ML093340503

Peach Bottom Draft - Outlines with Facility Ltr. Dtd. 10/19/09 (Folder 2)

ML093340503
Person / Time
Site:	Peach Bottom
Issue date:	01/18/2010
From:	Fish T H Operations Branch I
To:	Exelon Generation Co, Exelon Nuclear
Hansell S
References
50-277/09-302, 50-278/09-302, TAC U01792
Download: ML093340503 (24)

Text

Peach Bottom License No.: DPR-44, DPR-56 Docket No.: 50-277, 50-278 Operator Licensing Exam Dates: 12/07-12/18/09 Peach Bottom Draft -Outlines with Facility Letter (Folder 2) Chief Examiner:

T. Fish TAC No. U01792 Report No.: 50-277/50-278/09-302 Public and Non-Sensitive NRR-079, SUNSI Review Complete ADAMS Package Accession No. ML091380320 G:\ORS\OPERATIONS BRANCH\BIXLER\EXAMMASTERFOLOERS\FY1 0\12-07 -09PB.DOC Rev. 10/06/09 Peach Bottom License No.: DPR-44, DPR-56 Docket No.: 50-277, 50-278 Operator Licensing Exam Dates: 12/07 -12/18/09 Peach Bottom Draft -Outlines with Facility Letter (Folder 2) Chief Examiner:

T. Fish TAC No. U01792 Report No.: 50-277/50-278/09-302 Public and Non-Sensitive NRR-079, SUNSI Review Complete ADAMS Package Accession No. ML091380320 G:\ORS\OPERATIONS BRANCH\BIXLER\EXAMMASTERFOLOERS\FY1 0\12-07 -09PB.DOC Rev. 10/06/09 Exelon Nuclear wwwexeloncorp.com Peach Bottom Atomic Power Station Nuclear 1848 lay Road 10 CPR 55.40 Delta. PA 173'4-9032 October 19, 2009 Mr. Samuel J. Regional U. S. Nuclear Regulatory Commission Region 475 Allendale King of Prussia, PA Peach Bottom Atomic Power Station (pBAPS), Units 2 and 3 Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Submittal of Integrated Initial License Training Examination Materials Enclosed are the examination materials, which Peach Bottom Atomic Power Station is submitting in support of the Initial License Examination scheduled for the weeks of 1217/09 and 12114/09, at Peach Bottom Atomic Power Station. This submittal includes the Senior Reactor Operator and Reactor Operator Written Examinations, Job Performance Measures, and Integrated Plant Operation Scenario Guides. These examination materials have been developed in accordance with NUREG-102 t, "Operator Licensing Examination Standards," Revision 9, Supplement I. Please note that reference materials are attached to each individual examination question or item. Some minor modifications have been made to the Integrated Examination Outline with regards to the Job Performance Measures and operational scenarios in order to improve balance and content. These changes improve examination quality and are in compliance with NUREG-I 021, Revision 9, "Operator Licensing Examination Standards," Supplement L Some modifications or adjustments to the examination material may be required due to procedural changes. In accordance with NUREG-1021, Revision 9, Supplement 1, Section ES-201, please ensure that these materials are withheld from public disclosure until after the examinations are complete.

Should you have any questions concerning this letter, please contact David Foss at 717-456-4311.

For questions concerning examination materials, please contact Fred Bruns at 717-456-3793. Site Vice Peach Bottom Atomic Power CCN: 09-74 Exelon Nuclear Peach Bottom Atomic Power Station 1848 lay Road Delta. PA 173'4-9032 October 19, 2009 Mr. Samuel J. Collins Regional Administrator wwwexeloncorp.com U. S. Nuclear Regulatory Commission Region I 475 Allendale Rd. King of Prussia, PA 19406-1415 Peach Bottom Atomic Power Station (pBAPS), Units 2 and 3 Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 Nuclear 10 CPR 55.40

Subject:

Submittal of Integrated Initial License Training Examination Materials Enclosed are the examination materials, which Peach Bottom Atomic Power Station is submitting in support of the Initial License Examination scheduled for the weeks of 1217/09 and 12114/09, at Peach Bottom Atomic Power Station. This submittal includes the Senior Reactor Operator and Reactor Operator Written Examinations, Job Performance Measures, and Integrated Plant Operation Scenario Guides. These examination materials have been developed in accordance with NUREG-102 t, "Operator Licensing Examination Standards," Revision 9, Supplement I. Please note that reference materials are attached to each individual examination question or item. Some minor modifications have been made to the Integrated Examination Outline with regards to the Job Performance Measures and operational scenarios in order to improve balance and content. These changes improve examination quality and are in compliance with NUREG-I 021, Revision 9, "Operator Licensing Examination Standards," Supplement L Some modifications or adjustments to the examination material may be required due to procedural changes. In accordance with NUREG-1021, Revision 9, Supplement 1, Section ES-201, please ensure that these materials are withheld from public disclosure until after the examinations are complete.

Should you have any questions concerning this letter, please contact David Foss at 717-456-4311.

For questions concerning examination materials, please contact Fred Bruns at 717-456-3793. Site Vice President Peach Bottom Atomic Power Station CCN: 09-74

Enclosures:

(Hand carry to Todd Fish. Chief Examiner.

NRC Region 1) RO/SRO Composite Examination with references attached Contml Room Systems and Facility Walk-Through Job Performance Measures with references attached Administrative Topic Job Performance Mea.sures with references attached Integrated Plant Operation Scenario Guides Completed Checklists: Operating Test Quality Checklist (Form ES-301-3) Simulator Scenario Quality Checklist (Form ES-30 1-4) Transient and Event Checklist (Fonn ES-301-5) Competenc ies Checklist (Form ES-.30 1-6) Written Exam Quality Checklist (Form ES-401-6) Examination Security Agreements (Form ES-201-3) Record of Rejected KlAs (Form ES-401-4) F. L. Bower, Senior Resident Inspector.

USNRC, PBAPS (without attachments)

Chief. NRC Operator Licensing Branch (without attachments)

Note: Enclosures to be withheld from public disclosure in accordance with NUREG-I021, Rev. 9, ES*201 until examinations are completed.

Enclosures:

(Hand carry to Todd Fish. Chief Examiner.

NRC Region 1)

• RO/SRO Composite Examination with references attached

• Contml Room Systems and Facility Walk-Through Job Performance Measures with references attached

• Administrative Topic Job Performance Mea.sures with references attached

• Integrated Plant Operation Scenario Guides

• Completed Checklists:

o Operating Test Quality Checklist (Form ES-301-3) o Simulator Scenario Quality Checklist (Form ES-30 1-4) o Transient and Event Checklist (Fonn ES-301-5) o Competenc ies Checklist (Form ES-.30 1-6) o Written Exam Quality Checklist (Form ES-401-6)

• Examination Security Agreements (Form ES-201-3)

• Record of Rejected KlAs (Form ES-401-4) cc: F. L. Bower, Senior Resident Inspector.

USNRC, PBAPS (without attachments)

Chief. NRC Operator Licensing Branch (without attachments)

Note: Enclosures to be withheld from public disclosure in accordance with NUREG-I021, Rev. 9, ES*201 until examinations are completed.

