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G

• Total A2 G* Total 1. Emergency

& Abnormal Plant Evolutions 1 3 3 3 N/A 4 4 N/A 3 20 4 3 7 2 1 2 1 1 5 1 5 1 7 2 1 3 Tier Totals 4 5 4 4 27 6 4 10 2. Plant Systems 1 2 2 2 3 3 2 3 2 2 2 3 26 3 2 5 2 1 1 1 1 1 1 1 2 1 1 1 12 0 2 1 3 Tier Totals 3 3 3 4 4 3 4 4 3 3 4 38 5 3 8 3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 3 4 7 2 3 3 2 2 1 2 2 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA 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 RO exam must total 75 points and the SRO-only exam must total 25 points. 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those KlAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.* The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KlAs. 8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category.

Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams. 9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3.

Limit SRO selections to KlAs that are linked to 10 CFR 55.43. ES-40 1, Page 17 of 34 ES-401 BWR Examination Outline FORM ES-401-1 Facility Name: Peach Bottom Date of Exam: 12/07/2009 RO KJA Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4

• 1. 1 3 3 3 4 4 3 20 4 3 7 Emergency

& Abnormal 2 1 2 1 N/A 1 1 N/A 1 7 2 1 3 Plant Evolutions Tier Totals 4 5 4 5 5 4 27 6 4 10 1 2 2 2 3 3 2 3 2 2 2 3 26 3 2 5 2. Plant 2 1 1 1 1 1 1 1 2 1 1 1 12 0 2 1 3 Systems Tier Totals 3 3 3 4 4 3 4 4 3 3 4 38 5 3 8 3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 Categories 10 7 2 3 3 2 2 1 2 2 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA 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 RO exam must total 75 points and the SRO-only exam must total 25 points. 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those KlAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.* The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KlAs. 8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category.

Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams. 9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3.

Limit SRO selections to KlAs that are linked to 10 CFR 55.43. ES-40 1, Page 17 of 34 ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1/Group 1 (RO) EIAPE # I Name I Safety Function K KIA Topic(s) IR 95001 Partial or Complete Loss of Forced Core Flow Circulation 11 &4 95003 Partial or Complete Loss of AC 16 295004 Partial or Total Loss of DC Pwr 16 95005 Main Turbine Generator Trip 13 295006 SCRAM I 1 295016 Control Room Abandonment 17 295018 Partial or Total Loss of CCW 18 295019 Partial or Total Loss of Ins!. Air I 8 295021 Loss of Shutdown Cooling 14 295023 Refueling Acc I 8 295024 High Drywell Pressure I 5 295025 High Reactor Pressure 13 295026 Suppression Pool High Water emp./5 295027 High Containment Temperature 15 295028 High Drywell Temperature 15 295030 Low Suppression Pool Wtr LviI 5 95031 Reactor Low Water Level 12 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 11 295038 High Off-site Release Rate I 9 600000 Plant Fire On Site 18 700000 Generator Voltage and Electric Grid Disturbances I 6 tegory Totals: o o 2 333 Core flow Indication 2.9 Failsafe component design 2.6 4 3.5 r/shutdown implications In accident (e.g .. t or loss of residual heat removal) 3.8 4.2 3.3 4.2 Component cooling water systems: Plant-Specific 3.1 3.6 4.2 4,1 3.9 3.2 ECCS systems (NPSH considerations):

Plant-Specific 3.6 4,2 Reactor water level effects on reactor power 4.1 3,6 Fire fighling eqUipment used on each class of fire 2.6 3.9 # 0 ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1/Group 1 (RO) EIAPE # I Name I Safety Function K KIA Topic(s) IR 95001 Partial or Complete Loss of Forced Core Flow Circulation 11 & 4 95003 Partial or Complete Loss of AC 16 295004 Partial or Total Loss of DC Pwr 16 95005 Main Turbine Generator Trip 13 295006 SCRAM I 1 295016 Control Room Abandonment 17 295018 Partial or Total Loss of CCW 18 295019 Partial or Total Loss of Ins!. Air I 8 295021 Loss of Shutdown Cooling 14 295023 Refueling Acc I 8 295024 High Drywell Pressure I 5 295025 High Reactor Pressure 13 295026 Suppression Pool High Water emp./5 295027 High Containment Temperature 15 295028 High Drywell Temperature 15 295030 Low Suppression Pool Wtr LviI 5 95031 Reactor Low Water Level 12 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 11 295038 High Off-site Release Rate I 9 600000 Plant Fire On Site 18 700000 Generator Voltage and Electric Grid Disturbances I 6 tegory Totals: o o 2 333 Core flow Indication 2.9 Failsafe component design 2.6 4 3.5 r/shutdown implications In accident (e.g .. t or loss of residual heat removal) 3.8 4.2 3.3 4.2 Component cooling water systems: Plant-Specific 3.1 3.6 4.2 4,1 3.9 3.2 ECCS systems (NPSH considerations):

Plant-Specific 3.6 4,2 Reactor water level effects on reactor power 4.1 3,6 Fire fighling eqUipment used on each class of fire 2.6 3.9 # 0 ES-401 3 Form ES-401-1 BWR Examination Outline Form Emergency and Abnormal Plant Evolutions

-Tier 1/Group 2 (RO) EIAPE # I Name I Safety Loss of Main Condenser Vac I High Reactor Pressure High Reactor Water Levell Low Reactor Water Levell 10 High Drywell Pressure 11 High Containment Temp I 2 High Drywell Temperature 3 High Suppression Pool Temp. 4 Inadvertent Reactivity Addition 15 Incomplete SCRAM 17 High Off-site Release Rate I 1I?(l"ri?n Inadvertent ConI. Isolation I 5 &

Loss of CRD Pumps 1129 1 50.19 High Suppression Pool Wtr LviI iation Levels I 9 Radiathm 19 Pressure I 5 High Secondary Containment High Secondary Containment Area Secondary Containment Ventilation Secondary Containment High Secc)ndary Containment High mp/Area Water Levell High CTMT Hydrogen Conc, Category KKK 2 3 o 2 o KJA Topic(s) power water level indication ventilation/cooling relationship Corrtrol rod blocks : I Kn",,"edlle of EOP entry conditions and immediate action steps, 01 the high water level P()intTotal:

IR # o 3.4 o 3.3 o 3,7 o o o o o .4 o 7 o ES-401, Page 19 of ES-401 3 Form ES-401-1 BWR Examination Outline Form ES-401-Emergency and Abnormal Plant Evolutions

-Tier 1/Group 2 (RO) EIAPE # I Name I Safety Function Loss of Main Condenser Vac I 3 High Reactor Pressure 13 High Reactor Water Levell 2 Low Reactor Water Levell 2 10 High Drywell Pressure 15 11 High Containment Temp I 5 2 High Drywell Temperature 15 3 High Suppression Pool Temp. 15 4 Inadvertent Reactivity Addition 11 15 Incomplete SCRAM 11 17 High Off-site Release Rate I 9 1I?(l"ri?n Inadvertent ConI. Isolation I 5 & 7 Loss of CRD Pumps 11 1129 1 50.19 High Suppression Pool Wtr LviI 5 High Secondary Containment Area 15 High Secondary Containment Area iation Levels I 9 Secondary Containment Ventilation Radiathm 19 Secondary Containment High Pressure I 5 Secc)ndary Containment High mp/Area Water Levell 5 High CTMT Hydrogen Conc, 15 Category Totals: KKK o 2 3 o 2 o KJA Topic(s) power water level indication ventilation/cooling relationship Corrtrol rod blocks : I Kn",,"edlle of EOP entry conditions and immediate action steps, 01 the high water level P()intTotal:

ES-401, Page 19 of 34 IR # o 3.4 o 3.3 o 3,7 o o o o o .4 o 7 ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems -Tier 2/Group 1 (RO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2. A 3 A 4 G KIA Topic(s) IR # 203000 RHR/LPCI:

Injection Mode 0 3 0 2 Initiation logic: Core cooling methods 2.7; 3.5 2 205000 Shutdown Cooling 0 3 Heat removal mechanisms 2.8 1 206000 HPCI 0 7 D.C. power: BWR-2. 3. 4 3.7 1 207000 Isolation (Emergency)

Condenser 0 209001 LPCS 0 1 04. 50 Reactor water level; Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. 3.8; 4.2 2 209002 HPCS 0 211000 SLC 0 4 Explosive valve operation 3.1 1 212000 RPS 0 1 RPS motor-generator output voltage 2.8 1 2150031RM 0 2 Reactor power indIcation response to rod position changes 3.7 1 215004 Source Range Monitor 0 215005 APRM / LPRM 0 7 Flow biased trip setpoints 3.7 1 217000 RCIC 0 7 0 3 Suppression pool level; System valves 3.3; 3.4 2 218000 ADS 0 5 12 A.C. power: Plant-Specific; Knowledge of surveillance procedures.