ES-401 BWR Examination Outline FORM ES-401-1 Facility Name: Peach Bottom Date of Exam: 12/07/2009 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G 1 2 3 4 6 1 2 3 4 Total A2 1. 3 3 3 4 4 3 20 4 Emergency

& Abnormal 2 2 N/A N/A 7 2 3 Plant Evolutions Totals 4 5 4 5 5 4 27 6 4 10 2 2 3 26 2 5 2. Plant 2 12 0 3 Systems Tier Totals 3 3 3 8 3. Generic Knowledge and Abilities 4 7 Note: 1. 2. 3. 4. 5. 6. 7.* 8. 9. Categories 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). The point total for each group and tier in the proposed outline must match that specified in the table. 2 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. 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.

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.

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.

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

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. 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. For Tier 3, select topiCS from Section 2 of the KIA catalon, 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 ES-401 BWR Examination Outline FORM ES-401*1 Facility Name: Peach Bottom Date of Exam: 12/07/2009 RO KIA Category Points Tier Group K K K K K K A A A A G 1 2 3 4 5 6 1 2 3 4

• Total A2 1. 1 3 3 3 4 4 3 20 4 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 2Eg2 3 26 3 2 5 2. Plant 2 1 1 111 1 12 0 2 1 3 Systems Totals 3 3 3 4 4 3 4 38 5 3 8 3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 10 7 Categories 2 3 I 3 2 2 1 I 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 oullines (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 catalon, 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 ES401 2 Form ES401-1 !ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant E\lolutions Tier 1/Group 1 (RO) I,------------------------,r-,r-.--.

E/APE # I Name I Safety Function 295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 &4 295003 Partial or Complete Loss of AC 16 295004 Partial or Total Loss of DC Pwr 16 295005 Main Turbine Generator Trip 13 295006 SCRAM I 1 295016 Control Room Abandonment 17 295018 Partial or Total Loss ofCCW 18 295019 Partial or Total Loss of Inst. Air 18 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 !Temp./5 1295027 High Containment Temperature 15 295028 High Drywell Temperature I 5 295030 Low Suppression Pool Wtr L\l11 5 295031 Reactor Low Water Le\lell 2 295037 SCRAM Condition Present and Reactor Power Abo\le APRM Downscale or Unknown 11 295038 High Off-site Release Rate 19 600000 Plant Fire On Site I 8 and Electric Grid KKK A G 123 2 o 5 o 6 KIA Topic(s) lR Core flow indication 2.9 I Fa,ilsafe coml:JO'nent design Knowledge of limiting conditions for operations and safety limits. u****I';:li,.I!.'*1 o 3 o 2 3 o 1 o 3 3 o 2 o 2 3 Pressure effects on reactor level Knowledge of low power/shutdown Imphcations In aCCident (e.g., loss of coolant accident or lOSS of residual heat removal) 3 mitigation strategies.

4.2 Reactor

power reduction 3.3 4.2 ICampon,enl cooling water systems' Plant-Specific Area radiation levels Drywall pressure Suppression pool cooling Components internal to the drywell 3.2 ECCS systems (NPSH conSiderations)

Plant-Specific I 3.6 Reactor pressure 4.2 Reactor water level effects on reactor power 4.1 Plant ventilation systems Fife fighting equipment used on each class of fire Engineered safety features Group Point Total: # 20 ES401 2 Form ES401-1 !ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant E\lolutions Tier 1/Group 1 (RO) I,------------------------,r-,r-.--.

E/APE # I Name I Safety Function 295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 & 4 295003 Partial or Complete Loss of AC 16 295004 Partial or Total Loss of DC Pwr 16 295005 Main Turbine Generator Trip 13 295006 SCRAM I 1 295016 Control Room Abandonment 17 295018 Partial or Total Loss ofCCW 18 295019 Partial or Total Loss of Inst. Air 18 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 !Temp./5 1295027 High Containment Temperature 15 295028 High Drywell Temperature I 5 295030 Low Suppression Pool Wtr L\l11 5 295031 Reactor Low Water Le\lell 2 295037 SCRAM Condition Present and Reactor Power Abo\le APRM Downscale or Unknown 11 295038 High Off-site Release Rate 19 600000 Plant Fire On Site I 8 and Electric Grid KKK A G 123 2 o 5 o 6 KIA Topic(s) lR Core flow indication 2.9 I Fa,ilsafe coml:JO'nent design Knowledge of limiting conditions for operations and safety limits. u****I';:li,.I!.'*1 o 3 o 2 3 o 1 o 3 3 o 2 o 2 3 Pressure effects on reactor level Knowledge of low power/shutdown Imphcations In aCCident (e.g., loss of coolant accident or lOSS of residual heat removal) 3 mitigation strategies.

4.2 Reactor

power reduction 3.3 4.2 ICampon,enl cooling water systems' Plant-Specific Area radiation levels Drywall pressure Suppression pool cooling Components internal to the drywell 3.2 ECCS systems (NPSH conSiderations)

Plant-Specific I 3.6 Reactor pressure 4.2 Reactor water level effects on reactor power 4.1 Plant ventilation systems Fife fighting equipment used on each class of fire Engineered safety features Group Point Total: # 20 ES-401 3 Form ES-401*1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier llGroup 2 (RO) EIAPE # I Name I Safety Function KJlWA A G KIA Topic(s) IR #1 1 2 295002 Loss of Main Condenser Vac I 3 0 295007 High Reactor Pressure I 3 0 2 ....* i-* Reactor power 3.8 1 295008 High Reactor Water Levell 2 I ... r '.. 0 295009 Low Reactor Water Levell 2 ti-tfReactor water level indication 3.9 1 r;" 295010 High Drywell Pressure 15 Drywell ventilation/cooling 3.4 1 1295011 High Containment Temp I 5 . 0 295012 High Drywel1 Temperature I 5 0 .... 3.3 1 1 Pressure/temperature relationship 295013 High Suppression Pool Temp. 15 0 295014 Inadvertent Reactivity Addition 11 Control rod blocks 3.7 1 295015 Incomplete SCRAM I 1 F';: 0 295017 High Off-site Release Rate I 9 04. Knowledge of EOP entry conditions and immediate action steps. 4.6 1 01 1295020 Inadvertent Cont. Isolation I 5 & 7 0 295022 Loss of CRD Pumps 11 I 0 . < i ***** ('. 295029 High Suppression Pool Wtr Lvii 5 *** 0 295032 High Secondary Containment Area 0 Temperature 15 I ******i*:.

i""C' 295033 High Secondary Containment Area 0 Radiation Levels 19 295034 Secondary Containment Ventilation 1>:::< 0 High Radiation I 9 295035 Secondary Containment High E Differential Pressure I 5 295036 Secondary Containment High 0 Cause of the high water level 3.4 1 SumplArea Water Levell 5 3 500000 High CTMT Hydrogen Conc. I 5 ++ . 0". KIA Category Totals: Group Point Total: 1 7 ES-401, Page 19 of ES-401 3 Form ES-401*1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier llGroup 2 (RO) EIAPE # I Name I Safety Function KJlWA A G KIA Topic(s) IR # 1 1 2 295002 Loss of Main Condenser Vac I 3 0 295007 High Reactor Pressure I 3 0 2 ....* i-* Reactor power 3.8 1 295008 High Reactor Water Levell 2 I ... r ' .. 0 295009 Low Reactor Water Levell 2 ti-tf Reactor water level indication 3.9 1 r;" 295010 High Drywell Pressure 15 Drywell ventilation/cooling 3.4 1 1295011 High Containment Temp I 5 . 0 295012 High Drywel1 Temperature I 5 0 .... 3.3 1 1 Pressure/temperature relationship 295013 High Suppression Pool Temp. 15 0 295014 Inadvertent Reactivity Addition 11 Control rod blocks 3.7 1 295015 Incomplete SCRAM I 1 F';: 0 295017 High Off-site Release Rate I 9 04. Knowledge of EOP entry conditions and immediate action steps. 4.6 1 01 1295020 Inadvertent Cont. Isolation I 5 & 7 0 295022 Loss of CRD Pumps 11 I 0 . < i ***** ('. 295029 High Suppression Pool Wtr Lvii 5 *** 0 295032 High Secondary Containment Area 0 Temperature 15 I ******i*:.