3; 3.7 2 223002 PCIS/Nuclear Steam Supply Shutoff 0 2 Valve closures 3.5 1 239002 SRVs 0 1 0 5 Reactor pressure control; Allows for SRV operation from more than one location:

Plant*Specific 3.9; 3.6 2 259002 Water Level Control 0 3 All individual component controllers when transferring from manual to automatic modes 3.8 1 261000 SGTS () 6, Valve closures 2.9 1 262001 AC Electrical Distribution 0 9 Exceeding voltage limitations 3.1 1 262002 UPS (AC/DC) 0 3 Static inverter 2.7 1 263000 DC Electrical Distribution 0 1 Major D.C. loads 3.1 1 264000 EDGs 0 1 Emergency generator trips (normal) 3.5 1 300000 Instrument Air 0 4 01. 20 Cooling water to compressor; Ability to interpret and execute procedure steps 2.8; 4.6 2 400000 Component Cooling Water 0 1 SetpOlnts on instrument signal levels for nonnal operations, warnings, and trips that are applicable to the CCWS 3 1 0 KIA Category Totals: 2 2 2 3 3 2 3 2. 2 2 3 Group Point Total: 26 ES-401, Page 20 of 34 ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems -Tier 2/Group 1 (RO) System # / Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2. 3 4 203000 RHR/LPCI:

Injection Mode 0 0 2.7; 2 3 2 Initiation logic: Core cooling methods 3.5 205000 Shutdown Cooling 0 3 Heat removal mechanisms 2.8 1 206000 HPCI 0 7 D.C. power: BWR-2. 3. 4 3.7 1 207000 Isolation (Emergency) 0 Condenser 0 04. Reactor water level; Ability to verify system alarm setpoints 3.8; 209001 LPCS 1 50 and operate controls identified in the alarm response 2 manual. 4.2 209002 HPCS 0 211000 SLC 0 4 Explosive valve operation 3.1 1 212000 RPS 0 RPS motor-generator output voltage 2.8 1 1 2150031RM 0 Reactor power indIcation response to rod position changes 3.7 1 2 215004 Source Range Monitor 0 215005 APRM / LPRM 0 7 Flow biased trip setpoints 3.7 1 217000 RCIC 0 0 Suppression pool level; System valves 3.3; 2 7 3 3.4 218000 ADS 0 A.C. power: Plant-Specific; Knowledge of surveillance 3; 3.7 2 5 12 procedures.

223002 PCIS/Nuclear Steam Supply 0 Valve closures 3.5 1 Shutoff 2 239002 SRVs 0 0 Reactor pressure control; Allows for SRV operation from 3.9; 2 1 5 more than one location:

Plant*Specific 3.6 259002 Water Level Control 0 All individual component controllers when transferring from 3.8 1 3 manual to automatic modes 261000 SGTS () 6, Valve closures 2.9 1 262001 AC Electrical Distribution 0 Exceeding voltage limitations 3.1 1 9 262002 UPS (AC/DC) 0 3 Static inverter 2.7 1 263000 DC Electrical Distribution 0 Major D.C. loads 3.1 1 1 264000 EDGs 0 Emergency generator trips (normal) 3.5 1 1 300000 Instrument Air 0 01. Cooling water to compressor; Ability to interpret and 2.8; 2 4 20 execute procedure steps 4.6 400000 Component Cooling Water 0 SetpOlnts on instrument signal levels for nonnal operations, 3 1 1 warnings, and trips that are applicable to the CCWS 0 KIA Category Totals: 2 2 2 3 3 2 3

2. 2 2 3 Group Point Total: 26 ES-401, Page 20 of 34 ES-401-1 5 Form ES-401-1 ES-401 BWR Examination Outline Plant Systems -Tier 2/Group 2 (RO) Form ES-401-1 System # I Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G KIA Topic(s) IR # 201001 CRD Hydraulic 0 201002 RMCS 0 201003 Control Rod and Drive Mechanism 0 201004 RSCS 0 201005 RCIS 0 201006 RWM 0 1 Rod position:

P-Spec(Not-BWR6) 3.2 1 202001 Recirculation 0 9 : Reactor water level 3.7 1 202002 Recirculation Flow Control 02, 37: Ability to determine operability and/or availability of safety related equipment.

3.6 1 RWCU 0 214000 RPIS 0 215001 Traversing In-core Probe 0 1 Primary containment isolation*

Mark-I&II(Not-BWR

1) 3.4 1 215002 RBM 0 216000 Nuclear Boiler Inst. 0 219000 RHR/LPCI:

Torus/Pool Cooling Mode 0 1 Valve operation 3.3 1 223001 Primary CTMT and Aux. 0 226001 RHR/LPCI:

CTMT Spray Mode 0 230000 RHR/LPCI:

Torus/Pool Spray Mode 1 2 Valve logic failure 3.2 1 1233000 Fuel Pool Cooling/Cleanup 0 2 RHR pumps 2.8 1 234000 Fuel Handling Equipment 0 239001 Main and Reheat Steam 0 239003 MSIV Leakage Control 0 241000 ReactorlTurbine Pressure Regulator 0 245000 Main Turbine Gen. / Aux. 0 2 Turbine operation and limitations 2.8 1 256000 Reactor Condensate 0 259001 Reactor Feedwater 0 5 Recirculation pump NPSH 2.9 1 268000 Radwaste 0 6 Drywell floor drains 2.9 1 2710000ffgas 0 9 Valve closures 2.6 1 272000 Radiation Monitoring 0 286000 Fire Protection 0 288000 Plant Ventilation 0 3 Plant air systems 2.7 1 290001 Secondary CTMT 0 290003 Control Room HVAC 0 290002 Reactor Vessel Internals 0 KIA Category Totals: 1 1 1 1 1 1 1 2 1 1 1 Group Point Total: 12 ES-401, Page 22 of 34 ES-401-1 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems -Tier 2/Group 2 (RO) System # I Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 0 201002 RMCS 0 201003 Control Rod and Drive Mechanism 0 201004 RSCS 0 201005 RCIS 0 201006 RWM 0 1 Rod position:

P-Spec(Not-BWR6) 3.2 1 202001 Recirculation 0 9 : Reactor water level 3.7 1 202002 Recirculation Flow Control 02, Ability to determine operability and/or availability of safety 3.6 1 37: related equipment.

RWCU 0 214000 RPIS 0 215001 Traversing In-core Probe 0 1 Primary containment isolation*

Mark-I&II(Not-BWR

1) 3.4 1 215002 RBM 0 216000 Nuclear Boiler Inst. 0 219000 RHR/LPCI:

Torus/Pool Cooling Mode 0 Valve operation 3.3 1 1 223001 Primary CTMT and Aux. 0 226001 RHR/LPCI:

CTMT Spray Mode 0 230000 RHR/LPCI:

Torus/Pool Spray Mode 1 Valve logic failure 3.2 1 2 1233000 Fuel Pool Cooling/Cleanup 0 RHR pumps 2.8 1 2 234000 Fuel Handling Equipment 0 239001 Main and Reheat Steam 0 239003 MSIV Leakage Control 0 241000 ReactorlTurbine Pressure Regulator 0 245000 Main Turbine Gen. / Aux. 0 Turbine operation and limitations 2.8 1 2 256000 Reactor Condensate 0 259001 Reactor Feedwater 0 Recirculation pump NPSH 2.9 1 5 268000 Radwaste 0 Drywell floor drains 2.9 1 6 2710000ffgas 0 Valve closures 2.6 1 9 272000 Radiation Monitoring 0 286000 Fire Protection 0 288000 Plant Ventilation 0 Plant air systems 2.7 1 3 290001 Secondary CTMT 0 290003 Control Room HVAC 0 290002 Reactor Vessel Internals 0 KIA Category Totals: 1 1 1 1 1 1 1 2 1 1 1 Group Point Total: 12 ES-401, Page 22 of 34 ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Emergency and Abnormal Plant Evolutions

-Tier 1/Group 1 (SRO) Form ES-401-1 E/APE # I Name I Safety Function K 1 K 2 K 3 A 1 A 2 G KJA Topic(s) IR # 295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 & 4 , , , 0 295003 Partial or Complete Loss of AC I 6 " ,', OZ.'37> Ability to determine operability and/or availability of safety related equipment 4,6 1 295004 Partial or Total Loss of DC Pwr 16 02. 22 Knowledge of limiting conditions for operations and safety limits, 4,7 1 295005 Main Turbine Generator Trip 13 0 295006 SCHAM I 1 0 295016 Control Room Abandonment 17 0 295018 Partial or Total Loss of CCW 18 0 3 Cause for partial or complete loss 3,5 1 295019 Partial or Total Loss of Ins\, Air 18 0 1" ,' , Instrument air system pressure 3,6 1 295021 Loss of Shutdown Cooling 14 , 0 295023 Refueling Acc I 8 0 295024 High Drywell Pressure I 5 0 295025 High Reactor Pressure I 3 0 295026 Suppression Pool High Water Temp, 5 0 295027 High Containment Temperature I 5 0 295028 High Drywell Temperature I 5 04.' ,,21 " ,,",vnovu,., t" , ""u 'UII'" u""u <u """""" the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system 4,6 1 295030 Low Suppression Pool Wtr Lvii 5 0 295031 Reactor Low Water Levell 2 0 295037 SCHAM Condition Present and Reactor Power Above APRM Downscale or Unknown 11 0 3 SBlC tank level 4,4 1 295038 Higll Off-site Release Rate 19 0: 4: Source of off-site release 4,5 1 600000 Plant Fire On Site I 8 0 700000 Generator Voltage and Electric Grid Disturbances I 6 0 KJA Category Totals: 0 0 0 0 '" 3 Group Point Total: 7 ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1/Group 1 (SRO) E/APE # I Name I Safety Function K K K A A G KJA Topic(s) IR # 1 2 3 1 2 .. 295001 Partial or Complete Loss of Forced 0 Core Flow Circulation I 1 & 4 .. .. OZ *. Ability to determine operability and/or availability of 295003 Partial or Complete Loss of AC I 6 37, ': safety related equipment.

4.6 1 295004 Partial or Total Loss of DC Pwr 16 02. Knowledge of limiting conditions for operations and 4.7 1 22 safety limits. 295005 Main Turbine Generator Trip 13 0 295006 SCHAM I 1 0 295016 Control Room Abandonment 17 0 295018 Partial or Total Loss of CCW 18 0 Cause for partial or complete loss 3.5 1 3 295019 Partial or Total Loss of Ins!. Air 18 0 Instrument air system pressure 3.6 1 1 , ... , 295021 Loss of Shutdown Cooling 14 0 295023 Refueling Acc I 8 0 295024 High Drywell Pressure I 5 0 295025 High Reactor Pressure I 3 0 295026 Suppression Pool High Water Temp. 0 5 295027 High Containment Temperature I 5 0 04,' '" 'v ...

v, "'" "D' D" ,,,,,,, D' 'v

'v 295028 High Drywell Temperature I 5 the status of safety functions, such as reactivity control, 4.6 1 .. 21 : core cooling and heat removal, reactor coolant system 295030 Low Suppression Pool Wtr Lvii 5 " 0 295031 Reactor Low Water Levell 2 0 295037 SCHAM Condition Present 0 and Reactor Power Above APRM SBlC tank level 4.4 1 Downscale or Unknown 11 3 295038 Higll Off-site Release Rate 19 0: 4: Source of off-site release 4.5 1 600000 Plant Fire On Site I 8 0 700000 Generator Voltage and Electric Grid 0 Disturbances I 6 KJA Category Totals: 0 0 0 0 '" 3 Group Point Total: 7 ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1/Group 2 (SRO) EIAPE # I Name I Safety Function 295002 Loss of Main Condenser Vac I 3 K 1 K 2 K 3 A 1 A 2 G KIA Topic(s) IR # 0 295007 High Reactor Pressure 13 0 295008 High Reactor Water Levell 2 0 295009 Low Reactor Water Levell 2 0 295010 High Drywell Pressure 15 0 295011 High Containment Temp I 5 0 295012 High Drywell Temperature 15 295013 High Suppression Pool Temp. 15 295014 Inadvertent Reactivity Addition 11 0 2 Localized heating/stratification 3.5 0 1 0 295015 Incomplete SCRAM 11 0 295017 High Off-site Release Rate I 9 295020 Inadvertent Cont. Isolation 15 & 7 295022 Loss of CRD Pumps 11 295029 High Suppression Pool Wtr LviI 5 295032 High Secondary Containment Area Temperature I 5 295033 High Secondary Containment Area Radiation Levels I 9 295034 Secondary Containment Ventilation High Radiation 19 295035 Secondary Containment High Differential Pressure I 5 0 3 ... .. 04. 06 Reactor power Knowledge of EOP mitigation strategies 3.7 4.7 0 1 0 0 0 0 1 0 295036 Secondary Containment High SumplArea Water Levell 5 0 500000 High CTMT Hydrogen Conc. 15 KIA Category Totals: 0 0 0 0 2 1 Group Point Total: 0 3 ES-401, Page 19 of 34 ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1/Group 2 (SRO) EIAPE # I Name I Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 295002 Loss of Main Condenser Vac I 3 0 295007 High Reactor Pressure 13 0 295008 High Reactor Water Levell 2 0 295009 Low Reactor Water Levell 2 0 295010 High Drywell Pressure 15 0 295011 High Containment Temp I 5 0 295012 High Drywell Temperature 15 0 295013 High Suppression Pool Temp. 15 0 Localized heating/stratification 3.5 1 2 295014 Inadvertent Reactivity Addition 11 0 295015 Incomplete SCRAM 11 0 295017 High Off-site Release Rate I 9 0 295020 Inadvertent Cont. Isolation 15 & 7 0 Reactor power 3.7 1 3 295022 Loss of CRD Pumps 11 0 ... 295029 High Suppression Pool Wtr LviI 5 0 .. 295032 High Secondary Containment Area 0 Temperature I 5 295033 High Secondary Containment Area 0 Radiation Levels I 9 295034 Secondary Containment Ventilation

04. Knowledge of EOP mitigation strategies 4.7 1 High Radiation 19 06 295035 Secondary Containment High 0 Differential Pressure I 5 295036 Secondary Containment High 0 SumplArea Water Levell 5 500000 High CTMT Hydrogen Conc. 15 0 KIA Category Totals: 0 0 0 0 2 1 Group Point Total: 3 ES-401, Page 19 of 34 ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems -Tier 2/Group 1 (SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K A 6 1 A 2 A A 3 4 KIA Topic(s) IR # 203000 FtHRlLPCI:

Injection 0 205000 Shutdown Cooling Mode 0 206000 HPCI 0 207000 Isolation (Emergency)

Condenser 0 209001 LPCS 0 209002 HPCS 0 211000 SLC 0 4 Inadequate system flow 3.4 1 212000 RPS 0 2150031RM

02. 40 Ability to apply Technical Specifications for a system 4.7 1 215004 Source Range Monitor . 0 215005 APRM / LPRM 0 217000 RCIC 0 218000 ADS 0 PCIS/Nuclear Steam Supply Shutoff 1 1 Standby liquid initiation 3.9 1 239002 SRVs 0 259002 Reactor Water Level Control 0 261000 SGTS 0 262001 AC Electrical Distribution 1 () Exceeding current limitations 3.4 1 UPS (ACIDC) 0 263000 DC Electrical Distribution

04. 09 Knowledge of low power/shutdown Implications in accid (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies ent 4.2 1 EDGs ... 0 300000 Instrument Air 0 400000 Component Cooling Water 0 0 KIA Category Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Total: 5 ES-401, Page 20 of 34 ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems -Tier 2/Group 1 (SRO) System # / Name K K K K K K A A A A KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 203000 FtHRlLPCI:

Injection 0 205000 Shutdown Cooling Mode 0 206000 HPCI 0 207000 Isolation (Emergency) 0 Condenser 209001 LPCS 0 209002 HPCS 0 211000 SLC 0 Inadequate system flow 3.4 1 4 212000 RPS 0 2150031RM

02. 40 Ability to apply Technical Specifications for a system 4.7 1 215004 Source Range Monitor . 0 215005 APRM / LPRM 0 217000 RCIC 0 218000 ADS 0 PCIS/Nuclear Steam Supply 1 3.9 1 Shutoff 1 Standby liquid initiation 239002 SRVs 0 259002 Reactor Water Level Control 0 261000 SGTS 0 262001 AC Electrical Distribution 1 () Exceeding current limitations 3.4 1 UPS (ACIDC) 0 04. Knowledge of low power/shutdown Implications in accident 263000 DC Electrical Distribution 09 (e.g., loss of coolant accident or loss of residual heat 4.2 1 removal) mitigation strategies EDGs ... 0 300000 Instrument Air 0 400000 Component Cooling Water 0 0 KIA Category Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Total: 5 ES-401, Page 20 of 34 ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Plant Systems -Tier 2/Group 2 (SRO) Form ES-401-1 System # I Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A <I A 3 A 4 G KIA Topic(s) IR # ;201001 CRD Hydraulic 0 201002 RMCS 0 201003 Control Rod and Drive Mechanism 0 701004 RSCS 0 201005 RCIS 0 201006 RWM 0 202001 Recirculation 0

Re,circulation Flow Control 0 RWCU 0 1714000 RPIS 0 1215001 Traversing In-core Probe 0 215002 RBM 0 1716000 Nuclear Boiler Ins!. 0 19000 RHR/LPCI:

Torus/Pool Cooling Mode 0 1223001 Primary CTMT and Aux. 0 1726001 RHR/LPCI:

C,TMT Spray Mode 02. 40 Ability to apply Technical Specifications for a system 4.7 1 1230000 RHR/LPCI:

Torus/Pool Spray Mode 0 1233000 Fuel Pool Cooling/Cleanup 0

Fuel Handling Equipment 0 1239001 Main and Reheat Steam 0 MSIV Leakage Control 0 741000 ReactorfTurbine Pressure Regulator 0 245000 Main Turbine Gen. / Aux. 0 256000 Reactor Condensate 0 259001 Reactor Feedwater 0 268000 Radwaste 0 271000 Offgas 0 272000 Radiation Monitoring 0

Fire Protection 0 1288000 Plant Ventilation 0 790001 Secondary CTMT 0 5 High area temperature 3.3 1 790003 Control Room HVAC 0 290002 Reactor Vessel Internals 0 II Exceeding safety limits 4.5 1 KIA Category Totals: 0 0 0 0 0 0 0 2 0 0 1" Group Point Total: 3 ES-401, Page 21 of 34 ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems -Tier 2/Group 2 (SRO) System # I Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1