i""C' 295033 High Secondary Containment Area 0 Radiation Levels 19 295034 Secondary Containment Ventilation 1>:::< 0 High Radiation I 9 295035 Secondary Containment High E Differential Pressure I 5 295036 Secondary Containment High 0 Cause of the high water level 3.4 1 SumplArea Water Levell 5 3 500000 High CTMT Hydrogen Conc. I 5 ++ . 0 ". KIA Category Totals: Group Point Total: 1 7 ES-401, Page 19 of 34 ES-401 4 Form ES-401-1 !ES-401 BWR Examination Outline Fonm ES-401-1 Plant Systems -Tier 2/Group 1 (RO) System # / Name K K K K K t<iA Top;",) IR #1 2 3 4 5 203000 RHRlLPCI:

Injection Mode 0 0 logiC, Core cooling methods 2.7; 2 3.5 205000 Shutdown Cooling I ,

Heat removal mechanisms 2.8 1 206000 HPCI ottt ,; D.C. power-BWR-2, 3, 4 3,7 1 7 .*'.'..,***.,..**.. 207000 Isolation (Emergency) 0 Condenser

209001 LPCS 0 '""-..m '""","' 3,8: 1 and operate controls identified in the alarm response 4.2 2 manual. '209002 HPCS .......*..

"..........

0 211000 SLC 0 rn Explosive valve operation 3.1 1 4 :212000 RPS 0 1"---,,, 1 1 .* ,;; Ifi, RPS motor-generalor output voltage 2,8 1 o i' 2150031RM I 2 i.. Reactor power indication response to rod posItion changes 37 1 215004 Source Range Monitor ........, 0 215005 APRM ILPRM 0 I Flow biased trip setpoints 3,7 1 7 i217000 RCIC 0 0 Suppression pool level, System valves 3,3; 2 7 3 3,4 1218000 ADS 0 ;2' ()2. A C power. Plant-Specific; Knowledge of surveillance 3; 3,7 2 5 fii2 procedures 223002 PCIS/Nuclear Steam Supply 0 I',;' 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" 36 , ."" All individual component controllers when transferrfng from 259002 Reactor Water Level Control t 3 manual to automatic modes 3.8 1 1261000 SGTS () /52 Valve closures 2,9 1 *l 11262001 AC Electrical Distribution X". Exceeding voltage limitations 3,1 1 i 0 ,i 262002 UPS (ACIDC) Static Inverter 2.7 1 3 .... 1263000 DC Electrical Distribution 0 I'c, i;; 3.1 1 1 Major D.C loads " 264000 EDGs 0 1+ Emergency generator trips {normal} 3.5 1 1 !300000 Instrument Air 0 Cooling water to compressor; Ability to interpret and 2.8; 2 4 execute procedure steps 4.6 400000 Component Cooling Water ff 0 on instrument Signal levelS for normal operations" 3 1 1 gs, and lrips that are applicable to the CCWS 0 KIA Category Totals: 212 2 3 3 2 Point Total: 26 ES-401, Page 20 of ES-401 4 Form ES-401-1 !ES-401 BWR Examination Outline Fonm ES-401-1 Plant Systems -Tier 2/Group 1 (RO) System # / Name K K K K K KIA Topic(s) IR # 1 2 3 4 5 203000 RHRlLPCI:

Injection Mode 0 0 I!}Z:'" II00tiation logic, Core cooling methods 2.7; 2 'TIt-f.1 3.5 205000 Shutdown Cooling I , ...*... Heat removal mechanisms 2.8 1 ottt 206000 HPCI i.; D.C. power-BWR-2, 3, 4 3,7 1 7 .*.*. , .. , ....**. 207000 Isolation (Emergency) 0 Condenser

209001 LPCS 0 " .. " '."m ,-.. 3,8: 1 and operate controls identified in the alarm response 4,2 2 manual. 1209002 HPCS I***.*." .*. * ** *** ,it 211000 SLC 0 rn r*' Explosive valve operation 4 :212000 RPS 0 Irj .** RPS motor-generator output voltage 2,8 1 I ** *** *** 1 I .* ,;; o I, 2150031RM I 2 I.; Reactor power indication response to rod posItion changes 37 1 215004 Source Range Monitor ........ , 0 215005 APRM ILPRM 0 I Flow biased trip setpoints 3,7 1 7 i217000 RCIC 0 0 Suppression pool level, System valves 3,3; 2 7 3 3,4 1218000 ADS 0 ;2' ()2. A C power. Plant-Specific; Knowledge of surveillance 3; 3.7 2 5 .12 procedures 223002 PCIS/Nuclear Steam Supply 0 3,5 1 res Shutoff 2 239002 SRVs 0 0 Reactor pressure control: Allows for SRV operation from 3.9; 2 1 5 more than one location" 36 , ."" 0, ****. All individual component controllers when transferrfng from 259002 Reactor Water Level Control t manual to automatic modes 3.8 1 3 1261000 SGTS

.. 2,9 1 11262001 AC Electrical Distribution

! I , .... ,_ ...

3,1 1 i 0 262002 UPS (ACIDC) t':-,-:

StatiC Inverter 2,7 1 3 .... 1263000 DC Electrical Distribution 0 I'c, f2 3.1 1 1 E Major D.C loads (: " -mEmergency generator trips (normal) 264000 EDGs 0 1'< 3,5 1 1 ft !300000 Instrument Air 0 Cooling water to compressor; Ability to interpret and 2.8; 2 4 execute procedure steps 4.6 400000 Component Cooling Water ff O! jnts on instrument Signal levelS for normal operations" 3 1 1 rnings. and lrips that are applicable to Ihe CCWS 0 KIA Category Totals: 212 2 3 3 2 2 Group Point Total: 26 ES-401, Page 20 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 KIA Topic(s) IR #1 2 3 4 5 6 1 2 3 201001 CRD Hydraulic

• • • • • 0 RMCS 003 Control Rod and Drive Mechanism 201004 RSCS 0 101005 RCIS 0 01006 RWM 0 Rod posilion P-Spec(Not-6WR6) 3.2 1 1 2001 Recirculation 0 Reactor water level 3.7 1 9 Abilily 10 determine operability andlor ava,lability of safety 3.6 1 I 37 retated equipment t__ 0 1214000 RPIS .........

0 .... 215001 Traversing In-core Probe 0 Primary containment isolation Mark-I!'.II( Not-8WR1) 3.4 1 1 215002 RBM 0 216000 Nuclear Boiler Inst ........ 0 219000 RHR/LPCI:

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

CTMT Spray Mode < 0 230000 RHR/LPCI:

Torus/Pool Spray Mode 1 Valve logiC tailure 3.2 1 2 ley 233000 Fuel Pool Cooling/Cleanup 0 R pumps tEE 2 234000 Fuel Handling Equipment IT;: ....: ****:*1.*:***:

I.* ... I; . : 239001 Main and Reheat Steam 0 239003 MSIV Leakage Control 0 ' . 241000 Reactor/Turbine Pressure Regulator

....... 0 Main Turbine Gen. / Aux. 0 Turbine operation and 2.8 1 2 256000 Reactor Condensate I 0 259001 Reactor Feedwater 0 12. Recirculation pump NPSH 2.9 1 5 268000 Radwaste 0 Drywall Hoar drains 2.9 1 6 271000 Offgas l:-i. Vaws closures 2.6 1 272000 Radialion Monitoring f-: m < 0 286000 Fire Protection 0 288000 Plant Ventilation 0 Plant air systems 3 2.7 1 290001 Secondary CTMT ie/ 0 290003 Control Room HVAC 0 290002 Reactor Vessel Internals f,f 0 bd KIA Category Totals: 1 1 1 1 1 1 1 2_ 1 1 .1 Group Point Total: ES-401, Page 22 of 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 KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 201001 CRD Hydraulic

• • • • • 0 RMCS 003 Control Rod and Drive Mechanism 201004 RSCS 0 101005 RCIS 0 01006 RWM 0 Rod posilion P-Spec(Not-6WR6) 3.2 1 1 2001 Recirculation 0 Reactor water level 3.7 1 9 Abilily 10 determine operability andlor ava,lability of safety 3.6 1 I 37 retated equipment t __ 0 1214000 RPIS .........

0 .... 215001 Traversing In-core Probe 0 Primary containment isolation Mark-I!'.II( Not-8WR1) 3.4 1 1 215002 RBM 0 216000 Nuclear Boiler Inst ........ 0 219000 RHR/LPCI:

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

CTMT Spray Mode < 0 230000 RHR/LPCI:

Torus/Pool Spray Mode 1 Valve logiC tailure 3.2 1 2 ley 233000 Fuel Pool Cooling/Cleanup 0 R pumps tEE 2 234000 Fuel Handling Equipment IT;: .... : ****:*1.*:***:

I.* ... I; I-. : 239001 Main and Reheat Steam 0 239003 MSIV Leakage Control 0 ' . 241000 Reactor/Turbine Pressure Regulator

....... 0 Main Turbine Gen. / Aux. 0 Turbine operation and 2.8 1 2 256000 Reactor Condensate I 0 259001 Reactor Feedwater 0 12. Recirculation pump NPSH 2.9 1 5 268000 Radwaste 0 [-Drywall Hoar drains 2.9 1 6 271000 Offgas l:-i. Vaws closures 2.6 1 272000 Radialion Monitoring f-: m< 0 286000 Fire Protection 0 288000 Plant Ventilation 0 Plant air systems 3 2.7 1 290001 Secondary CTMT ie/ 0 290003 Control Room HVAC 0 290002 Reactor Vessel Internals f,f 0 bd KIA Category Totals: 1 1 1 1 1 1 1 2_ 1 1 .1 Group Point Total: 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 # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G KIA Topic(s) IR # 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 0 295003 Partial or Complete Loss of AC / 6 02. 3"7 Ability to determine operability and/or availability of safety related equipment 4.6 1 295004 Partial or Total Loss of DC Pwr / 6 02. 22 Knowledge of limiting conditions for operations and safety limits. 4.7 1 295005 Main Turbine Generator Trip / 3 0 295006 SCRAM / 1 0 295016 Control Room Abandonment / 7 0 295018 Partial or Total Loss of CCW / 8 0 3 Cause for partial or complete loss 3.5 1 295019 Partial or Total Loss of Ins!. Air /8 0 1 Instrument air system pressure 3.6 1 295021 Loss of Shutdown Cooling / 4 .. 0 295023 Refueling Acc / 8 0 295024 High Drywell Pressure / 5 0 295025 High Reactor Pressure / 3 0 295026 Suppression Pool High Water Temp. 5 0 295027 High Containment Temperature / 5 ," 0 295028 High Drywell Temperature / 5 04. 21 v'

.... _"" v. 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 Lvi / 5 0 295031 Reactor Low Water Level / 2 0 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 0 3 SBLC tank level 4.4 1 295038 High Off-site Release Rate / 9 <4 Source of off-site release 4.5 1 Plant Fire On Site / 8 0 700000 Generator Voltage and Electric Grid Disturbances / 6 0 KIA Category Totals: 0 0 0 0 <4 .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 # / Name / Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 295001 Partial or Complete Loss of Forced 0 Core Flow Circulation / 1 & 4 295003 Partial or Complete Loss of AC / 6 02. Ability to determine operability and/or availability of 4.6 1 3"7 safety related equipment 295004 Partial or Total Loss of DC Pwr / 6 02. Knowledge of limiting conditions for operations and 4.7 1 22 safety limits. 295005 Main Turbine Generator Trip / 3 0 295006 SCRAM / 1 0 295016 Control Room Abandonment / 7 0 295018 Partial or Total Loss of CCW / 8 0 Cause for partial or complete loss 3.5 1 3 295019 Partial or Total Loss of Ins!. Air /8 0 Instrument air system pressure 3.6 1 1 .. 295021 Loss of Shutdown Cooling / 4 0 295023 Refueling Acc / 8 0 295024 High Drywell Pressure / 5 0 295025 High Reactor Pressure / 3 0 295026 Suppression Pool High Water Temp. 0 5 295027 High Containment Temperature / 5 ," 0 v'

... . _"" v. 04. 295028 High Drywell Temperature / 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 Lvi / 5 0 295031 Reactor Low Water Level / 2 0 295037 SCRAM Condition Present 0 and Reactor Power Above APRM SBLC tank level 4.4 1 Downscale or Unknown / 1 3 295038 High Off-site Release Rate / 9 0-Source of off-site release 4.5 1 <4 Plant Fire On Site / 8 0 700000 Generator Voltage and Electric Grid Disturbances / 6 0 KIA Category Totals: 0 0 0 0 <4 .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) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G KIA Topic(s) IR # 295002 Loss of Main Condenser Vac / 3 0 295007 High Reactor Pressure / 3 0 295008 High Reactor Water Level / 2 0 295009 Low Reactor Water Level / 2 0 295010 High Drywell Pressure / 5 0 295011 High Containment Temp / 5 , 0 295012 High Drywell Temperature / 5 0 295013 High Suppression Pool Temp, / 5 0 2 Localized heating/stratification 3,5 1 295014 Inadvertent Reactivity Addition / 1 0 295015 Incomplete SCRAM / 1 0 295017 High Off-site Release Rate / 9 0 295020 Inadvertent Cont Isolation / 5 & 7 0 3 Reactor power 3,7 1 295022 Loss of CRD Pumps / 1 0 295029 High Suppression Pool Wtr Lvi / 5 0 295032 High Secondary Containment Area Temperature / 5 0 295033 High Secondary Containment Area Radiation Levels / 9 0 295034 Secondary Containment Ventilation High Radiation / 9 04. 06 Knowledge of EOP mitigation strategies 4,7 1 295035 Secondary Containment High Differential Pressure / 5 0 295036 Secondary Containment High Sump/Area Water Levell 5 0 500000 High CTMT Hydrogen Conc, /5 0 KIA Category Totals: 0 0 0 0 2 1 Group Point Total: 3 ES-40 1, 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) E/APE # / Name / Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 0 295007 High Reactor Pressure / 3 0 295008 High Reactor Water Level / 2 0 295009 Low Reactor Water Level / 2 0 295010 High Drywell Pressure / 5 0 , 295011 High Containment Temp / 5 0 295012 High Drywell Temperature / 5 0 295013 High Suppression Pool Temp, / 5 0 Localized heating/stratification 3,5 1 2 295014 Inadvertent Reactivity Addition / 1 0 295015 Incomplete SCRAM / 1 0 295017 High Off-site Release Rate / 9 0 295020 Inadvertent Cont Isolation / 5 & 7 0 Reactor power 3,7 1 3 295022 Loss of CRD Pumps / 1 0 295029 High Suppression Pool Wtr Lvi / 5 0 295032 High Secondary Containment Area 0 Temperature / 5 295033 High Secondary Containment Area 0 Radiation Levels / 9 295034 Secondary Containment Ventilation