<I 3 4 ;201001 CRD Hydraulic 0 201002 RMCS 0 201003 Control Rod and Drive Mechanism 0 701004 RSCS 0 201005 RCIS 0 201006 RWM 0 202001 Recirculation 0

Re,circulation Flow Control 0 RWCU 0 1714000 RPIS 0 1215001 Traversing In-core Probe 0 215002 RBM 0 1716000 Nuclear Boiler Ins!. 0 19000 RHR/LPCI:

Torus/Pool Cooling 0 Mode 1223001 Primary CTMT and Aux. 0 1726001 RHR/LPCI:

C,TMT Spray Mode 02. Ability to apply Technical Specifications for a system 4.7 1 40 1230000 RHR/LPCI:

Torus/Pool Spray Mode 0 1233000 Fuel Pool Cooling/Cleanup 0

Fuel Handling Equipment 0 1239001 Main and Reheat Steam 0 MSIV Leakage Control 0 741000 ReactorfTurbine Pressure Regulator 0 245000 Main Turbine Gen. / Aux. 0 256000 Reactor Condensate 0 259001 Reactor Feedwater 0 268000 Radwaste 0 271000 Offgas 0 272000 Radiation Monitoring 0

Fire Protection 0 1288000 Plant Ventilation 0 790001 Secondary CTMT 0 High area temperature 3.3 1 5 790003 Control Room HVAC 0 290002 Reactor Vessel Internals 0 Exceeding safety limits 4.5 1 II KIA Category Totals: 0 0 0 0 0 0 0 2 0 0 1" Group Point Total: 3 ES-401, Page 21 of 34 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility Name:Peach 8i.JLLUIII Date of Exam: 12/07/2009 Category KJA# Topic RO SRO-Only IR # IR # 2.1. 25 lAbility to interpret reference materials, such as graphs, curves, tables, etc. 3.9 1 2.1. 30 lAbility to locate and operate components, including local controls.

4.4 1 1. 2.1. 06 lAbility to manage the control room crew during plant transients.

4.8 1 Conduct of 2.1. 32 lAbility to explain and apply system limits and precautions. 4 1 r. '-',,","'I CLIVI 2.1. 2.1. I Subtotal 2.2. 01 IAbili pre-startup procedures for the facility, including operating those controls 4.5 1 lassociated with plant equipment that could affect reactivity.

2.2. 07 Knowledge of the process for conducting special or infrequent tests. 2.9 1 2. 2.2. 14 Knowledge of the process for controlling equipment configuration or status 3.9 1 Equipment 2.2. 43 Knowledge of the process used to track Inoperable alarms. 3.3 1 Control 2.2. 2.2. I Subtotal 3 2.3. 04 Knowledge of radiation exposure limits under normal or emergency conditions.

3.2 1 2.3. 05 Ability to use radiation mOnitoring systems, such as fixed radiation monitors and alarms, 2.9 1 portable survey instruments, personnel monitoring equipment, etc. 3. 2.3. 07 Ability to comply with radiation work permit requirements during normal or abnormal 3.5 1 Iditi< . Radiation 2.3. 11 Ability to control radiation releases.

4.3 1 Control I of I safety pertaining to licensed operator duties, such 2.3. 13 as , W' "g"""

• ern *"0. fuel handling 3.8 1 iii* . access to locked areas. alianina filters. etc. 2.3. Subtotal 3 2.4. 17 I of EOP terms and definitions.

3.9 1 2.4. 39 I ....

of RO responsibilities in emergency plan implementation 3.9 1 14. 2.4. 46 Ability to verify that the alarms are consistent with the plant conditions.

4.2 1 Emergency Ability perform without reference to procedures those actions that require immediate Procedures 2.4. 49 operation of system components and controls.

4.4 1 II Plan 2.4. 2.4. Subtotal 2 2 ITier 3 Point Total 10 7 ES-401, Page 27 of ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility Name:Peach 8i.JLLUIII Date of Exam: 12/07/2009 Category KJA# Topic RO SRO-Only IR # IR # 2.1. 25 lAbility to interpret reference materials, such as graphs, curves, tables, etc. 3.9 1 2.1. 30 lAbility to locate and operate components, including local controls.

4.4 1 1. 2.1. 06 lAbility to manage the control room crew during plant transients.

4.8 1 Conduct of 2.1. 32 lAbility to explain and apply system limits and precautions. 4 1 r. '-',,","'I CLIVI 2.1. 2.1. I Subtotal 2.2. 01 IAbili pre-startup procedures for the facility, including operating those controls 4.5 1 lassociated with plant equipment that could affect reactivity.

2.2. 07 Knowledge of the process for conducting special or infrequent tests. 2.9 1 2. 2.2. 14 Knowledge of the process for controlling equipment configuration or status 3.9 1 Equipment 2.2. 43 Knowledge of the process used to track Inoperable alarms. 3.3 1 Control 2.2. 2.2. I Subtotal 3 2.3. 04 Knowledge of radiation exposure limits under normal or emergency conditions.

3.2 1 2.3. 05 Ability to use radiation mOnitoring systems, such as fixed radiation monitors and alarms, 2.9 1 portable survey instruments, personnel monitoring equipment, etc. 3. 2.3. 07 Ability to comply with radiation work permit requirements during normal or abnormal 3.5 1 Iditi< . Radiation 2.3. 11 Ability to control radiation releases.

4.3 1 Control I of I safety pertaining to licensed operator duties, such 2.3. 13 as , W' "g"""

• ern * "0. fuel handling 3.8 1 iii* . access to locked areas. alianina filters. etc. 2.3. Subtotal 3 2.4. 17 I of EOP terms and definitions.

3.9 1 2.4. 39 I ....

of RO responsibilities in emergency plan implementation 3.9 1 14. 2.4. 46 Ability to verify that the alarms are consistent with the plant conditions.

4.2 1 Emergency Ability perform without reference to procedures those actions that require immediate Procedures 2.4. 49 operation of system components and controls.

4.4 1 II Plan 2.4. 2.4. Subtotal 2 2 ITier 3 Point Total 10 7 ES-401, Page 27 of 34 ES-401 Record of Rejected KlAs Form ES-401-4 Tier / Group Randomly Selected KIA Reason for Rejection RO 'I /1 0#53 295023 M2.05 Emergency Plan entry conditions are not required knowledge for ROs. (Replaced with KIA 295023 AA2.01) RO 1 /1 0#54 295004 G2.2.25 Tech Spec LCO bases are not required knowledge for ROs. (Replaced with KIA 295004 G2.2.22) RO 'I /1 0#58 295030 EA1.03 Peach Bottom does not have HPCS. (Replaced with KIA 295030 EA 1.01) RO 2/1 0#21 218000 G2.2.38 There are no conditions and limitations in the facility license associated with ADS. (Replaced with KIA 218000 G2.2.12) RO 2/1 0#25 262002 K4.01 Unable to construct another inverter question for this KlA. .. too similar to KIA 262002 K6.03 for Ouestion #12. (Replaced with KIA 239002 K4.05) RO 2/2 0#29 259001 K3.09 Unable to construct a question for this KIA -there is no significant effect on the Extraction Steam System from a Feedwater System malfunction. (Replaced with KIA 259001 K3.05) RO 3/2 0#68 G2.2.23 Tracking Tech Spec LCOs is not required knowledge for ROs. (Replaced with KIA G2.2.14) RO 3/2 0#75 G2.2.18 Managing maintenance (risk assessments, work prioritization, etc.) is not required knowledge for ROs. (Replaced with KIA G2.2.1) RO 3/4 0#73 G2.4.40 SRO responsibilities during emergency plan implementation are not required knowledge for ROs. (Replaced with KIA G2.4.39) SRO 1 /1 0#79 295025 G2.4.20 Unable to construct an SRO question for this KIA that meets the requirements of NUREG-1021. (Replaced with KIA 295004 G2.2.22) SRO 1 /2 0#84 295034 G2.4.49 Immediate operator actions are RO knowledge. (Replaced with KIA 295034 G2.4.6) SRO 1 /2 0#89 215003 G2.2.38 There are no conditions and/or limitations in the facility license associated with the IRM (WRNM) System. (Replaced with KIA 215003 G2.2.40) SRO 2 / 2 0#92 226001 G2.2.4 This KIA is not tied to 10CFR55.43(b), as required by NUREG-1021. (Replaced with KIA 226001 G2.2.40) SRO 3/3 0#94 G2.1.14 Unable to construct an SRO question for this KIA that meets the requirements of NUREG-1021 (ROs make plant announcements). (Replaced with KIA G2.1.32) SRO 3/3 0#98 G2.3.5 Not SRO-only; duplicate to KIA in Tier-3 RO section. (Replaced with KIA G2.3.13) ES-401, Page 27 of 33 ES-401 Record of Rejected KlAs Form ES-401-4 Tier / Randomly Group Selected Reason for Rejection KIA RO 'I /1 295023 Emergency Plan entry conditions are not required knowledge for ROs. 0#53 M2.05 (Replaced with KIA 295023 AA2.01) RO 1 /1 295004 Tech Spec LCO bases are not required knowledge for ROs. 0#54 G2.2.25 (Replaced with KIA 295004 G2.2.22) RO 'I /1 295030 Peach Bottom does not have HPCS. 0#58 EA1.03 (Replaced with KIA 295030 EA 1.01) RO 2/1 218000 There are no conditions and limitations in the facility license associated 0#21 G2.2.38 with ADS. (Replaced with KIA 218000 G2.2.12) RO 2/1 262002 Unable to construct another inverter question for this KlA. .. too similar to 0#25 K4.01 KIA 262002 K6.03 for Ouestion #12. (Replaced with KIA 239002 K4.05) RO 2/2 259001 Unable to construct a question for this KIA -there is no significant effect 0#29 K3.09 on the Extraction Steam System from a Feedwater System malfunction. (Replaced with KIA 259001 K3.05) RO 3/2 G2.2.23 Tracking Tech Spec LCOs is not required knowledge for ROs. 0#68 (Replaced with KIA G2.2.14) RO 3/2 G2.2.18 Managing maintenance (risk assessments, work prioritization, etc.) is not 0#75 required knowledge for ROs. (Replaced with KIA G2.2.1) RO 3/4 G2.4.40 SRO responsibilities during emergency plan implementation are not 0#73 required knowledge for ROs. (Replaced with KIA G2.4.39) SRO 1 /1 295025 Unable to construct an SRO question for this KIA that meets the 0#79 G2.4.20 requirements of NUREG-1021. (Replaced with KIA 295004 G2.2.22) SRO 1 /2 295034 Immediate operator actions are RO knowledge.

0#84 G2.4.49 (Replaced with KIA 295034 G2.4.6) SRO 1 /2 215003 There are no conditions and/or limitations in the facility license associated 0#89 G2.2.38 with the IRM (WRNM) System. (Replaced with KIA 215003 G2.2.40) SRO 2 / 2 226001 This KIA is not tied to 10CFR55.43(b), as required by NUREG-1021.

0#92 G2.2.4 (Replaced with KIA 226001 G2.2.40) SRO 3/3 Unable to construct an SRO question for this KIA that meets the 0#94 G2.1.14 requirements of NUREG-1021 (ROs make plant announcements). (Replaced with KIA G2.1.32) SRO 3/3 G2.3.5 Not SRO-only; duplicate to KIA in Tier-3 RO section. 0#98 (Replaced with KIA G2.3.13) ES-401, Page 27 of 33 Administrative Topics Outline Form ES-301-1 Facility:

Peach Date of Examination:

12/07/2009 Examination Level: RO SRO Operating Test Number: NRC Administrative Topic Type Describe activity to be performed (See Note) Code* G2.1.32 -Complete Attachment 1 of AO 3.8 "Evaluation of Conduct of Operations High CRD Temperature on Control Rod Scram Time" (PLOR-266C)

N, RIS G2.1.7 -Compliance with Asymmetric Feedwater Heating Conduct of Operations D, RIS Operation (AFTO) (PLOR-251 C) G2.2.41 -Isolate the 2A Turbine Building Closed Cooling Equipment Control Water Pump Due to a System Leak (P&ID M-316) (PLOR-257C)

(2008 NRC Exam) D, P, RIS N/A Radiation Control N/A G2.4.39 -Identify Errors on State and Local Notification N, RIS Emergency Plan Form -Return Form to SED for Correction (PLOR-341 CA) All items (5 total) are required for SROs. RO applicants require only 4 items unless they are iNOTE: retaking only the administrative topics, when 5 are required.

• Type Codes

& (C)ontrol room, (S)imulator, or (D)irect from bank (.:: 3 for ROs; .:: 4 for SROs & RO (N)ew or (M)odified from bank 1) (P)revious 2 exams (.:: 1; randomly selected)

ES 301, Page 22 of 27 ES-301 Administrative Topics Outline Form ES-301-1 Facility:

Peach Bottom Date of Examination:

12/07/2009 Examination Level: RO SRO D Operating Test Number: NRC Administrative Topic Type Describe activity to be performed (See Note) Code* G2.1.32 -Complete Attachment 1 of AO 3.8 "Evaluation of Conduct of Operations N, RIS High CRD Temperature on Control Rod Scram Time" (PLOR-266C)

Conduct of Operations D, RIS G2.1.7 -Compliance with Asymmetric Feedwater Heating Operation (AFTO) (PLOR-251 C) G2.2.41 -Isolate the 2A Turbine Building Closed Cooling Equipment Control D, P, RIS Water Pump Due to a System Leak (P&ID M-316) (PLOR-257C)

(2008 NRC Exam) Radiation Control N/A N/A Emergency Plan N, RIS G2.4.39 -Identify Errors on State and Local Notification Form -Return Form to SED for Correction (PLOR-341 CA) I NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (.:: 3 for ROs; .:: 4 for SROs & RO retakes) (N)ew or (M)odified from bank 1) (P)revious 2 exams (.:: 1; randomly selected)

ES 301, Page 22 of 27 II Administrative OD!ICS Outline Form ES-301-1 Facility:

Peach Bottom Date of Examination:

12/07/2009 Examination Level: RO SRO r8J Operating Test Number: NRC Administrative Topic (See Note) Conduct of Operations Conduct of Operations Equipment Control Radiation Control Type Code* Describe activity to be performed N, RIS i G2.1.32 Review Attachment 1 of AO 3.8 "Evaluation of High CRD Temperature on Control Rod Scram Time" and Identify and declare SLOW Control Rod (PLOR-340CA SRO) D, RIS

• G2.1.7 -Compliance with Asymmetric Feedwater Heating Operation (AFTO) (PLOR-252C)

D, P, RIS G2.2.21 -Determination of Required Post-Maintenance Testing (PLOR-242C)

(2007 NRC Exam) M, RIS G2.3.14 -Review and Authorize Issuance of Thyroid

• Blocking Agent (KI) (PLOR-215C)

I I G2.4.41 EAL Classification with State and Local Emergency Plan D, RIS Notifiations

-Alert due to RPS Failure (PLOR-233C)

I All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank 3 for ROs; 4 for SROs & RO retakes) (N)ew or (M)odified from bank 1) (P)revious 2 exams 1; randomly selected)

ES 301) Page 22 of 27 ES-301 Administrative Outline Form ES-301-1 Facility:

Peach Bottom Date of Examination:

12/07/2009 Examination Level: RO SRO r8J Operating Test Number: NRC Administrative Topic Type Describe activity to be performed (See Note) Code* i G2.1.32 Review Attachment 1 of AO 3.8 "Evaluation of Conduct of Operations N, RIS High CRD Temperature on Control Rod Scram Time" and Identify and declare SLOW Control Rod (PLOR-340CA SRO) Conduct of Operations D, RIS

• G2.1.7 -Compliance with Asymmetric Feedwater Heating Operation (AFTO) (PLOR-252C)

Equipment Control D, P, RIS G2.2.21 -Determination of Required Post-Maintenance Testing (PLOR-242C)

(2007 NRC Exam) II Radiation Control M, RIS G2.3.14 -Review and Authorize Issuance of Thyroid

• Blocking Agent (KI) (PLOR-215C)

I I Emergency Plan D, RIS G2.4.41 EAL Classification with State and Local Notifiations

-Alert due to RPS Failure (PLOR-233C)

I NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank 3 for ROs; 4 for SROs & RO retakes) (N)ew or (M)odified from bank 1) (P)revious 2 exams 1; randomly selected)

ES 301) Page 22 of 27 ES-301 Control Room/In-Plant Sv,=.t".,rnct.

Outline Form ES-301-2 Facility:

Peach Bottom Date of Examination:

12/07/2009 Exam Level: RO [8J SRO-I D SRO-U D Operating Test Number: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function a. 295037 EA 1.04 -Standby Liquid Control System I Inject SBLC (Alternate A, 0, EN, 1 Path Low SBLC Discharge Pressure) (PLOR-331 CAl [Set 1] L, S b. 217000 A4.03 -Reactor Core Isolation Cooling System I Manually A, 0, EN, 2 Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

L, S [Set 2] -"" . c. 239001 A4.01 -Reopen Main Steam Isolation Valves after a Group I D,L, P,S 3 ! Isolation (2008 NRC Exam) [Set 3] d. 206000 A2.09 -High Pressure Coolant Injection System I Manual A,L, N,S 4 System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low CST Level) (PLOR-343CA)

[Set 1] e. 223001 A4.10 -Primary Containment System and Auxiliaries 1 CAD D,S 5 System Nitrogen Addition To Containment During Normal Operations (PL.OR-034C)

[Set 2] f. 264000 A4.04 -Emergency Generators I Diesel Generator Load Test I A,D,S 6 'I (Alternate Path -Load Control Difficulti'} (PLOR-322CAl

[Set 4] !i g. 400000 A4.01 -Component Cooling Water System I ECW System N,S 8 . Makeup to Tower Using ESW System (PLOR-270C)

[Set 3] 261000 A4.03 Standby Gas Treatment (SBGT) System I Manually Start D,S 9 SBGT System on Equipment Cell Exhaust

[Set 4] "" In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. 239001 A2.12 -Main and Reheat Steam System 1 Close A Stuck Open A, D, L,R 4 I MSIV (Alternate Path -Removing Fuse Fails To Close MSIV -Unit 3) (pLOR-313PA}

I j. 295037 EA1.01 -Reactor Protection System I Scram Solenoid De-C, 0, E, 7

-Unit 2 (T-213-2) (PLOR-07§P)

L, R k. 2180000 A2.03 -Backup Instrument Nitrogen to ADS System Startup and D,E, R 3 Operation 3} (PLOR-271 P) @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlaQ those tested in the control room.