04. Knowledge of EOP mitigation strategies 4,7 1 High Radiation / 9 06 295035 Secondary Containment High 0 Differential Pressure / 5 295036 Secondary Containment High 0 Sump/Area Water Levell 5 500000 High CTMT Hydrogen Conc, /5 0 KIA Category Totals: 0 0 0 0 2 1 Group Point Total: 3 ES-40 1, 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 # I Name K K K A'.'A** A A G KIA Topic(s) IR #1 2 3 612.3 4 203000 RHR/LPCI:

Injection i:7.i; 0 205000 Shutdown Cooling Mode I.**.*. 0 206000 HPCI 1.. **1 0 -h,,;. 207000 Isolation (Emergency) 0 Condenser

.... 209001 LPCS ..* ; 0 :i%§ 209002 HPCS 0 211000 SLC 0 ....... 4 Inadequate system now 3.4 1 212000 RPS 0 2150031RM "02. . 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 223002 PCIS/Nuclear Steam Supply 1 .... Shutoff 1 Standby liquid initiation 3.9 1 i239002 SRVs 1,:'\' 0 i.'; 259002 Reactor Water Level Control I": 0 r" 261000 SGTS I:;

0 1,'*.;.::.. 262001 AC Electrical Distribution 1< L.,; 3.4 1 I,. I 262002 UPS (AC/DC) I:* I::*; 0 I" Knowledge of low power/shutdown implicaMns In aCCIdent 263000 DC Electrical Distribution (e g .. loss of coolant accident or loss of reSidual heal 4.2 1 /.:: removal) mitigation strategies 264000 EDGs 0 ::s:: 300000 Instrument Air . 0 Component Cooling Water 0 *c. 0 KIA Category Totals: 0 0 0 0 0 0 0 I: 5 ES-401, Page 20 of ES*401 4 Form ES-401*1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems Tier 2/Group 1 (SRO) System # I Name K K K A'.'A ** A A G KIA Topic(s) IR # 1 2 3 612.3 4 203000 RHR/LPCI:

Injection i:7.i; 0 205000 Shutdown Cooling Mode I.**.*. 0 206000 HPCI 1 .. **1 0 -h,,;. 207000 Isolation (Emergency) 0 Condenser

.. .. 209001 LPCS ..* ; 0 :i%§ 209002 HPCS 0 211000 SLC 0 ....... 4 Inadequate system now 3.4 1 212000 RPS 0 2150031RM "02. . 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 223002 PCIS/Nuclear Steam Supply 1 .... Shutoff 1 Standby liquid initiation 3.9 1 i239002 SRVs 1,:'\' 0 i.'; 259002 Reactor Water Level Control I": 0 r" 261000 SGTS I:;

0 1,' *. ;.:: .. 262001 AC Electrical Distribution 1< L.,; 3.4 1 I,. I 262002 UPS (AC/DC) I:* I::*; 0 I" Knowledge of low power/shutdown implicaMns In aCCIdent 263000 DC Electrical Distribution (e g .. loss of coolant accident or loss of reSidual heal 4.2 1 /.:: removal) mitigation strategies 264000 EDGs 0 ::s:: 300000 Instrument Air . 0 Component Cooling Water 0 *c. 0 KIA Category Totals: 0 0 0 0 0 0 0 I: 5 ES-401, Page 20 of 34 ES-401 5 Form ES-401*1 ES-401 System # I Name 001 CRD Hydraulic Control Rod and Drive Mechanism

,201004 RSCS RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 1,215001 Traversing In-core Probe 11215002 RBM 1216000 Nuclear Boiler Inst ,219000 RHR/LPCI:

Torus/Pool Cooling iMode I Primary CTMT and Aux. P26001 RHR/LPCI:

CTMT Spray Mode F30000 RHR/LPCI:

Torus/Pool Spray Mode Fuel Pool Cooling/Cleanup BWR Examination Outline Plant Systems Tier 2/Group 2 (SRO) KKK KKK A A A A 1 2 3 4 5 6 1

I 3 4 (\¢1 KIA Topic(s) .*.... i.. I -: Ability to apply Technical Specifications for a system. Form ES-401-111 IR o o o o o o o o o o o o o o o 4.7 o Fuel Handling Equipment I>i Main and Reheat Steam MSIV Leakage Control 241000 ReactorlTurbine Pressure Regulator

.... o 245000 Main Turbine Gen. / Aux. o 256000 Reactor Condensate o

268000 Radwaste " 271000 Offgas o 272000 Radiation Monitoring o 286000 Fire Protection o .'. a Plant Ventilation 290001 Secondary CTMT High area temperature

3.3 290003

Control Room HVAC : . o 290002 Reactor Vessel Internals Exceeding safety limits 4.5 .. KIA Category Totals: t\ t\ " o 11 Group PointTotal:

ES-401. Page 21 of ES-401 ES-401 System # I Name 001 CRD Hydraulic Control Rod and Drive Mechanism

,201004 RSCS RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 1,215001 Traversing In-core Probe 11215002 RBM 1216000 Nuclear Boiler Inst ,219000 RHR/LPCI:

Torus/Pool Cooling iMode I Primary CTMT and Aux. P26001 RHR/LPCI:

CTMT Spray Mode F30000 RHR/LPCI:

Torus/Pool Spray Mode Fuel Pool Cooling/Cleanup 5 Form ES-401*1 BWR Examination Outline Plant Systems Tier 2/Group 2 (SRO) KKK KKK A A A A 1 2 3 4 5 6 1

I 3 4 (\¢1 KIA Topic(s) .* .... i .. I -: Ability to apply Technical Specifications for a system. Form ES-401-111 IR o o o o o o o o o o o o o o o 4.7 o Fuel Handling Equipment I>i Jli-+-+-::+------..-

Main and Reheat Steam MSIV Leakage Control 241000 ReactorlTurbine Pressure Regulator

.... o 245000 Main Turbine Gen. / Aux. o 256000 Reactor Condensate o

268000 Radwaste " 271000 Offgas o 272000 Radiation Monitoring o 286000 Fire Protection o .'. a Plant Ventilation 290001 Secondary CTMT ,-----High area temperature

3.3 290003

Control Room HVAC : . o 290002 Reactor Vessel Internals Exceeding safety limits 4.5 .. KIA Category Totals: t\ t\ " o 11 Group PointTotal:

ES-401. Page 21 of 34 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 1. Conduct of Operations to interpret reference matenals, such as graphs. curves. tables. etc Ability to I,ocate and operate components, including loeal controls Abillity to Ima"ag,e ttle (:onltrolroc,m

<ore,. d,Jrin.g plant transients.

Abillity to ,explain and apply system limits and precautions pre-startup procedures for Ihe facility.

including operating those controls equipment that could affect reactivity.

2.2. 07 Knowledge of the process for conducting special or ,nfrequent tests. 2.2. 14 of the process for controlling eqUipment configuration or status I Knlowled!,e of the process used to track inoperable alarms. 3.9 1 4.4 1 4.8 1 4 4.5 1 2.9 1 3.9 1 3.3 I Knowled,le of radiation exposure limits under normal or emergency conditions use radiation mOnitoring systems, such as fb<ed radiation monitors and alarms. 1 2.9 instruments, personnel monitoring eqUipment etc 3.5 1 Ie c,f E*OP terms and definitions.

Ie of R() responsibilities in emergency plan implementation.

3.9 1 4.2 without reference to procedures those actions that require Immediate

4.4 components

and controls ES-401, Page 27 of 34 ES-401 1. Conduct of Operations Generic Knowledge and Abilities Outline (Tier 3) to interpret reference matenals, such as graphs. curves. tables. etc Ability to I,ocate and operate components, including loeal controls Abillity to Ima"ag,e ttle (:onltrolroc,m

<ore,. d,Jrin.g plant transients.

Abillity to ,explain and apply system limits and precautions pre-startup procedures for Ihe facility.

including operating those controls equipment that could affect reactivity.

2.2. 07 Knowledge of the process for conducting special or ,nfrequent tests. 2.2. 14 of the process for controlling eqUipment configuration or status I Knlowled!,e of the process used to track inoperable alarms. I Knowled,le of radiation exposure limits under normal or emergency conditions use radiation mOnitoring systems, such as fb<ed radiation monitors and alarms. instruments, personnel monitoring eqUipment etc Ie c,f E*OP terms and definitions.

Ie of R() responsibilities in emergency plan implementation.

without reference to procedures those actions that require Immediate components and controls ES-401, Page 27 of 34 Form ES-401-3 3.9 1 4.4 1 4.8 1 4 4.5 1 2.9 1 3.9 1 3.3 2.9 1 3.5 1 3.9 1 4.2 4.4 ES-401 Record of Rejected KlAs Form ES-401-4 Tier / Group Randomly Selected KIA Reason for Rejection RO 1 /1 0#53 295023 AA2.05 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 1 /1 0#58 295030 EA1.03 Peach Bottom does not have HPCS. (Replaced with KIA 2950:30 EA1.01) RO 2 /1 0#21 218000 G2.2.38 There are no conditions and limitations in the facility license associated with ADS. (Replaced witlh KIA 218000 G2.2.12) RO 2/1 0#25 262002 K4.01 Unable to construct another inverter question for this KIA ... too similar to KIA 262002 K6.03 for Ouestion #12. (Replaced with KIA 239002 K4.05) R02/2 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 G2.214) RO 3/2 0#75 G2.2.18 Managing maintenance (risk assessments, work prioritization, etc.) is not required knowledge for R.Os. (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 R.Os. (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 1 /1 295023 Emergency Plan entry conditions are not required knowledge for ROs. 0#53 AA2.05 (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 1 /1 295030 Peach Bottom does not have HPCS. 0#58 EA1.03 (Replaced with KIA 2950:30 EA1.01) RO 2 /1 218000 There are no conditions and limitations in the facility license associated 0#21 G2.2.38 with ADS. (Replaced witlh KIA 218000 G2.2.12) RO 2/1 262002 Unable to construct another inverter question for this KIA ... too similar to 0#25 K4.01 KIA 262002 K6.03 for Ouestion #12. (Replaced with KIA 239002 K4.05) R02/2 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 G2.214) RO 3/2 G2.2.18 Managing maintenance (risk assessments, work prioritization, etc.) is not 0#75 required knowledge for R.Os. (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 R.Os. (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 Outline Form ES-301-1 Facility:

Peach Bottom Date of Examination:

12/07/2009 Examination Level: RO k8J SRO D Operating Test Number: NRC Administrative Topic (See Note) Conduct of Operations Conduct of Operations Equipment Control N. RIS D. RIS D. P, RIS Describe activity to be performed G2.1.32 -Complete Attachment 1 of AO 3.8 "Evaluation of High CRD Temperature on Control Rod Scram Time" (PLOR-266C)

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

(2008 NRC Exam) I Radiation Control N/A N/A G2.4.39 -Identify Errors on State and Local Notification i Emergency Plan N. RIS Form -Return Form to SED for Correction (PLOR-341CA)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retakin 9 onl Y the administrative to p iCs. when 5 are re q uired. .. Type Codes & (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from banl< (.::: 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 I i Administrative Outline Form ES-301-1 Facility:

Peach Bottom Date of Examination:

12/07/2009 Examination Level: RO k8J SRO D Operating Test Number: NRC I 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-341CA)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retakin onl the administrative to iCs. when 5 are re uired. 9 Y p q .. Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from banl< (.::: 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 Administrative ODICS Outline Form ES-301-1 SRO Operating Test Number: !...!N.!...!R=C

___ Administrative Topic Type Describe activity to be performed (See Note) Code* G2.1.32 Review Attachment 1 of AO 3.8 "Evaluation of High CRD Temperature on Control Rod Scram Time" and Conduct of Operations N, RIS Identify and Declare SLOW Control Rod * (PLOR-340CA

-SRO)

• G2.1.7 -Compliance with Asymmetric Feedwater Heating Conduct of Operations D, RIS Operation (AFTO) (PLOR-252C)

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

(2007 NRC Exam) Review and Authorize Issuance of Thyroid Radiation Control M, RIS I

• Blocking Agent (KI) (PLOR-215C) -EAL Classification with State and Local Emergency Plan D, RIS Notifiations

-Alert due to RPS Failure (PLOR-233C) 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 & (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) ES 301, Page 22 of 27 ES-301 Administrative Topic (See Note) Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan Administrative Outline Form ES-301-1 SRO [gJ Type Code* N, RIS D, RIS Operating Test Number: !...!N.!...!R-",C

__ _ Describe activity to be performed 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)

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

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

(2007 NRC Exam) M, RIS D, RIS I G2.3.4 Review and Authorize Issuance of Thyroid

• Blocking Agent (KI) (PLOR-215C)

G2.4.41 -EAL Classification with State and Local Notifiations

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

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) ES 301, Page 22 of 27 ES-301 Outline Form ES-301-2 Control Room/In-Plant

....u.:."Alm Facility:

Peach Bottom Date of Examination:

12/07/2009 Exam Level: RO I:8J SRO-I 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 EA1.04 -Standby Liquid Control System 1 Inject SBLC (Alternate A, 0, EN, 1 Path -Low SBLC Discharge Pressure) (PLOR-331 CAl [Se!1] ! L, S i b. 217000 A4.03 -Reactor Core Isolation Cooling System 1 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 0, L,P, S I 3 Isolation (PLOR-083C)

(2008 NRC Exam) [Set 3] ; d. 206000 A2.09 -High Pressure Coolant Injection System 1 Manual A, System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low I CST Level) (PLOR-343C;A)

[Set 1] e. 223001 A4.10 -Primary Containment System and Auxiliaries 1 CAD D,S 5 l System Nitrogen Addition To Containment During Normal Operations I§et 2] i f. 264000 A4.04 -Emergency Generators 1 Diesel Generator Load Test A,D,S ! 6 * (Alternate Path -Load Control Difficulty) (PLOR-322CA)

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

[Set;3J 261000 A4.03 -Standby Gas Treatment (SBGT) System I Manually Start D,S 9 I SBGT System on Equipment Cell Exhaust (PLOR-018C}

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. 0, L, R 4 MSIV (Alternate Path -Removing Fuse Fails To Close MSIV Unit 3) (PLOR-313PA)

j. 295037 EA1.01 -Reactor Protection System / Scram Solenoid De-C,D, 7 energization Unit 2 (T-213-2) (PLOR-075P)

L,R

• k. 2180000 A203 -Backup Instrument Nitrogen to ADS System Startup and 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 overlap those tested in the control room.

• Type Codes I Criteria for RO 1 SRO-I / SRO-U ES-301 Facility:

Peach Bottom Exam Level: RO I:8J SRO-I Control Room/In-Plant

.... \I'C."AIITI SRO-U D Outline Form ES-301-2 Date of Examination:

12/07/2009 Operating Test Number: :..;:.N.:....:R.=C

__ _ Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System I JPM Title Type Code* a. 295037 EA1.04 -Standby Liquid Control System 1 Inject SBLC (Alternate A, 0, EN, Safety Function Path -Low SBLC Discharge Pressure) (PLOR-331 CAl [Se!1-'-'--

___ -,*,----=L::!.., -=.S_-+ ___ ---1 i b. 217000 A4.03 -Reactor Core Isolation Cooling System 1 Manually A, 0, EN,

• Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

L, S c. 239001 A4.01 -Reopen Main Steam Isolation Valves after a Group I 0, L,P, S (PLOR-083C)

(2008 NR:.;;::.C-,E=x=ac:..:..mL)

..L[S=-=e:.;;::.t

-=.31L-______ ---t--------j-----.JI

d. 206000 A2.09 -High Pressure Coolant Injection System 1 Manual System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low CST Level) (PLOR-343C;A)

[Set 1] e. 223001 A4.10 -Primary Containment System and Auxiliaries I CAD l System Nitrogen Addition To Containment During Normal Operations I§et 2] I f. 264000 A4.04 -Emergency Generators 1 Diesel Generator Load Test * (Alternate Path -Load Control Difficulty) (PLOR-322CA)

[Set 4] :1 g. 400000 A4.01 -Component Cooling Water System 1 ECW System i Makeup to Tower Using ESW System (PLOR-270C)

[Set;3J 261000 A4.03 -Standby Gas Treatment (SBGT) System I Manually Start SBGT System on Equipment Cell Exhaust (PLOR-018C}

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 I Close A Stuck Open MSIV (Alternate Path -Removing Fuse Fails To Close MSIV Unit 3) (PLOR-313PA)

j. 295037 EA1.01 -Reactor Protection System / Scram Solenoid De-energization Unit 2 (T-213-2) (PLOR-075P)

• k. 2180000 A203 -Backup Instrument Nitrogen to ADS System Startup and Operation (Unit 3) (PLOR-271 P) A, I D,S A,D,S ! N,S D,S A. 0, L, R C,D, L,R D,E, R @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety 5 6 8 9 4 7 3 functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes I Criteria for RO 1 SRO-I / SRO-U i I II (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L}ow-Power I Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator 4-6 I I > 1 I > 1 I >2 I > <3 I > 1 I > I I < I > I 1 (control room system) I I > I < 2 (randomly selected)

I ES-301, Page 23 of (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank 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) (L}ow-Power I Shutdown > 1 I 1 I 1 (N)ew or (M)odified from bank including 1 (A) >2 I > 2 I > 1 (P)revious 2 exams <3 I 3 I < 2 (randomly selected) (R)CA > 1 I > 1 I 1 (S)imulator II ES-301, Page 23 of 27 Outline Form ES-301-2 Control Room/In-Plant

....U.IU*OIM Facility:

Peach Bottom Date of Examination:

Exam level: RO 0 SRO-I SRO-U 0 Operating Test Number: Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

• I Safety System 1 JPM Title Type Code

• Function A, D, EN, a. 295037 EA 1.04 -Standby Liquid Control System 1 Inject SBlC (Alternate 1 Path -low SBlC Discharge Pressure)

{pLOR-331CA)

[Set 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] I . c. 239001 A4.01 -Reopen Main Steam Isolation Valves after a Group I I D,l, P,S 3 Isolation

{PlOR-083C}

{2008 NRC Exam) [Set 3] I d. 206000 A2.09 High Pressure Coolant Injection System I Manual I A, l, N, S 4

• System Start (Alternate Path -Suction Valves Fail to Auto Swap on low II CST_Level) (pLOR-343CA)

[Set 1J I e, I 6 I (Alternate Path -load Control DifficultY)iPlOR-322CA)

[Set 4] I f. 264000 A4.04 -Emergency Generators 1 Diesel Generator load Test A, D,S I g. 400000 A4,01 -Component Cooling Water System 1 ECW System N,S 8 Makeup to Tower Usingf:.SW System 3] , h, 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) 4 MSIV (Alternate Path -Removing Fuse Fails To Close MSIV -Unit 3) (PlOR-313PA)

i. 239001 A2.12 -Main and Reheat Steam System 1 Close A Stuck Open A,D,l, R C, D, E, 7 energization

-Unit 2 (T-213-2) (PlOR-075P)

j. 295037 EA 1.01 -Reactor Protection System I Scram Solenoid De-L, R I 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 overlap those tested in the control room. k. 2180000 A2,03 -Backup Instrument Nitrogen to ADS System Startup and

• Type Criteria for RO I SRO-II SRO-U I ES-301 Control Room/In-Plant

.... u'lu*ornC!

Outline Form ES-301-2 Facility:

Peach Bottom Date of Examination:

12/07/2009 Exam level: RO 0 SRO-I SRO-U 0 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* I Safety Function a. 295037 EA 1.04 -Standby Liquid Control System 1 Inject SBlC (Alternate A, D, EN, 1 Path -low SBlC Discharge Pressure)

{pLOR-331CA)

[Set 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 I D,l, P,S 3 Isolation

{PlOR-083C}

{2008 NRC Exam) [Set 3] I d. 206000 A2.09 High Pressure Coolant Injection System I Manual A, l, N, S I 4

• System Start (Alternate Path -Suction Valves Fail to Auto Swap on low II CST_Level) (pLOR-343CA)

[Set 1J -e, I I f. 264000 A4.04 -Emergency Generators 1 Diesel Generator load Test I A, D,S 6 I (Alternate Path -load Control DifficultY)iPlOR-322CA)

[Set 4] g. 400000 A4,01 -Component Cooling Water System 1 ECW System N,S 8 Makeup to Tower Usingf:.SW System 3] , h, 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 MSIV (Alternate Path -Removing Fuse Fails To Close MSIV -Unit 3) (PlOR-313PA)

j. 295037 EA 1.01 -Reactor Protection System I Scram Solenoid De-C, D, E, 7 I energization

-Unit 2 (T-213-2) (PlOR-075P)

L, R k. 2180000 A2,03 -Backup Instrument Nitrogen to ADS System Startup and 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 overlap those tested in the control room.

• Type Codes I Criteria for RO I SRO-II SRO-U (A)lternate path 4-6 I 4-6 I (C)ontrol room (D)irect from bank I :: 8 I < (E)mergency or abnormal in-plant I 1 I > (EN)gineered safety feature I -I > 1 (control room system) (L)ow-Power I Shutdown > 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 I :: 8 I < 4 (E)mergency or abnormal in-plant I 1 I > 1 (EN)gineered safety feature I -I > 1 (control room system) (L)ow-Power I Shutdown > 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 Room/In-Plant

,",u'",r""rTI"" Outline Form ES-301-2 Facility:

Peach Bottom Exam Level: RO D SRO-I D SRO-U I:8J Date of Examination:

12/07/2009 Operating Test Number: !..!.!..!.:::::.-

__ Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title a. b. 217000 A4.03 -Reactor Core Isolation Cooling System 1 Manually Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

Set 2 d. 206000 A2.09 -High Pressure Coolant Injection System 1 Manual System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low e. f. CST Level

____ 37 EA1.01 -Reactor Protection System 1 Scram Solenoid ization -lJllil?{I-213-2) (PLOR-075P) 00 A2.03 -Backup Instrument Nitrogen to ADS System Startup and tion Unit 3 PLOR-271 P Type Code* A, D, EN, L, S A, L, N, S C, D, E, L, R D, E, R @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety 4 7 3 functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may hose tested in the control room.

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

,",u'",r""rTI"" Outline Form ES-301-2 Facility:

Peach Bottom Exam Level: RO D SRO-I D SRO-U I:8J Date of Examination:

12/07/2009 Operating Test Number: !..!.!..!.:::::.-

__ Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title a. b. 217000 A4.03 -Reactor Core Isolation Cooling System 1 Manually Initiate RCIC (Alternate Path -RCIC fails to isolate) (PLOR-332CA)

Set 2 d. 206000 A2.09 -High Pressure Coolant Injection System 1 Manual System Start (Alternate Path -Suction Valves Fail to Auto Swap on Low e. f. CST Level

___ _ 37 EA1.01 -Reactor Protection System 1 Scram Solenoid ization -lJllil?{I-213-2) (PLOR-075P) 00 A2.03 -Backup Instrument Nitrogen to ADS System Startup and tion Unit 3 PLOR-271 P Type Code* A, D, EN, L, S A, L, N, S C, D, E, L, R D, E, R @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety 4 7 3 functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may hose tested in the control room.

• Type Codes Criteria for RO 1 SRO-II SRO-U (A)lternate path 4-6 1 4-6 1 2-3 (C)ontrol room (D)irect from bank I 8 I (E)mergency or abnormal in-plant > 1 1 1 1 1 (EN)gineered safety feature I -I 1 (control room system) (L)ow-Power I Shutdown I 1 I 1 (N)ew or (M)odified from bank including 1 (A) / 2 1 1 (P)revious 2 exams ::3 1 3 I :: 2 (randomly selected) (R)CA .::.1 1 1 / .::. 1 (S)imulator Scenario ES-D-1 Simulation Facility Peach Bottom Scenario No. #1 (new} OpTest No. 2009 NRC Examiners Operators CRS (SRO) URO (ATC) PRO (BOP)

The scenario begins with the reactor at approximately 5% power during a reactor startup. Following 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 lC-91, 5% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event No. No. Type* 1 PRO CRS 2 URO CRS 3 C URO TS PRO CRS 4 PRO CRS 5 TS CRS 6 URO CRS 7 ALL 8 URO CRS C PRO CRS .. 9 Secure drywell purge Power ascension with control rods Drifting control rod (Tech ' *.! Drywell chiller trip / place standby chiller in service ARI power supply failure (Tech Spec) APRM trip with auto scram failure / manual reactor scram Torus leak into secondary containment

/ emergency blowdown Startup level control valve (LCV-8091) failure Turbine bypass jack fails (prevents rapid depressurization to the main condenser)

• (N)ormal, (R)eactlvlty, (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 lC-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 / place standby chiller in service CRS 5 TS CRS ARI power supply failure (Tech Spec) 6 I URO APRM trip with auto scram failure / manual reactor scram CRS 7 M ALL Torus leak into secondary containment

/ 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)eactlvlty, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec Scenario ES-D-1 Simulation Facility Peach Bottom Scenario NC). #2 {new} OpTest No. 2009 NRC Examiners Operators CRS (SRO) URO (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 'D' 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 ATWS 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. No. Type* 1 PRO CRS 2 URO CRS 3 TS CRS 4 C PRO TS CRS 5 URO CRS 6 ALL 7 URO CRS 8 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 A TWS) 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)eacttvlty, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario NC). #2 {new} OpTest No. 2009 NRC Examiners Operators CRS (SRO) URO (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 'D' 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 ATWS 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 A TWS) 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)eacttvlty, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec Scenario ES-D-1 Simulation Facility Peach I3l llVI Scenario No. #3 (modified)

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

The scenario begins with the reactor at approximately 88% power and HPCI out of service due to emergent maintenance.

Following shift turnover, the crew will perform ST-O-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 drywell 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 Tripsllsolations to Satisfy Tech SpecfTRM Requirements".

When this is complete, the 'A' 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 Bus will trip on overcurrent, causing a loss ofthe 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. A small reactor coolant leak inside the drywell will be greater than the capacity of RCIC (the only available 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 RCIC flow controller will complicate efforts to feed with RCIC and require the operator to transfer RCIC 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 Blowdown".

Low pressure ECCS will be available to recover reactor level. Initial IC-93, 88% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Type*No. Description 1 Main Turbine stop valve functional test I RPS failure (Tech Spec) TS CRS N PRO 2 C URO Service Water pump trip 1 manual start of the standby pump CRS 3 Drywell 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 Loss of #2 auxiliary bus 1 loss of condensate

& feedwater I reactor coolant leak inside the drywell 5 M ALL 6 RPS faiture requires ARI to scram the reactor CRS C URO RCIC flow controller fails in automatic 1 transfer to manual control CRS 7 I URO Containment spray logic failure hampers effort to spray the CRS I PRO 8 containment, requiring crew to use alternate RH R loop ..* (N)ormal, (R)eactlvlty, (I)nstrument. (C)omponent, (M)aJor, (TS) Tech Spec Scenario Outline ES-D-1 Simulation Facility Peach I3l llVI 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 and HPCI out of service due to emergent Summary maintenance.

Following shift turnover, the crew will perform ST-O-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 drywell 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 Tripsllsolations to Satisfy Tech SpecfTRM Requirements".

When this is complete, the 'A' 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 Bus will trip on overcurrent, causing a loss ofthe 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. A small reactor coolant leak inside the drywell will be greater than the capacity of RCIC (the only available 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 RCIC flow controller will complicate efforts to feed with RCIC and require the operator to transfer RCIC 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 Blowdown".

Low pressure ECCS will be available to recover reactor level. Initial IC-93, 88% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 N PRO Main Turbine stop valve functional test I RPS failure (Tech Spec) TS CRS 2 C URO Service Water pump trip 1 manual start of the standby pump CRS 3 I PRO Drywell 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 faiture requires ARI to scram the reactor CRS 7 I URO RCIC flow controller fails in automatic 1 transfer to manual control CRS 8 I PRO Containment spray logic failure hampers effort to spray the CRS containment, requiring crew to use alternate RH R loop .. * (N)ormal, (R)eactlvlty, (I)nstrument. (C)omponent, (M)aJor, (TS) Tech Spec