• Type Codes I Criteria for RO 1 SRO-II SRO-U ES-301 Control Room/In-Plant Outline Form ES-301-2 Facility:

Peach Bottom Date of Examination:

12/07/2009 Exam Level: RO [8J SRO-I D SRO-U D Operating Test Number: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function a. 295037 EA 1.04 -Standby Liquid Control System I Inject SBLC (Alternate A, 0, EN, 1

• Path Low SBLC Discharge Pressure) (PLOR-331 CAl [Set 1] L, S b. 217000 A4.03 -Reactor Core Isolation Cooling System I Manually A, 0, EN, 2 Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

L, S [Set 2] _ ..

• c. 239001 A4.01 -Reopen Main Steam Isolation Valves after a Group I D,L, P,S 3 ! .. Isolation (2008 NRC Exam) [Set 3] d. 206000 A2.09 -High Pressure Coolant Injection System I Manual A,L, N,S 4 System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low CST Level) (PLOR-343CA)

[Set 1] e. 223001 A4.10 -Primary Containment System and Auxiliaries 1 CAD D,S 5 System Nitrogen Addition To Containment During Normal Operations (PL.OR-034C)

[Set 2] f. 264000 A4.04 -Emergency Generators I Diesel Generator Load Test I A,D,S 6 'I (Alternate Path -Load Control Difficulti'} (PLOR-322CAl

[Set 4] !i g. 400000 A4.01 -Component Cooling Water System I ECW System N,S 8

• Makeup to Tower Using ESW System (PLOR-270C)

[Set 3] 261000 A4.03 -Standby Gas Treatment (SBGT) System I Manually Start D,S 9 SBGT System on Equipment Cell Exhaust

[Set 4] .. In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. 239001 A2.12 -Main and Reheat Steam System 1 Close A Stuck Open A, D, L,R 4 I MSIV (Alternate Path -Removing Fuse Fails To Close MSIV -Unit 3) (pLOR-313PA}

I j. 295037 EA1.01 -Reactor Protection System I Scram Solenoid De-C, 0, E, 7

-Unit 2 (T-213-2) (PLOR-07§P)

L, R k. 2180000 A2.03 -Backup Instrument Nitrogen to ADS System Startu D,E, R 3 Operation 3} (PLOR-271 P) @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlaQ those tested in the control room.

• Type Codes I Criteria for RO 1 SRO-II SRO-U (A)lternate path 4-6 I 4-6 I (C)ontrol room (D)irect from bank <9 I .:: 8 I .:: (E)mergency or abnormal in-plant > 1 I 1 I (EN)gineered safety feature I -I 1 (control room system) (Wow-PowerIShuWown > 1 I > 1 I (N)ew or (M)odified from bank including 1 (A) >2 I I (P)revious 2 exams .::3 I .:: 3 I .:: 2 (randomly (R)CA > 1 I 1 I (S)imulator ES-301, Page 23 of (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank <9 I .:: 8 I .:: 4 (E)mergency or abnormal in-plant > 1 I 1 I 1 (EN)gineered safety feature I -I 1 (control room system) (Wow-PowerIShuWown > 1 I > 1 I 1 (N)ew or (M)odified from bank including 1 (A) >2 I I 1 (P)revious 2 exams .::3 I .:: 3 I .:: 2 (randomly selected) (R)CA > 1 I 1 I 1 (S)imulator ES-301, Page 23 of 27 ES-301 Control Roomlln-Plant

!ii;\I'l:I;tArne.

Outline Form ES-301-2 Facility:

Peach Bottom Date of Examination:

12/07/2009 EXam Level: RO D SRO-I I:8J SRO-U D Operating Test Number: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function a. 295037 EA 1.04 -Standby Liquid Control System 1 Inject SBLC {Alternate A, D, EN, 1 Path -Low SBLC Discharge Pressure}

1] L, S b. 217000 A4.03 Reactor Core Isolation Cooling System 1 Manually A, D, EN, 2 Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

L, S [Set 2] c. 239001 A4.01 -Reopen Main Steam Isolation Valves after a Group I D, L, P,S 3 Isolation (PLOR-083C)

(2008 NRCJ:;>>(am}

[Set 3] d. 206000 A2.09 -High Pressure Coolant Injection System 1 Manual A,L,N,S 4 System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low CST Level)

[Set 1] e . . f. 264000 A4.04 -Emergency Generators 1 Diesel Generator Load Test A, D, S (Alternate Path -Load Control Difficulty) (PLOR-322CA)

[Set 4] :' 400000 A4.01 -Component Cooling Water System 1 ECW System N,S 8 Makeup to Tower Using ESW System (PLOR-270C)

[Set 3] h. 261000 A4.03 -Standby Gas Treatment (SBGT) System 1 Manual .... SBGT Syste ....' ent Cell Exhaust (PLOR-018C)

[Set 4]

• D,S 9 i In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. 239001 A2.12 -Main and Reheat Steam System 1 Close A Stuck Open A, D, L, R 4 MSIV (Alternate Path -Removing Fuse Fails To Close MSIV -Unit 3) (PLOR-313PA)

J' 5037 EA 1.01 -Reactor Protection System 1 Scram Solenoid De-C, D, E, 7 energization -Unit 2 (T-213-2) (PLOR-075P)

L, R k. 2180000 A2.03 -Backup Instrument Nitrogen to ADS System St D,E, R 3 Operation (Unit 3) (PLOR-271 P) @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may ClYf?rlal>

those tested in the control room,

• Type Codes I Criteria for RO 1 SRO-II SRO-U ES-301 Control Roomlln-Plant

!ii;\I'l:I;tArne.

Outline Form ES-301-2 Facility:

Peach Bottom Date of Examination:

12/07/2009 EXam Level: RO D SRO-I I:8J SRO-U D Operating Test Number: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function a. 295037 EA 1.04 -Standby Liquid Control System 1 Inject SBLC {Alternate A, D, EN, 1 Path -Low SBLC Discharge Pressure}

1] L, S b. 217000 A4.03 Reactor Core Isolation Cooling System 1 Manually A, D, EN, 2 Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

L, S [Set 2] c. 239001 A4.01 -Reopen Main Steam Isolation Valves after a Group I D, L, P,S 3 Isolation (PLOR-083C)

(2008 NRCJ:;>>(am}

[Set 3] d. 206000 A2.09 -High Pressure Coolant Injection System 1 Manual A,L,N,S 4 System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low CST Level)

[Set 1] e . . f. 264000 A4.04 -Emergency Generators 1 Diesel Generator Load Test A, D, S (Alternate Path -Load Control Difficulty) (PLOR-322CA)

[Set 4] :' 400000 A4.01 -Component Cooling Water System 1 ECW System N,S 8 Makeup to Tower Using ESW System (PLOR-270C)

[Set 3] h. 261000 A4.03 -Standby Gas Treatment (SBGT) System 1 Manual .... SBGT Syste ....' ent Cell Exhaust (PLOR-018C)

[Set 4]

• D,S 9 i In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. 239001 A2.12 -Main and Reheat Steam System 1 Close A Stuck Open A, D, L, R 4 MSIV (Alternate Path -Removing Fuse Fails To Close MSIV -Unit 3) (PLOR-313PA)

J' 5037 EA 1.01 -Reactor Protection System 1 Scram Solenoid De-C, D, E, 7 energization -Unit 2 (T-213-2) (PLOR-075P)

L, R k. 2180000 A2.03 -Backup Instrument Nitrogen to ADS System St D,E, R 3 Operation (Unit 3) (PLOR-271 P) @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may ClYf?rlal>

those tested in the control room,

• Type Codes I Criteria for RO 1 SRO-II SRO-U II (A)lternate path (C)ontrol room (D)irect from (E)mergency or abnormal (EN)gineered safety feature (L)ow-Power

/ Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams 4-6 I 4-6 <9 / .:': 8 > 1 / .:::. 1 / -> 1 / > 1 .:::.2 / > 2 <3 / .:': 3 > 1 / > 1 I 2-3 / .:': 4 / > 1 / > 1 (control room system) / > 1 / > 1 / .:': 2 (randomly selected) / > 1 ES-301, Page 23 of 27 (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank <9 / .:': 8 / .:': 4 (E)mergency or abnormal in-plant > 1 / .:::. 1 / > 1 (EN)gineered safety feature / -/ > 1 (control room system) (L)ow-Power

/ Shutdown > 1 / > 1 / > 1 (N)ew or (M)odified from bank including 1(A) .:::.2 / > 2 / > 1 (P)revious 2 exams <3 / .:': 3 / .:': 2 (randomly selected)

II (R)CA > 1 / > 1 / > 1 (S)imulator ES-301, Page 23 of 27 ES-301 Control Room/In-Plant SV'l::tArnc Outline Form ES-301-2 Facility:

Peach Bottom Date of Examination:

12/07/2009 Exam Level: RO SRO-I D SRO-U fZJ Operating Test Number: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function a. b. 217000 A4.03 -Reactor Core Isolation Cooling System I Manually A, D, EN, 2 Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

L,S [Set 21 c. d. 206000 A2.09 -High Pressure Coolant Injection System I Manual A, L,N, S 4 System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low CST Level) (PLOR-343CA)

[Sl?tJ1 e. f. g. h. 261000 A4.03 -Standby Gas Treatment (SBGT) System 1 Manually Start D,S 9 SBGT System on Equipment Cell Exhaust (PLOR-018C)

[Set 4] In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. j. 295037 EA 1.01 -Reactor Protection System 1 Scram Solenoid De-C, D, E, 7 energizati()rl:Unit 2 (T ...(PLOR-075P)

L, t< I k. 2180000 A2.03 -Backup Instrument Nitrogen to ADS System Startup and D, Operation (Unit 3) (PLOR-271 P) @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U "Y"l",llI", must serve different safety functions; in-plant systems and functions may Vy",**co .... those tested in the control room

• Type Codes Criteria for RO 1 SRO-I 1 SRO-U (A)lternate path 4-6 1 4-6 1 2-3 (C}ontrol room (D}irect from bank :5.9 1 :5. 8 I :5. 4 (E}mergency or abnormal in-plant > 1 1 > 1 1 1 (EN}gineered safety feature 1 1 (control room system) (L)ow-Power 1 Shutdown > 1 I > 1 I 1 (N}ew or (M}odified from bank including 1 (A) I I 1 (P}revious 2 exams <3 I < 3 I :5. 2 (randomly selected) (R)CA > II -(S}imulator ES-301 Control Room/In-Plant SV'l::tArnc Outline Form ES-301-2 Facility:

Peach Bottom Date of Examination:

12/07/2009 Exam Level: RO SRO-I D SRO-U fZJ Operating Test Number: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function a. b. 217000 A4.03 -Reactor Core Isolation Cooling System I Manually A, D, EN, 2 Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

L,S [Set 21 c. d. 206000 A2.09 -High Pressure Coolant Injection System I Manual A, L,N, S 4 System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low CST Level) (PLOR-343CA)

[Sl?tJ1 e. f. g. h. 261000 A4.03 -Standby Gas Treatment (SBGT) System 1 Manually Start D,S 9 SBGT System on Equipment Cell Exhaust (PLOR-018C)

[Set 4] In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. j. 295037 EA 1.01 -Reactor Protection System 1 Scram Solenoid De-C, D, E, 7 energizati()rl:Unit 2 (T ... (PLOR-075P)

L, t< I k. 2180000 A2.03 -Backup Instrument Nitrogen to ADS System Startup and D, Operation (Unit 3) (PLOR-271 P) @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U "Y"l",llI", must serve different safety functions; in-plant systems and functions may Vy", ** co .... those tested in the control room

• Type Codes Criteria for RO 1 SRO-I 1 SRO-U (A)lternate path 4-6 1 4-6 1 2-3 (C}ontrol room (D}irect from bank :5.9 1 :5. 8 I :5. 4 (E}mergency or abnormal in-plant > 1 1 > 1 1 1 (EN}gineered safety feature 1 1 (control room system) (L)ow-Power 1 Shutdown > 1 I > 1 I 1 (N}ew or (M}odified from bank including 1 (A) I I 1 (P}revious 2 exams <3 I < 3 I :5. 2 (randomly selected) (R)CA > II -(S}imulator Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #1 (new} OpTest No. 2009 NRC Examiners Operators CRS (SRO) URO (ATC) PRO (BOP) Scenario The scenario begins with the reactor at approximately 5% power during a reactor startup. Following Summary shift turnover, the crew is directed to secure drywell purge in preparation for inerting the drywell. Once drywell purge is secured, the crew should continue with the reactor startup by pulling control rods in accordance with the approved startup sequence.

During this evolution a control rod will drift out, requiring the crew to execute ON-121 "Drifting Control Rod" and declare the affected control rod inoperable in accordance with Tech Specs. After the actions for the drifting control rod are complete, the 'B' drywell chiller will trip. The crew should place a standby drywell chiller in service in accordance with the system operating procedure.

Next, a blown fuse will cause an ARI power supply failure, requiring the crew to initiate repairs and evaluate ARI-RPT operability per Tech Specs. This will be followed by an APRM trip with an auto scram failure, requiring the crew to initiate a manual scram. A SULCV failure will complicate RPV level control post-scram.

Following the scram, a leak will develop in the torus, requiring the crew to enter T-103 "Secondary Containment Control" and T-102 "Primary Containment Control".

A failure of the turbine bypass jack will require the crew to use alternate methods to depressurize the reactor in accordance with T-101 "RPV Control".

Torus level will continue to lower to the point where the crew will be required to perform T-112 "Emergency Blowdown".

Initial IC-91, 5% power Conditions Turnover See Attached "Shift Turnover" Sheet Event No. Malfunction No. Event Type* Event Description 1 N PRO CRS Secure drywell purge 2 R URO CRS Power ascension with control rods 3 C URO TS PRO CRS Drifting control rod (Tech Spec) 4 C PRO CRS Drywell chiller trip I place standby chiller in service 5 TS CRS ARI power supply failure (Tech Spec) 6 I URO CRS APRM trip with auto scram failure I manual reactor scram 7 M ALL Torus leak into secondary containment I emergency blowdown 8 I URO CRS Startup level control valve (LCV-8091) failure 9 C PRO CRS Turbine bypass jack fails (prevents rapid depressurization to the main condenser)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #1 (new} OpTest No. 2009 NRC Examiners Operators CRS (SRO) URO (ATC) PRO (BOP) Scenario The scenario begins with the reactor at approximately 5% power during a reactor startup. Following Summary shift turnover, the crew is directed to secure drywell purge in preparation for inerting the drywell. Once drywell purge is secured, the crew should continue with the reactor startup by pulling control rods in accordance with the approved startup sequence.

During this evolution a control rod will drift out, requiring the crew to execute ON-121 "Drifting Control Rod" and declare the affected control rod inoperable in accordance with Tech Specs. After the actions for the drifting control rod are complete, the 'B' drywell chiller will trip. The crew should place a standby drywell chiller in service in accordance with the system operating procedure.

Next, a blown fuse will cause an ARI power supply failure, requiring the crew to initiate repairs and evaluate ARI-RPT operability per Tech Specs. This will be followed by an APRM trip with an auto scram failure, requiring the crew to initiate a manual scram. A SULCV failure will complicate RPV level control post-scram.

Following the scram, a leak will develop in the torus, requiring the crew to enter T-103 "Secondary Containment Control" and T-102 "Primary Containment Control".

A failure of the turbine bypass jack will require the crew to use alternate methods to depressurize the reactor in accordance with T-101 "RPV Control".

Torus level will continue to lower to the point where the crew will be required to perform T-112 "Emergency Blowdown".

Initial IC-91, 5% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 N PRO Secure drywell purge CRS 2 R URO Power ascension with control rods CRS 3 C URO Drifting control rod (Tech Spec) TS PRO CRS 4 C PRO Drywell chiller trip I place standby chiller in service CRS 5 TS CRS ARI power supply failure (Tech Spec) 6 I URO APRM trip with auto scram failure I manual reactor scram CRS 7 M ALL Torus leak into secondary containment I emergency blowdown 8 I URO Startup level control valve (LCV-8091) failure CRS 9 C PRO Turbine bypass jack fails (prevents rapid depressurization to the main CRS condenser)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec Scenario ES-D-1 Simulation Facility Peach Bottom Scenario No. #2 (new) OpTestNo.

2009 NRC Examiners Operators CRS (SRO) LlRO (ATC) PRO (BOP)

The scenario begins with the reactor at 100% power. After taking the shift the crew is required to verify operability of the Startup Source load tap changer due to an earlier thunderstorm.

Shortly after this, the running CRD pump will trip, requiring the crew to execute ON-107 "loss of CRD Regulating Function" and place the standby CRD pump in service. Additional thunderstorms in the area will result in a loss of the SBO line, which will require the CRS to enter and evaluate the TRM. Following this, the '0' SRV will inadvertently open, requiring the crew to take actions to close the valve in accordance with OT-114 "Inadvertent Opening of a Relief Valve". Power will be reduced in accordance with GP-9-2 "Fast Power Reduction", and the crew will be successful in closing the SRV by lowering reactor pressure in accordance with OT-114. Next, a sustained loss of Stator Cooling will occur, requiring the crew to scram the reactor. An A TWS (electrical) will require the crew to execute T -101 "RPV Control" and T -117 "level/Power Control".

The main turbine will trip several minutes into this event as a result of the loss of Stator Cooling, complicating the crew's efforts to respond to the A TWS and challenging Primary Containment due to SRVactuation.

When SBlC is initiated, RWCU will fail to automatically isolate, requiring the crew to manually isolate RWCU. In addition, the crew will not be able to restore normal instrument nitrogen, which will require aligning a backup source of nitrogen to the SRVs to ensure they are available for reactor pressure control. After RPV level has been lowered to control power, the ATWS will be terminated using T-214 "Venting the Scram Air Header". Initial IC-92, 100% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event No. Type*No. 1 PRO CRS 2 URO CRS 3 TS CRS 4 C PRO TS CRS 5 URO CRS 6 All 7 URO CRS 8 C PRO CRS ..

Verify operability of Startup Source load tap changer CRD pump trip I place standby CRD pump in service SBO line failure (TRM) SRV inadvertently opens (Tech Spec) I maximize torus cooling Fast power reduction I pressure reduction due to SRV failure loss of stator cooling water I scram (electric ATWS) RWCU fails to isolate on SBlC initiation I manually isolate RWCU Unable to restore drywell instrument nitrogen I place alternate instrument nitrogen system(s) in service .. (N)ormal, (R)eactlvlty, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #2 (new) OpTestNo.

2009 NRC Examiners Operators CRS (SRO) LlRO (ATC) PRO (BOP) Scenario The scenario begins with the reactor at 100% power. After taking the shift the crew is required to Summary verify operability of the Startup Source load tap changer due to an earlier thunderstorm.

Shortly after this, the running CRD pump will trip, requiring the crew to execute ON-107 "loss of CRD Regulating Function" and place the standby CRD pump in service. Additional thunderstorms in the area will result in a loss of the SBO line, which will require the CRS to enter and evaluate the TRM. Following this, the '0' SRV will inadvertently open, requiring the crew to take actions to close the valve in accordance with OT-114 "Inadvertent Opening of a Relief Valve". Power will be reduced in accordance with GP-9-2 "Fast Power Reduction", and the crew will be successful in closing the SRV by lowering reactor pressure in accordance with OT-114. Next, a sustained loss of Stator Cooling will occur, requiring the crew to scram the reactor. An A TWS (electrical) will require the crew to execute T -101 "RPV Control" and T -117 "level/Power Control".

The main turbine will trip several minutes into this event as a result of the loss of Stator Cooling, complicating the crew's efforts to respond to the A TWS and challenging Primary Containment due to SRVactuation.

When SBlC is initiated, RWCU will fail to automatically isolate, requiring the crew to manually isolate RWCU. In addition, the crew will not be able to restore normal instrument nitrogen, which will require aligning a backup source of nitrogen to the SRVs to ensure they are available for reactor pressure control. After RPV level has been lowered to control power, the ATWS will be terminated using T-214 "Venting the Scram Air Header". Initial IC-92, 100% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 N PRO Verify operability of Startup Source load tap changer CRS 2 C URO CRD pump trip I place standby CRD pump in service CRS 3 TS CRS SBO line failure (TRM) 4 C PRO SRV inadvertently opens (Tech Spec) I maximize torus cooling TS CRS 5 R URO Fast power reduction I pressure reduction due to SRV failure CRS 6 M All loss of stator cooling water I scram (electric ATWS) 7 I URO RWCU fails to isolate on SBlC initiation I manually isolate RWCU CRS 8 C PRO Unable to restore drywell instrument nitrogen I place alternate CRS instrument nitrogen system(s) in service .. .. (N)ormal, (R)eactlvlty, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec Scenario ES-D-1 Simulation Facility Peach Bottom Scenario No. #3 (modified)

Op Test No. 2009 NRC Operators CRS (SRO) URO (ATC) PRO (BOP)

The scenario begins with the reactor at approximately 88% power. Following shift turnover, the crew will perform ST-0-001-200-2 "Turbine Stop Valve Closure and EOC-RPT Functional Test". An RPS failure during the test will require the crew to make a Tech Spec declaration.

Next, the running Service Water pump will trip on overcurrent, requiring the crew to place the standby pump in service using the system operating procedure.

Following this, a dryweil pressure instrument will fail upscale without causing the expected half scram. The crew will apply Tech Specs and (with time-compression) insert a half scram lAW GP-25 "Installation of Trips/lsolations to Satisfy Tech SpecfTRM Requirements".

When this is complete, the 'N Condensate pump will trip without the expected Recirc System runback. Power must be manually reduced using recirc flow to prevent a low-level scram. When conditions have stabilized, #2 Auxiliary Sus will trip on overcurrent, causing a loss of the remaining Condensate pumps. An RPS failure will prevent the automatic and manual scrams, requiring entry into T-101 "RPV Control" and the use of Alternate Rod Insertion (ARI) to shutdown the reactor. HPCI will trip shortly after it starts and will not be recoverable.

A small reactor coolant leak inside the drywell will be greater than the capacity of RCIC (the only remaining high-pressure feed source) and require the use of containment sprays. The crew should enter T-111 "Level Restoration" and T-102 "Primary Containment Control".

A failure of the RCfC flow controller will complicate efforts to feed with RCIC and require the operator to transfer RCfC control to manual. A containment spray logic failure will complicate the crew's efforts to spray containment; the other loop of RHR will be available and should be used to spray containment As level deteriorates, the crew should start available low pressure ECCS pumps and when it is determined that level cannot be restored and maintained above -172 inches, the reactor should be depressurized in accordance with T-112 "Emergency Slowdown".

Low pressure ECCS will be available to recover reactor level. Initial IC-93, 88% power Conditions i Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event No. Type" Description 1 N PRO Main Turbine stop valve functional test 1 RPS failure (Tech Spec) TS CRS 2 Service Water pump trip 1 manual start of the standby pump CRS C URO 3 Orywe/l pressure instrument fails upscale without the expected half TS CRS I PRO scram (Tech Spec) 1 insert half scram lAW GP-25 4 Condensate pump trip with recirc runback failure 1 power reduction CRS R URO 5 Loss of #2 auxiliary bus 1 loss of condensate

& feedwater I reactor coolant leak inside the drywell M ALL 6 C URO RPS failure requires ARI to scram the reactor CRS 7 HPCI trip 1 RCIC flow controller fails in automatic (transfer to manual CRS I URO control) 8 I PRO Containment spray logic failure hampers effort to spray the CRS containment, requiring crew to use alternate RHR loop .. " (N)ormal, (R)eacbvlty, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #3 (modified)

Op Test No. 2009 NRC Examiners Operators CRS (SRO) URO (ATC) PRO (BOP) Scenario The scenario begins with the reactor at approximately 88% power. Following shift turnover, the crew will Summary perform ST-0-001-200-2 "Turbine Stop Valve Closure and EOC-RPT Functional Test". An RPS failure during the test will require the crew to make a Tech Spec declaration.

Next, the running Service Water pump will trip on overcurrent, requiring the crew to place the standby pump in service using the system operating procedure.

Following this, a dryweil pressure instrument will fail upscale without causing the expected half scram. The crew will apply Tech Specs and (with time-compression) insert a half scram lAW GP-25 "Installation of Trips/lsolations to Satisfy Tech SpecfTRM Requirements".

When this is complete, the 'N Condensate pump will trip without the expected Recirc System runback. Power must be manually reduced using recirc flow to prevent a low-level scram. When conditions have stabilized, #2 Auxiliary Sus will trip on overcurrent, causing a loss of the remaining Condensate pumps. An RPS failure will prevent the automatic and manual scrams, requiring entry into T-101 "RPV Control" and the use of Alternate Rod Insertion (ARI) to shutdown the reactor. HPCI will trip shortly after it starts and will not be recoverable.

A small reactor coolant leak inside the drywell will be greater than the capacity of RCIC (the only remaining high-pressure feed source) and require the use of containment sprays. The crew should enter T-111 "Level Restoration" and T-102 "Primary Containment Control".

A failure of the RCfC flow controller will complicate efforts to feed with RCIC and require the operator to transfer RCfC control to manual. A containment spray logic failure will complicate the crew's efforts to spray containment; the other loop of RHR will be available and should be used to spray containment As level deteriorates, the crew should start available low pressure ECCS pumps and when it is determined that level cannot be restored and maintained above -172 inches, the reactor should be depressurized in accordance with T-112 "Emergency Slowdown".

Low pressure ECCS will be available to recover reactor level. Initial IC-93, 88% power Conditions i Turnover See Attached "Shift Turnover" Sheet -Event Malfunction Event Event No. No. Type" Description 1 N PRO Main Turbine stop valve functional test 1 RPS failure (Tech Spec) TS CRS 2 C URO Service Water pump trip 1 manual start of the standby pump CRS 3 I PRO Orywe/l pressure instrument fails upscale without the expected half TS CRS scram (Tech Spec) 1 insert half scram lAW GP-25 4 R URO Condensate pump trip with recirc runback failure 1 power reduction CRS 5 M ALL Loss of #2 auxiliary bus 1 loss of condensate

& feedwater I reactor coolant leak inside the drywell 6 C URO RPS failure requires ARI to scram the reactor CRS 7 I URO HPCI trip 1 RCIC flow controller fails in automatic (transfer to manual CRS control) 8 I PRO Containment spray logic failure hampers effort to spray the CRS containment, requiring crew to use alternate RHR loop .. " (N)ormal, (R)eacbvlty, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec