ML15169A652

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Draft - Outlines (Folder 2)
ML15169A652
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 01/23/2015
From: Millard C
Exelon Generation Co
To: Peter Presby
Operations Branch I
Shared Package
ML14254A311 List:
References
TAC U01911
Download: ML15169A652 (34)
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Text

ES-401 Written Examination Outline Form ES-401-1 Peach Bottom Facility:

ILT 13-1 2015 NRC Date of Exam:

03/23/15 Exam RO Kl A Category Points SRO-Only 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 G*

Total

1.

1 3

4 3

4 3

3 20 2

5 7

Emergency 2

1 1

2 1

1 1

7 1

2 3

Plant Tier Evolutions Totals 4

5 5

5 4

4 27 3

7 10 1

3 2

3 2

3 2

2 3

2 2

2 26 3

2 5

2.

Plant 2

1 1

1 0

3 1

1 1

1 1

1 12 0

1 2

3 Systems Tier Totals 4

3 4

2 6

3 3

4 3

3 3

38 4

4 8

3. Generic Knowledge & Abilities 1

2 3

4 l

2 3

4 10 7

Categories 3

3 2

2

")

l 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 Kl A category shall not be less than two).

2.

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-I 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. l.b ofES-401, for guidance regarding elimination of inappropriate Kl A 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 KAs having an importance rating (IR) of2.5 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 I and 2 from the shaded systems and KIA categories.

7.*

The generic (G) K/As in Tiers I 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. I.b of ES-401 for the applicable K/A's

8.

On the following pages, enter the Kl A numbers, a brief description of each topic, the topics' importance ratings (IR) 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. 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 K/As that are linked to IOCFR55.43

ES-401 2

Form ES-401-1 ILT 13-1 2015 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 ion K1 K2 K3 A1 A2 G

KJA Topic(s) 295016 Control Room 2.2.14 *-Knowledge of the process for Abandonment I 7 x

contmlling equipment conliguration or status.

AA2.04 - Abilitv to determine and/or 295020 lnadvcrl.:nt Containment x

interpret the following as they apply to Jsc1Jatic111 I 5 INADVLRTENT COl'\\TAl\\./MlKI" ISOL.AT!Ol'\\ :Reactor Pressure EA2.04 - Ability lo determine and/or 2950J I Rcilcllir I.ow Wakr L.evel I x

interpret the following us they apply to 2

REACTOR LOW WATER LEVEL

  • Atkquate core cooling 2.2. l 8 -Kntiwledgc of the process l(ir 295030 I,ciw Suppression Poc1I x

managing maintenance activities during shut W ntcr Levd I 5 down operatitms, such as risk assc,sments, work nrioriti1:alion, de.

295019 Partial or Total Loss of x

2.1.20 - Conduct of Operations: Abiliry to Inst. Air/ 8 interpret and execute procedure steps.

295003 Partial or C:ompkk Lo'5 of 2.1.2 - Conduct of Operations: Knowkdgo:

x ofopcratcir responsibilities during all modes AC if11lant 011.:ration.

295018 Partial or Total Los'> nf 2.1.19 - Conduct nf Operations: Ability to ccw /8 x

us<.: plant computers to evaluate system tir cc1mpo11ent status.

AKl.02 - Knowledge of the operational 295001 Partial or Complete Loss implications of the following concepts as of Forced Core Flow Circulation/ I x

they apply to PARTIAL OR COMPLETE

&4 LOSS OF FORCED CORE FLOW ClRCULA TION : Power/flow distribution AKI.01 - Knowledge of the operational implications of the following concepts as 295018 Partial or Complete Loss x

they apply to PARTIAL OR COMPLETE of Component Cooling Water /8 LOSS OF COMPONENT COOLING WATER :Effects on component/system operations EKl.02 - Knowledge of the operational 295031 Reactor Low Water Level x

implications of the following concepts as 12 they apply to REACTOR LOW WATER LEVEL : Natural circulation: Plant-Specific EK2.09 - Knowledge of the interrelations 295037 SCRAM Conditions between SCRAM CONDITION PRESENT Present and Reactor Power Above x

AND REACTOR POWER ABOVE APRM APRM Downscale or Unknown/ l DOWNSCALE OR UNKNOWN and the following: Reactor water level AK2.02 - Knowledge of the interrelations 295005 Main Turbine Generator x

between MAIN TURBINE GENERATOR Trip/ 3 TRIP and the following: Feedwater temperature EK2.09 - Knowledge of the interrelations 295024 High Drywell Pressure I 5 x

between HIGH DRYWELL PRESSURE and the following: Suppression pool makeup: Plant-Specific AK3.06 - Knowledge of the reasons for 295003 Partial or Complete Loss x

the following responses as they apply to of AC 16 PARTIAL OR COMPLETE LOSS OF A.C. POWER: Containment isolation EK3.07 - Knowledge of the reasons for 295030 Low Suppression Pool the following responses as they apply to x

LOW SUPPRESSION POOL WATER Water Level I 5 LEVEL: NPSH considerations for ECCS pumps Imp.

Q#

4J 76 3.9 77 4.8 78 3.9 79 4.6 80 4.4 81 3.8 82 3.3 39 3.5 40 3.8 41 4.0 42 2.9 43 2.9 44 3.7 45 3.5 46

ES-401 2

Form ES-401-1 ILT 13-1 2015 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE #I Name Safety Function K1 K2 K3 A1 A2 G

KIA Topic(s)

AK3.02 - Knowledge of the reasons for 295019 Partial or Total Loss of the following responses as they apply to x

PARTIAL OR COMPLETE LOSS OF Inst. Air/ 8 INSTRUMENT AIR : Standby air compressor operation EA 1.01 - Ability to operate and/or 295026 Suppression Pool High monitor the following as they apply to x

SUPPRESSION POOL HIGH WATER Water Temp. I 5 TEMPERATURE: Suppression pool cooling EA 1.04 - Ability to operate and/or 295025 High Reactor Pressure I 3 x

monitor the following as they apply to HIGH REACTOR PRESSURE: HPCI:

Plant-Specific AA 1.04 - Ability to operate and/or 295006 SCRAM I I x

monitor the following as they apply to SCRAM : Recirculation system AA2.04 - Ability to determine and/or 295004 Partial or Total Loss of DC x

interpret the following as they apply to Pwr/6 PARTIAL OR COMPLETE LOSS OF D.C. POWER: System lineups EA2.04 - Ability to determine and/or 295028 High Drywell Temperature x

interpret the following as they apply to 15 HIGH DRYWELL TEMPERATURE:

Drywell pressure EA2.04 - Ability to determine and/or 295038 High Off-site Release Rate x

interpret the following as they apply to 19 HIGH OFF-SITE RELEASE RATE:

Source of off-site release 2.1.14 -Knowledge of criteria or 295021 Loss of Shutdown Cooling x

conditions that require plant-wide 14 announcements, such pump starts, reactor trios, mode chani:ies, etc..

2.4.8 - Emergency Procedures I Plan:

600000 Plant Fire On-site I 8 x

Knowledge of how abnormal operating procedures are used in conjunction with EOP's.

2.2.42 - Equipment Control:: Ability to 700000 Generator Voltage and x

recognize system parameters that are Electric Grid Disturbances entry-level conditions for Technical Specifications.

AA 1.01 - Ability to operate and/or 295023 Refueling Ace Cooling x

monitor the following as they apply to Mode I 8 REFUELING ACCIDENTS : Secondary containment ventilation AK2.03 - Knowledge of the interrelations 295016 Control Room x

between CONTROL ROOM Abandonment I 7 ABANDONMENT and the following:

Control room HVAC KIA Category Totals:

3 4

3 4

3/2 3/5 Group Point Total:

Imp.

Q#

3.5 47 4.1 48 3.8 49 3.1 50 3.2 51 4.1 52 4.1 53 3.1 54 3.8 55 3.9 56 3.3 57 2.9 58 I 20/7

ES-401 3

ILT 13-1 2015 NRC Exam Written Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE # I Name Safety Function K1 K2 K3 A1 A2 G

KIA Topic(s)

Imp.

Q#

1-/\\2.0 I - Ability to determine and/or 295035 Secondary Containmellt interpret the following as they apply to x

SECONDARY CONTAINMENT HIGH 3_()

83 High Differential Pressure I 5 DIFFERENTIAL PRESSURE: Secondary containment pressure: Plant-Specific 295017 High Ofl~site Release Rule x

2.2.12 - Equipment Cm1trol: Knowledge ol' 4.1 84

!'I surveillance procedures.

2.1.23-Ability to perform specitic system 295015 lnwmplctc SCRAlVl I I x

and int.:gratcd plant procedures during all 4.4 85 modes of plant opcrntion.

EKl.02 - Knowledge of the operational 295032 High Secondary implications of the following concepts as x

they apply to HIGH SECONDARY 3.6 59 Containment Area Temperature I 5 CONTAINMENT AREA TEMPERATURE: Radiation releases AK2.0l - Knowledge of the interrelations 295015 Incomplete SCRAM I I x

between INCOMPLETE SCRAM and the 3.8 60 following: CRD hvdraulics AK3.0l - Knowledge of the reasons for 295012 High Drywell x

the following responses as they apply to 3.5 61 Temperature I 5 HIGH DRYWELL TEMPERATURE:

Increased drywell cooling AA!.05 - Ability to operate and/or monitor 295002 Loss of Main Condenser x

the following as they apply to LOSS OF 3.2 62 Vacuum I 8 MAIN CONDENSER VACUUM :Main Turbine AA2.0l - Ability to determine and/or 295014 Inadvertent Reactivity x

interpret the following as they apply to 4.1 63 Addition I I INADVERTENT REACTIVITY ADDITION : Reactor power 295029 High Suppression Pool 2.4.3 I - Emergency Procedures I Plan:

Water Level I 5 x

Knowledge of annunciator alarms, 4.2 64 indications, or response procedures.

EK3.02 - Knowledge of the reasons for the 295035 Secondary Containment following responses as they apply to x

SECONDARY CONTAINMENT HIGH 3.3 65 High Differential Pressure I 5 DIFFERENTIAL PRESSURE : Secondary containment ventilation response KIA Category Totals:

1 1

2 1

1/1 1/2 Group Point Total:

I 713

ES-401 System #I Name K

K K

1 2

3 215004 Source Range Monitlir 2112002 UPS (AC/DC) 264000 EDGs 241000 Reactor/'furbinc Pressure R..:gulating System 2~9002 Reactor Water Level Control 259002 Reactor Water Level x

Control 209001 LPCS x

203000 RHR/LPCI: Injection Mode x

400000 Component Cooling x

Water 4

Form ES-401-1 ILT 13-1 2015 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K

K K

A A2 A

A G

4 5

6 1

3 4

i\\2.0J - Ability to (a) predict the impacts of the following on the SOURCE R;\\NGE MONlTOR (SRM) SYSTEM: and (b) based on x

those predictions. use procedures to correct. control. or mitigak the consequences uf those abnormal conditions or operations: Power supply degraded

/\\2.01 -Ability to (a) predict the impacts of the following on the L:NIN"l"ERRLJPTABI.I.' POWLR SUPPLY (A.C/D.C.): and (b) x based on those predictions, use pniccdurcs to C<.HTcct, control. or mitigate the consequences of those abnormal cnnditions c>r c>pcrations:

Under voltage x

2.4.11-Knnwledge of abnormal condition procedures.

2.1.7 -Ability to.:valuate plant pcrformanc.x and ma1'e operational:

x judgments based on operating characteristics. r..:actor behavior, and instrument interpretation.

i\\2.02 - Ability to (al predict the impacts of the following on the Rb\\CTOR WATER LEVFL CONTROL SYSTEiVI; and (b) x based on thc>se predictions. use procedures to correct. C<mtrol, or mitigate the consequences of those abnormal conditions or operatiLms:

Loss ot' any number or reactor feedwater flow inputs Kl.OS - Knowledge of the physical connections and/or cause-effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following:

Reactor feedwater system K1.05 - Knowledge of the physical connections and/or cause-effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: Automatic depressurization system K2.01 - Knowledge of electrical power supplies to the following:

Pumps K2.02 - Knowledge of electrical power supplies to the following:

CCWvalves Imp Q#

2.9 86 2.X 8"'

I 4.2 88 4.7 89 3.4 90 3.6 I

3.7 2

3.5 3

2.9 4

ES-401 System # I Name K

K K

1 2

3 211000 SLC x

223002 PCIS/Nuclear Steam x

Supply Shutoff 264000 EDGs 218000 ADS 215005 APRM I LPRM 26200 I AC Electrical Distribution 215003 !RM 211000 SLC 209001 LPCS 4

Form ES-401-1 ILT 13-1 2015 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K

K K

A A2 A

A G

4 5

6 1

3 4

K3.01 - Knowledge of the effect that a loss or malfunction of the STANDBY LIQUID CONTROL SYSTEM will have on following:

Ability to shutdown the reactor in certain conditions K3.12 - Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following: High pressure coolant injection: Plant-Specific K4.08 - Knowledge of EMERGENCY GENERATORS x

{DIESEL/JET) design feature(s) and/or interlocks which provide for the following: Automatic startup K4.01 - Knowledge of AUTOMATIC DEPRESSURIZA TION SYSTEM x

design feature(s) and/or interlocks which provide for the following: Prevent inadvertent initiatior of ADS logic K5.05 - Knowledge of the operational implications of the following concepts as they apply x

to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM :

Core flow effects on APRM trip setpoints K5.01 - Knowledge of the operational implications of the following concepts as they x

apply to AC. ELECTRICAL DISTRIBUTION: Principle involved with paralleling two AC. sources K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the x

INTERMEDIATE RANGE MONITOR (IRM) SYSTEM :

Reactor protection system (power supply): Plant-Specific K6.03 - Knowledge of the effect that a loss or malfunction of the x

following will have on the STANDBY LIQUID CONTROL SYSTEM : A.C. power A 1.04 - Ability to predict and/or monitor changes in parameters x

associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including: Reactor pressure Imp Q#

43 5

3.6 6

3.8 7

3.7 8

3.6 9

3.1 10 3.8 II 3.2 12 3.7 13

ES-401 System # I Name K

K K

1 2

3 218000 ADS 206000 HPCI 217000 RCIC 239002 SRVs 261000 SGTS 212000 RPS 215004 Source Range Monitor 300000 Instrument Air 263000 DC Electrical Distribution 4

Form ES-401-1 ILT 13-1 2015 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K

K K

A A2 A

A G

4 5

6 1

3 4

A 1.06 - Ability to predict and/or monitor changes in parameters associated with operating the x

AUTOMATIC DEPRESSURIZATION SYSTEM controls including: Suppression pool temperature A2.04 - Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM ; and (b) x based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: AC. failures: BWR-2,3,4 A2.10 - Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ;

and (b) based on those x

predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Turbine control system failures A3.02 - Ability to monitor automatic operations of the x

RELIEF/SAFETY VALVES including: SRV operation on high reactor pressure A3.03 - Ability to monitor automatic operations of the x

STANDBY GAS TREATMENT SYSTEM including: Valve operation A4.14 - Ability to manually x

operate and/or monitor in the control room: Reset system following system activation A4.03 - Ability to manually x

operate and/or monitor in the control room: CRT displays:

Plant-Specific 2.4.31 - Knowledge of x

annunciator alarms, indications, or response procedures.

2.4.21 - Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety x

functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Imp Q#

4.1 14 2.7 15 3.1 16 4.3 17 3.0 18 3.8 19 2.9 20 4.2 21 4.0 22

ES-401 System # I Name K

K K

1 2

3 205000 Shutdown Cooling 215004 Source Range Monitor x

261000 SOTS 217000 RCJC x

Kl A Category Totals:

3 2

3 4

Form ES-401-1 ILT 13-1 2015 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K

K K

A A2 A

A G

4 5

6 1

3 4

KS.03 - Knowledge of the operational implications of the following concepts as they apply x

to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) : Heat removal mechanisms K3.01 - Knowledge of the effect that a loss or malfunction of the SOURCE RANGE MONITOR (SRM) SYSTEM will have on following: RPS A2.04 - Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on x

those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High train moisture content K1.03 - Knowledge of the physical connections and/or cause-effect relationships between REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: Suppression pool 2

3 2

2 3/3 2

2 212 Group Point Total:

Imp Q#

2.8 23 3.4 24 2.5 25 3.6 26 I

2615

ES-401 System # I Name K

K K

1 2

3 215001 Traversing ln-cor.:

Probe 202002 Recircnlation Flow Control 233000 Fuel Pool Cooling/Cleanup 215002 RBM x

272000 Radiation Monitoring x

256000 Reactor Condensate x

290001 Secondary CTMT 268000 Radwaste 239001Main and Reheat Steam System 204000 RWCU 230000 RHR/LPCI:

Torus/Pool Spray Mode 5

Form ES-401-1 ILT 13-1 2015 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 2 K

K K

A A2 A

A G

4 5

6 1

3 4

A2.08 - Ability to (a) pr~dict th~

impacts of the following on the TRA VF RSI NG IN-CORI PROBF

and (b) based on those x

predictions. use procedures to correct, rnntrol, or mitigate the consequ.:nces ol"thosc abnormal conditions c>r 1ipcrations: Failure to retract to shield: (Not-13\\VR 1) x 2.2.40 *Ability to apply Technical Specifications for a svstcm.

2.2.22 - Equipment Control:

x Knowledge of limiting conditions for operations and safety limits.

Kl.06 - Knowledge of the physical connections and/or cause-effect relationships between ROD BLOCK MONITOR SYSTEM and the following: Control rod selection: BWR-3,4,5 K2.0l - Knowledge of electrical power supplies to the following:

Main steamline radiation monitors K3.08 - Knowledge of the effect that a loss or malfunction of the REACTOR CONDENSATE SYSTEM will have on following:

SJAE KS.01 - Knowledge of the operational implications of the x

following concepts as they apply to SECONDARY CONTAINMENT: Vacuum breaker operation KS.01 - Knowledge of the operational implications of the x

following concepts as they apply to RADWASTE : Units of radiation, dose and dose rate K6.02 - Knowledge of the effect that a loss or malfunction of the x

following will have on the MAIN AND REHEAT STEAM SYSTEM : Plant air system A 1.07 - Ability to predict and/or monitor changes in parameters x

associated with operating the REACTOR WATER CLEANUP SYSTEM controls including:

RWCU drain flow A2.14 - Ability to (a) predict the impacts of the following on the RHR/LPCI:

TORUS/SUPPRESSION POOL SPRAY MODE; and (b) based x

on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low (or negative) suppression pool pressure durino system operation Imp.

Q#

2.9 91 4.7 92 4.7 93 3.0 27 2.5 28 2.8 29 3.3 30 2.7 31 3.2 32 2.9 33 3.2 34

ES-401 System #I Name K

K K

1 2

3 233000 Fuel Pool Cooling/Cleanup 201002 RMCS 25900 I Reactor Feedwater 20200 I Recirculation Kl A Category Totals:

1 1

1 5

Form ES-401-1 ILT 13-1 2015 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 2 K

K K

A A2 A

A G

4 5

6 1

3 4

A3.03 - Ability to monitor automatic operations of the x

FUEL POOL COOLING AND CLEAN-UP including: System indicatina liahts and alarms A4.02 - Ability to manually x

operate and/or monitor in the control room: Emergency in/notch override switch 2.1.19 - Conduct of Operations:

x Ability to use plant computers to evaluate system or component status.

K5.01 - Knowledge of the operational implications of the x

following concepts as they apply to RECIRCULATION SYSTEM:

Indications of pump cavitation 0

3 1

1 1/1 1

1 1/2 Group Point Total:

Imp.

Q#

2.6 35 3.5 36 3.9 37 2.7 38 I

12/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility:

ILT 13-1 2015 NRC Exam Date:

03/23/15 Category KIA#

RO SRO-Only Topic IR Q#

IR Q#

2.1.41 Knowledge of the refueling process.

3.7 94 2.1.25 Ability to interpret reference materials, such 4.2 99 as qraphs, curves, tables, etc.

1.

2.1.2 Knowledge of operator responsibilities during 4.1 66 Conduct all modes of plant operation.

of Operations 2.1.27 Knowledge of system purpose and I or 3.9 67 function.

2.1.3 Knowledge of shift or short-term relief 3.7 75 turnover practices.

Subtotal 3

2 Ability to determine the expected plant 2.2.15 configuration using design and configuration 4.3 95 control documentation, such as drawings, line-ups, tag-outs, etc.

2.

2.2.7 Knowledge of the process for conducting 2.9 68 special or infrequent tests.

Equipment Knowledge of the process for managing Control 2.2.20 troubleshootinq activities.

2.6 69 Ability to perform pre-startup procedures for 2.2.1 the facility, including operating those controls 4.5 74 associated with plant equipment that could affect reactivitv.

Subtotal 3

1 2.3.11 Ability to control radiation releases.

4.3 96 Knowledge of radiological safety procedures pertaining to licensed operator duties, such 2.3.13 as response to radiation monitor alarms, 3.8 98 containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

3.

Radiation Ability to use radiation monitoring systems, Control 2.3.5 such as fixed radiation monitors and alarms, 2.9 70 portable survey instruments, personell monitoring equipment, etc.

2.3.4 Knowledge of radiation exposure limits under 3.2 71 normal or emerqencv conditions.

Subtotal 2

2

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility:

IL T 13-1 2015 NRC Exam Date:

03/23/15 Category KIA#

Topic RO SRO-Only IR Q#

IR Q#

2.4.18 Knowledge of the specific bases for EOPs.

4.0 97 Ability to verify system alarm setpoints and 2.4. 50 operate controls identified in the alarm 4.0 100 response manual.

4.

Knowledge of events related to system Emergency operation I status that must be reported to Procedures I 2.4.30 internal organizations or external agencies, 2.7 72 Plan such as the state, the NRC, or the transmission system operator.

2.4.14 Knowledge of general guidelines for EOP usage.

3.8 73 Subtotal 2

2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/A's Form ES-401-4 Tier I Group Randomly Reason for Rejection Selected Kl A RO 2I1 300000 I 2.4.41 EAL classification task for an RO question. Replaced with Q #21 KIA 300000 2.4.31.

R02/2 290001 I K4.01 Could not develop a question above General Employee Q#30 Training for this KIA. Replaced with KIA 290001 K5.01.

R02/2 201004 I K6.02 Software sampled RSCS. RSCS is no longer a PBAPS Q #32 system. Replaced with KIA 239001 K6.02.

RO 1I1 295028 I AKl.01 295028 was oversampled. Replaced with KIA 295018 Q#40 AKl.01.

RO 1I1 295021 2.1.19 G 2.1.19 had been selected three times. Replaced with Kl A Q#54 295021 2.1.14.

RO 1 /2 295034 I AAl.05 Too many similar KIA for Secondary Containment selected.

Q#62 Replaced with KIA 295002 AAl.05 SRO 1I1 295016 I AA2.05 Could not make a SRO question linking the Kl A. Replaced Q# 76 with KIA 295016 2.2.14.

SRO 1I1 295004 I AA2.04 Same Kl A selected for Q #51. Replaced with Kl A 295020 Q#77 AA2.04.

SRO 1I1 295030 I 2.2.39 Could not develop a SRO question. Replaced with Kl A Q#79 295030 2.2.18.

SRO 1I2 295015 I 2.1.28 Not a SRO task. Could not develop a SRO question.

Q#85 Replaced with KIA 295015 2.1.23.

SRO 2 I 1 264000 I 2.4.4 No EOP entry conditions bases on Emergency Diesel Q#88 Generators. Replaced with Kl A 26400 2.4.11.

SRO 2 I 1 217000 2.1.14 217000 was already sampled twice. 2.1.14 is not an SRO Q#89 task. Replaced with KIA 241000 2.1.7.

SRO 2 I 2 202002 I 2.2.39 Not a SRO task. Could not develop a SRO question.

Q#92 Replaced with KIA 202002 2.2.40.

SR03 2.3.6 Could not develop a SRO question. Replaced with 2.3.13.

Q#98

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 03/23/2015 Examination Level: RO [8J SRO 0 Operating Test Number: 2015 NRC Administrative Topic Type Describe activity to be performed (See Note)

Code*

D, P, R G2.1.45 (4.3) - Manually Calculate Drywell Bulk Average Conduct of Operations Temperature with Failed Temperature Points (PLOR-241 C) (2011 NRC)

Conduct of Operations D,S G2.1.31 (4.6) - Perform an APRM Scram Margin Check (PLOR-219C)

G2.2.41 (3.5) - Determine Status of Instrument Nitrogen Equipment Control D,R Compressor Discharge Solenoid Valve Using Station Piping and Instrumentation Drawings (PLOR-220C)

Radiation Control N/A Not Required G2.4.29 (3.1) - Emergency Response Organization Emergency Plan N,R Response Augmentation Using the Everbridge Web-based Call Out System (PLOR-92C)

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 Administrative Topics Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 03/23/2015 Examination Level: RO D SRO [8:]

Operating Test Number: 2015 NRC Administrative Topic Type Describe activity to be performed (See Note)

Code*

Conduct of Operations D,R G2.1.7 (4.7) - Resolution of Thermal Limit Violation (PLOR-21 SC)

Conduct of Operations D, R G2.1.5 (3.9) - Evaluate Overtime Work Request (PLOR-279C)

Equipment Control D,R G2.2.6 (3.6) - Review a Temporary Procedure Change -

Change of Intent (PLOR-222C)

Radiation Control D, P, R G2.3.13 (3.8) - Perform Primary Containment Purge I Vent Isolation Valve Cumulative Log (PLOR-256C)

(2013 NRC)

Emergency Plan N,R G2.4.41 (4.6) - Make EAL Classification And State/Local Notifications for ALERT - Inability to Maintain Cold Shutdown (PLOR-153C)

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

Peach Bottom 2015 ILT NRC Exam Summary of JPM Tasks JPM Cateaorv Title SettinQ I Summary RO Administrative Conduct of Manually Calculate Classroom.

Operations Drywell Bulk Average Temperature with The candidate is directed to perform certain steps of Failed Temperature RT-0-40C-530-2, "Drywell Temperature Points Monitoring", and document the results on the JPM cue sheet. They must recognize that all of the temperature points in Temperature Zone Number 4 of Tl-2501 are out of service making the calculation of Bulk Average temperature INVALID. They will have to manually calculate Drywell bulk average temperature by using Tl-2501, Point 136, and add 10°F as required by the RT and then report to the CRS that ON-120 "High Drywell Temperature" should be entered due to Approximate Drywell Bulk Average Temperature greater than 140°F.

Perform an APRM Simulator.

Scram Margin Check The candidate will be directed to perform an APRM scram margin check for APRM #4. The candidate will use the APRM NUMAC drawer to get the information needed to perform the scram margin calculation. The candidate reports the amount of scram margin to the CRS.

Equipment Determine Status of Classroom.

Control Instrument Nitrogen Compressor The candidate will be directed to determine the Discharge Solenoid status of a degraded component in the Instrument Valve Using Station Nitrogen System using the appropriate Piping and Piping and Instrumentation Drawings (P&IDS). The candidate Instrumentation should determine that:

Drawings

  • Solenoid Valve SV-5232A should be energized with the 3AK037 Compressor running
  • Solenoid Valve SV-5232A is currently closed (Valve is energized to close)
  • Starting the 3BK037 Compressor will allow a comparison of SV-5232A and SV-5232B 2015 Peach Bottom NRC Exam JPM Summary Page 1

provided the 3BK037 Compressor is run for longer than 0.5 seconds.

Emergency Emergency Response Classroom with computer access.

Plan Organization

Response

The candidate will be directed to initiate Emergency Augmentation Using Response Organization (ERO) Activation (Call-out) the Everbridge Web-in accordance with EP-AA-112-1 OO-F-06 Section 1, based Call Out Steps 1.3 - 1.11. This will require using a computer System to access the web-based Everbridge callout system.

SRO Administrative Conduct of Resolution of a Classroom.

Operations Thermal Limit Violation The candidate will be requested to review an official 30 Monitor Case (P1) edit and document on the cue sheet any unsatisfactory data points and document any actions that are required by applicable procedures or Technical Specifications I Technical Requirements Manual. The candidate should determine the following:

  • That MFLCPR is above 1.000 in one location (19-20)
  • GP-13, "Resolution of Thermal Limit Violations" needs to be entered
  • Reactor power must be reduced with the assistance of Reactor Engineering in accordance with GP-5, "Power Operations" to restore MFLCPR to below 1.000
  • If MFLCPR is not restored to below 1.000 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, thermal power must be reduced to below 25% RTP within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
  • Determine that Technical Specification LCO 3.2.2, "Minimum Critical Power Ratio (MCPR)"

Condition A entry is required 2015 Peach Bottom NRC Exam JPM Summary Page 2

Evaluate Overtime Classroom.

Work Request The candidate will be requested to determine whether or not the operator can cover the requested shift and whether any work hour limits have already been violated. The candidate will be given the following information:

  • Asked to work on a scheduled day off.
  • One month of previous work history.

With this information the candidate should determine that working would result in a work hour violation and determine that there was a violation in the previous work schedule Equipment Review a Temporary Classroom.

Control Procedure Change that Contains a A Temporary Change has been prepared for ST Change of Intent 080-520-2 "Reactor Vessel Head Flange Temperature Surveillance" step 6.1.2, to ensure reactor vessel flange and head flange temperature is greater than 65 degrees F. The candidate is assigned as the SRO Reviewer for the temporary change to the surveillance procedure. The candidate should identify that the proposed change to step 6. 1.2 constitutes a "change of intent" in accordance with procedure AD-PB-101-1003.

Radiation Perform Primary Classroom.

Control Containment Purge I Vent Isolation Valve The candidate is requested to perform the Plant Cumulative Log Staff review and approval of ST-0-007-560-2, "Primary Containment PurgeNent Isolation Valve Cumulative Hour Log", and annotate any errors on procedure copy, and inform CRS of any issues/errors. The candidate should recognize multiple calculation errors on Data Sheet 1 and determines that the "Accumulated Total Time Since Beginning of Year" is 93 Hr, 22 Min versus 80 Hr, 22 Min. They should report to Shift Management that the "Accumulated Total Time Since Beginning of Year" is greater than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />.

2015 Peach Bottom NRC Exam JPM Summary Page 3

Emergency Make EAL Classroom.

Plan Classification And State/Local The candidate is directed, as the Emergency Notifications for Director, to make an Emergency Action Level (EAL)

ALERT - Inability to classification and complete the State/Local Event Maintain Cold Notification form (if required) for an ALERT. The Shutdown conditions will require the candidate to reference the more infrequently used COLD MATRIX of the EALs.

EAL MU1 is referenced due to the loss of Off-site power. This is a correct EAL classification but not the highest classification that exists.

EAL CU5 is referenced. CU5 is a correct classification but not the highest classification.

EAL CA5 is the correct EAL classification for Unit 3.

The Reactor Head is removed which means that the RCS is not intact. Primary containment is not intact because of fuel handling and Reactor building differential pressure is positive therefore Secondary Containment is not established CONTROL ROOM SYSTEMS Reactor Operator Simulator.

Actions on a Recirculation Pump The candidate will be directed to trip the "A" Recirc Trip (Alternate Path -

Pump and perform the Immediate Operator Actions Thermal Hydraulic of OT-112, "Unexpected/Unexplained Change in Instability Exists Core Flow". As control rods are being inserted the Without Operable candidate should observe power oscillations OPRM System) indicating Thermal Hydraulic Instability (THI) and then should shut down the reactor by taking the reactor mode switch to the "shutdown" position.

Startup HPCI in the Simulator.

CST to CST Mode (Alternate Path -

The candidate will be directed you to start up HPCI Turbine Exhaust in the CST to CST Mode in accordance with Rapid Diaphragm High Response Card (RRC) 23.1-2, "HPCI System Pressure)

Operation During a Plant Event" and lower reactor pressure to 500 psig. The candidate should recognize that annunciator 221 E-3 is alarming for a 2015 Peach Bottom NRC Exam JPM Summary Page4

HPCI Exhaust Diaphragm rupture condition and that the HPCI system needs to be shut down. HPCI can be shut down using either SO 23.2.A-2, "HPCI System Shutdown" or the Rapid Response procedure RRC 23.1-2.

Operate the High Simulator.

Pressure Service Water System The candidate is directed to place Unit 2 HPSW in (Alternate Path - Low service with the 2A HPSW pump supplying the 2D Heat Exchanger Delta RHR heat exchanger using SO 32.1.A-2 "High P)

Pressure Service Water System Startup and Normal Operations". The candidate should recognize that the 2D RHR heat exchanger differential pressure is less than 20 psid and "RHR Heat Exchanger Tube to Shell Low Press" alarm 226 E-4 is NOT clear and must start the 2C HPSW pump to clear the low DP condition, then throttle open M0-2-10-898 to raise total HPSW flow to between 6600 and 10600 gpm.

Drywell Venting via Simulator.

the 2 Inch Vents (Alternate Path - Main The candidate will be directed to maximize venting Stack High Radiation) the drywell via the 2" vents in accordance with OT-101, Step 3. 7 "High Drywell Pressure" to lower drywell pressure to 0.70 psig. Once drywell venting is in progress, PR/RR-2-02-3-4048 will slowly rise and annunciator 003 D-2 "Main Stack Radiation High" will alarm. Examinee must terminate Drywell venting at this time by closing several valves.

Load Diesel Simulator.

Generator to 500kW (Alternate Path -

The candidate is directed to synchronize the E-4 Differential/G round Diesel Generator to the E-43 Bus and pick up 500 Fault)

KW in accordance with Section 4.2 of SO 52A.1.B, "Diesel Generator Operations". During the diesel generator run alarm "E-4 Diesel Generator Differential and Ground" is received. Guidance from the alarm response card will have the candidate trip the output breaker and trip the diesel generator.

2015 Peach Bottom NRC Exam JPM Summary Page 5

Initialize the Rod Simulator.

Worth Minimizer The candidate will be directed to initialize the Rod Worth Minimizer (RWM) using procedure SO 62A.1.A-2. This is accomplished using various pushbuttons for system initialization and system diagnostics, and on the RWM computer touch screen acknowledging all RWM messages and verifying system permissives are active.

ECW System Makeup Simulator.

to Emergency Cooling Tower Using ESW The candidate will be directed to makeup to the System Emergency Cooling Tower to a level of 18ft 3in, then restore to a normal lineup, using the "A" HPSW Pump I Heat Exchanger IAW SO 48.7.A "Emergency Cooling Water System Makeup To Tower Using A High Pressure Service Water Pump.

The task will require starting the HPSW pump and performing several motor-operated valve manipulations to send water to the Emergency Cooling Tower.

Manually Place Simulator.

Standby Gas Treatment System on The candidate is directed to place SBGT on Equipment Cell Equipment Cell Exhaust using SO 9A.7.G,Standby Exhaust Gas Treatment System (SBGT) Manual Startup on Equipment Cell Exhaust". Several SBGT and reactor building ventilation system dampers will need to be manipulated and the 'A' SBGT fan placed in service. Reactor building pressure differentials and SBGT system flow will need to be verified.

IN-PLANT Alternate RPV RCA.

SYSTEMS Injection Using the Standby Liquid The candidate will be directed to perform procedure Control Test Tank T-244-2, "Alternate Injection Using the SBLC Test Tank" up to and including Step 4.6. This task requires going to the Emergency Operating Proce-dure Tool Locker and obtaining the T-244 Tool Kit and 50 foot length of air hose. The candidate must then go to reactor building elevation 195' to the SBLC system area, install the section of hose, and verify/manipulate several hand operated valves.

2015 Peach Bottom NRC Exam JPM Summary Page 6

RPV Venting During RCA.

Containment Flooding The candidate will be directed to perform sections 4.3.and 4.4 of procedure T-252-2, "RPV Venting During Containment Flooding" to setup the HPCI and RCIC Steam Line Drain flow paths. The task includes going to the EOP Tool Locker and obtaining the T-252 Tool Kit, and then going to the Cable Spreading Room to remove several fuses and verifying several HPCI and RCIC system valve positons with the main control room.

Loss of RBCCW -

RCA.

Plant Actions for the Instrument Nitrogen The candidate will be directed to perform steps 2.10 System and 2.11 of procedure ON-113, "Loss of RBCCW" on Unit 2. The task includes opening several air-operated instrument air backup to instrument nitrogen system valves and shutting down the Instrument Nitrogen Compressors.

2015 Peach Bottom NRC Exam JPM Summary Page 7

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 03/23/2015 Exam Level: RO r8] SRO-I D SRO-U D Operating Test Number: 2015 NRC 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*

Safety Function

a. 295001 AA1.06 (3.3/3.4) - Reactor Operator Actions on a Recirculation A, N,S 1

Pump Trip (Alternate Path - Thermal Hydraulic Instability Exists Without Operable OPRM System)(PLOR-374CA)

b. 206000 A3.07 (3.9/3.8) - Startup HPCI in the CST to CST Mode A, D, EN, 2

(Alternate Path - Turbine Exhaust Diaphragm High Pressure) (PLOR-353CA) 2011 NRC Exam P,S

c. 203000 A4.02 (4.1/4.1) Manual Startup of LPCI for Injection (Alternate A,EN,N,S 4

Path - RHR Injection Valve Trips on Thermal Overload)

d. 223001 A2.07 (4.2/4.3) - Drywell Venting via the 2 Inch Vents (Alternate A,D,S 5

Path - Main Stack Hiqh Radiation) (PLOR-321CA)

e. 264000 A4.04 (3.7/3.7) - Load Diesel Generator to 500kW (Alternate A, N,S 6

Path - Differential/Ground Fault) (PLOR-373CA)

f. 201006 A3.01 (3.2/3.1) - Initialize the Rod Worth Minimizer (PLOR-D, L, S 7

366C)

g. 400000 A4.01 (3.1/3.0) - ECW System Makeup to Emergency Cooling D,EN, P,S 8

Tower Usinq ESW System (PLOR-270C) 2011 NRC Exam

h. 295017AA1.09 (3.6/3.8) - Manually Place Standby Gas Treatment D, EN,S 9

System on Equipment Cell Exhaust (PLOR-18C)

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. 295031 EA1.08 (3.8/3.9) -Alternate RPV Injection Using the Standby D,E, R 2

Liquid Control Test Tank (PLOR-105P)

j. 206000 K1.01 (3.8/3.8) - RPV Venting During Containment Flooding D, E, L, R 4

(PLOR-91 P)

k. 295018AA1.01 (3.3/3.4) - Loss of RBCCW (Plant Actions for the D, P, R 8

Instrument Nitroqen System) (PLOR-96P) 2011 NRC Exam All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; a/15 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank 5_9/5_8/5_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

11:::11:::1 (N)ew or (M)odified from bank including 1 (A)
21:::21:::1 (P)revious 2 exams
5. 3 I 5. 3 I 5. 2 (randomly selected)

(R)CA

11:::11:::1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 03/23/2015 Exam Level: RO D SRO-I IZI SRO-U D Operating Test Number: 2015 NRC Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I J PM Title Type Code*

Safety Function

a. 295001 AA1.06 (3.3/3.4) - Reactor Operator Actions on a Recirculation A,N,S 1

Pump Trip (Alternate Path - Thermal Hydraulic Instability Exists Without Operable OPRM System)(PLOR-374CA)

b. 206000 A3.07 (3.9/3.8) - Startup HPCI in the CST to CST Mode A, D, EN, 2

(Alternate Path - Turbine Exhaust Diaphragm High Pressure) (PLOR-P,S 353CA) 2011 NRC Exam

c. 203000 A4.02 (4.1/4.1) Manual Startup of LPCI for Injection (Alternate A, EN,N,S 4

Path - RHR Injection Valve Trips on Thermal Overload)

d. 223001 A2.07 (4.2/4.3) - Drywell Venting via the 2 Inch Vents (Alternate A,D,S 5

Path - Main Stack Hiqh Radiation) (PLOR-321CA)

e. 264000 A4.04 (3.7/3.7) - Load Diesel Generator to 500kW (Alternate A, N,S 6

Path - Differential/Ground Fault) (PLOR-373CA)

f.
g. 400000 A4.01 (3.1/3.0) - ECW System Makeup to Emergency Cooling D,EN, P,S 8

Tower Usinq ESW System (PLOR-270C) 2011 NRC Exam

h. 295017AA1.09 (3.6/3.8) - Manually Place Standby Gas Treatment D,EN,S 9

System on Equipment Cell Exhaust (PLOR-18C)

In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. 295031 EA 1.08 (3.8/3.9) - Alternate RPV Injection Using the Standby D, E, R 2

Liquid Control Test Tank (PLOR-105P)

j. 206000 K1.01 (3.8/3.8) - RPV Venting During Containment Flooding D, E, L, R 4

(PLOR-91 P)

k. 295018AA1.01 (3.3/3.4) - Loss of RBCCW (Plant Actions for the D,P, R 8

Instrument Nitroqen System) (PLOR-96P) 2011 NRC Exam

@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 Criteria for RO I SRO-I I SRO-U (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 I -

I

<:: 1 (control room system)

(L)ow-Power I Shutdown

~1/~1/~1 (N)ew or (M)odified from bank including 1 (A)

~2/~2/~1 (P)revious 2 exams

.:=: 3 I.:=: 3 I.:=: 2 (randomly selected)

(R)CA

~1/~1/~1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 03/23/2015 Exam Level: RO D SRO-I D SRO-U rgj Operating Test Number: 2015 NRC 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*

Safety Function

a. 295001 AA1.06 (3.3/3.4) - Reactor Operator Actions on a Recirculation A,N,S 1

Pump Trip (Alternate Path - Thermal Hydraulic Instability Exists Without Operable OPRM Svstem)(PLOR-374CA)

b. 206000 A3.07 (3.9/3.8) - Startup HPCI in the CST to CST Mode A, D, EN, 2

(Alternate Path - Turbine Exhaust Diaphragm High Pressure) (PLOR-353CA) 2011 NRC Exam P,S

c.
d.
e.
f.
g.
h. 295017 AA1.09 (3.6/3.8) - Manually Place Standby Gas Treatment D, EN, S 9

System on Equipment Cell Exhaust (PLOR-18C)

In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i.
j. 206000 K1.01 (3.8/3.8) - RPV Venting During Containment Flooding D, E, L, R 5

(PLOR-91 P)

k. 295018 AA1.01 (3.3/3.4) - Loss of RBCCW (Plant Actions for the D, P, R 8

Instrument NitroQen System) (PLOR-96P) 2011 NRC Exam 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 Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank 5.915.815.4 (E)mergency or abnormal in-plant
.11:::.11:::.1 (EN)gineered safety feature I -

I ;:: 1 (control room system)

(l)ow-Power I Shutdown

.11:::.11:::.1 (N)ew or (M)odified from bank including 1 (A)
.21:::.21:::.1 (P)revious 2 exams
5. 3 I 5. 3 I 5. 2 (randomly selected)

(R)CA

.11:::.11:::.1 (S)imulator ES-301, Page 23 of 27

Appendix D Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #1 Op Test No.

2015 NRC Examiners Scenario Summary Operator _______ CRS (SRO)

URO (ATC)

PRO (BOP)

The scenario begins with the reactor at approximately 5% power during a reactor startup.

Following shift turnover, the PRO will raise Reactor Pressure Set to 915 psig using GP-2-2, "Normal Plant Startup". Then the URO will continue the startup by raising reactor power by withdrawing control rods in accordance with the approved startup sequence until 3 main turbine bypass valves are open with reactor pressure at 915 psig using procedure GP-2-2, "Normal Plant Startup". During this evolution a control rod will become mispositioned due to a Reactor Manual Control System timer failure, requiring the crew to execute ON-122, "Mispositioned Control Rod" to return the control rod to the correct target position.

Following Control Rod withdrawal, the steam supply valve for in-service Steam Jet Air Ejector fails closed due to a loss of its normal air supply. The loss of steam to the air ejector will cause main condenser vacuum to get worse. The crew should recognize the lowering vacuum condition and enter procedure OT-106 "Condenser Low Vacuum". The CRS should direct the crew to place the steam supply valve alternate air supply in service and restore the air ejector to normal service and thereby reestablishing normal main condenser vacuum.

Once main condenser vacuum is normal there will be a loss of a 125 VDC bus, which will result in loss of multiple ECCS, RCIC, the E-1 Emergency Diesel Generator, and multiple main control room annunciators. The crew should recognize the condition and enter procedure SE-13 "Loss of a 125 or 250 VDC Safety Related Bus". The CRS should direct the crew to perform actions to minimize plant impact such as cross-tie electrical feeds on the 13kV auxiliary busses, place the control switch for the E-1 Emergency Diesel Generator in pull-to-lock, and functionally test all main control room annunciators. The crew should also enter procedure ON-123 "Loss of Control Room Annunciators". The CRS should reference Technical Specifications for required actions with multiple ECCS inoperable.

A Primary Containment Isolation valve will fail open. The failed valve will require the CRS to evaluate the situation in Tech Specs and determine that the penetration must be isolated within four hours.

When the CRS has determined the Tech Spec action, the startup level control system will experience a control signal failure resulting in the startup level control valve failing closed. The valve closure will halt any makeup to the RPV and subsequently RPV level will lower. The crew should recognize the lowering RPV level and enter procedure OT-100 "Reactor Level Low". Placing the startup level control valve controller into manual will not return control of the makeup valve. The URO will need to establish RPV level control using the "C" RFP discharge valve 2015 NRC Scenario #1 0-1 Rev 0

Appendix D Scenario Outline ES-D-1 and RFP speed.

Once RPV level is stabilized, a steam leak will develop in the primary containment with a stuck open Torus to Drywell vacuum breaker. The crew should recognize the rise in drywell temperature and pressure and enter procedure OT-101 "High Drywell Pressure". OT-101 actions include maximizing drywell cooling and isolating steam supply valves in the drywell in order to identify the possible leak location.

When drywell pressure reaches 1.2 psig the crew should attempt to scram the reactor. When the mode switch is placed in shutdown no control rods will insert due to an electric ATWS. The crew should enter procedure T-101 "RPV Control" to respond to the ATWS condition. The control rods will fully insert and the ATWS will be terminated when Alternate Rod Insertion is initiated using Rapid Response Card RRC 38.1-2 "ARI During a Plant Event". (Critical Task; Insert all control rods using ARI)

When drywell pressure reaches 2 psig the crew will enter procedure T-102 "Primary Containment Control" to respond to the degrading condition. The crew should spray the primary containment using procedure T-204 "Initiation of Containment Sprays Using RHR" to maintain below the Pressure Suppression Pressure Limit. (Critical Task; Spray the Drywell before the Pressure Suppression Pressure Limit Curve is exceeded) When Torus Sprays are placed in-service, the RHR pump will trip and another RHR will need to be placed in-service. The scenario will be terminated when Primary Containment pressure is stable due to spraying containment.

Initial IC-71 Approximately 5% power Conditions Turnover Unit 2 startup is in progress.

Drywell purge has just been secured. The extra RO will begin inerting Containment shortly after turnover.

Reactor Power is approximately 5% with direction to continue to raise Reactor power with control rods.

Reactor pressure is 450 psig with direction to raise pressure set to 915 psig.

2015 NRC Scenario #1 D-1 Rev 0

Appendix D Scenario Outline ES-D-1 Event Malfunction Event Event No.

No.

Type*

Description 1

See Scenario Guide N

PRO Raise Pressure set to 915 psig CRS 2

See Scenario Guide R

URO Raise reactor power by withdrawing control rods until 3 main CRS turbine bypass valves are open with reactor pressure at 940 psig 3

See Scenario Guide c

URO A control rod becomes mispositioned, requiring execution of CRS ON-122 "Mispositioned Control Rod" 4

See Scenario Guide c

PRO Steam supply valve for in-service Steam Jet Air Ejector fails CRS closed I lowering main condenser vacuum 5

See Scenario Guide c

PRO Loss of 125 VDC bus I multiple ECCS inoperable (Tech TS CRS Spec) 6 See Scenario Guide TS CRS Failure of a Primary Containment isolation valve 7

See Scenario Guide c

URO Startup level control valve fails closed I lowering RPV level CRS 8

See Scenario Guide M

ALL Reactor coolant leak inside the drywell I Torus to Drywell vacuum breaker fails open 9

See Scenario Guide c

URO A TWS I Control rods inserted using Alternate Rod Insertion CRS 10 See Scenario Guide c

PRO RHR pump running in Torus Spray trips CRS

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec 2015 NRC Scenario #1 D-1 Rev O

Appendix D Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #2 Op Test No.

2015 NRC Examiners Operator ______ _

CRS (SRO)

URO (ATC)

______ PRO (BOP)

Scenario The scenario begins with the reactor at 50% power during a plant startup.

Summary Following shift turnover the Crew will place the Reactor Core Isolation Cooling (RCIC) system in a test flow path at rated flow for data collection using procedure SO 13.1. B-2, "RCIC System Manual Operation". While the RCIC system is in service the Crew will continue the startup by raising reactor power by withdrawing control rods in accordance with the approved startup sequence using procedure GP-2-2, "Normal Plant Startup". During this evolution a control rod will become stuck due to mechanical binding, requiring the Crew to adjust control rod drive pressure by executing procedure SO 62.1.A-2, "Withdrawing/Inserting a Control Rod" in order to place the control rod to the correct target position.

Once the stuck control rod is at the correct target position, the 'A' station air compressor will trip on low oil pressure. The Crew should recognize the trip of the compressor and enter procedure ON-119, "Loss of Instrument Air". The Crew should start the backup air compressor and direct Equipment Operators to manipulate local valves to restore the 'A' Instrument air header to normal pressure.

Once the 'A" Instrument air system is returned to normal, the already in-service 'B' RHR Pump will experience an overcurrent condition. The Crew should recognize the higher than normal pump motor amperage and remove the 'B' RHR pump from service and the CRS should direct that Torus Cooling be secured using procedure SO 10.1.D-2, "RHR System Torus Cooling". The CRS should reference Technical Specifications for the inoperable RHR Pump.

Following the Technical Specification determination, the Motor Driven fire pump will become INOP. When the Crew determines that the MDFP is INPO the CRS will consult the TRM to determine appropriate actions.

Following the TRM determination, a primary system steam leak will occur downstream of the inboard Main Steam Isolation Valve (MSIV), allowing steam to enter the Reactor Building. The Crew should respond to high area temperature alarms and respond to the condition using procedure T-103 "Secondary Containment Control". The CRS should initiate evacuation of the Reactor Building and direct a manual scram of the reactor using procedure GP-4 "Scram". During the GP-4 shutdown, the 'A" Reactor Recirculation Pump (RRP) will fail to change speed. The Crew will be required to trip the 'A" RRP.

A CRD hydraulic malfunction will result in a hydraulic A TWS, requiring the Crew to execute T-101 "RPV Control" and T-117 "Level/Power Control." (Critical Task; Inhibit ADS before an automatic depressurization occurs.) The Crew should lower RPV level using T-240 "Termination and Prevention of Injection into the 2015 NRC Scenario #2 D-1 Rev 1

Appendix D Scenario Outline ES-D-1 RPV" and be inserting control rods by perform T-220 "Driving Control Rods During Failure to Scram" and T-216 "Control Rod Insertion by Manual Scram or Individual Scram Test Switches" to insert control rods. (Critical Task; Attempt to shutdown the Reactor by performing one or more of the following: T-216, "Control Rod Insertion by Manual Scram or Individual Scram Test Switched", T-220, "Driving Control Rods During a Failure to Scram",

Injecting Standby Liquid Control before Torus temperature exceeds 11 o°F}The MS IVs will fail to automatically isolate on a Group 1 isolation signal and will have to be manually closed by the Crew. (Critical Task; Close at least one MSIV in each steam line following receipt of a Group I isolation signal and failure of the MSIVs to isolate.}

The scenario may be terminated when the Crew has control of RPV pressure and RPV level has been lowered and is greater than -195 inches using T-240 "Termination and Prevention of Injection into the RPV". (Critical Task; Maintain core cooling by restoring and/or maintaining RPV level above -195 inches but below the level that will cause a sustained Reactor power rise above 25% power.}

Initial IC-13, Approximately 50% Reactor power Conditions Turnover Reactor power is approximately 50%. Continue the Reactor startup by withdrawing Control Rods.

The "B" Loop of RHR is running in Torus cooling to support the RCIC operation.

The PRO is directed to place RCIC in the CST to CST mode for data collection.

2015 NRC Scenario #2 D-1 Rev 1

Appendix D Scenario Outline ES-D-1 Event Malfunction Event Event No.

No.

Type*

Description 1

See Scenario Guide N

PRO Place RCIC in test flow mode of operation CRS 2

See Scenario Guide R

URO Raise reactor power by withdrawing control rods CRS 3

See Scenario Guide c

URO A control rod becomes stuck, requiring execution of SO CRS 62.1.A-2 "Withdrawing/Inserting a Control Rod" 4

See Scenario Guide c

PRO

'A' Instrument Air Compressor trips I place backup air CRS compressor in service 5

See Scenario Guide c

PRO

'B' RHR Pump overcurrent condition (Tech Spec)

TS CRS 6

See Scenario Guide TS CRS Motor Driven Fire Pump INOP 7

See Scenario Guide c

ALL Primary system steam leak into the Reactor Building 8

See Scenario Guide c

URO

'A" Reactor Recirculation Pump fails to operate during CRS manual scram 9

See Scenario Guide M

ALL ATWS (hydraulic) 10 See Scenario Guide c

PRO Main Steam Isolation Valves fail to automatically isolate I Close manually

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec 2015 NRC Scenario #2 D-1 Rev 1

Appendix D Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #4 (Spare)

Op Test No.

2015 NRC Examiners Scenario Summary Operator ------- CRS (SRO)

URO (ATC)

PRO (BOP)

The scenario begins with the reactor at 100% power with the 'B' Emergency Service Water (ESW) Pump in service for an evaluation of flow through the Emergency Diesel Generator heat exchangers.

Shortly after taking the shift the Crew will swap Electrohydraulic Control (EHC)

Pumps using procedure SO 1 D.6.A-2 "Placing the EHC Oil System Standby Pump in Service". The 'B' EHC pump will be started ad the 'A' EHC pump will be shut down.

Once the "B" EHC pump is in service, an Equipment Operator will report a Core Spray snubber is INOP. The CRS will review the TRM and Tech Spec and determine that the Core Spray loop is INOP.

After the Tech Spec determination is made, the in-service Steam Packing Exhauster (SPE) blower will trip. The Crew should recognize the trip condition and restart the tripped blower or start the standby blower using alarm response procedures.

After the SPE blower is placed in service a leak will develop on the discharge of the running 'B' ESW Pump requiring the Crew to recognize the condition and secure the 'B' ESW pump. The CRS should reference Technical Specifications for the inoperable ESW pump and also for inoperable fire barriers due to doors being intentionally left open in response to the flooding.

Once the Technical Specification determinations have been made, the running RBCCW pump will trip and the standby pump will fail to start, resulting in a complete loss of RBCCW. The Crew should reduce RBCCW loads (e.g. RWCU) and reduce reactor power as directed by ON-113 "Loss of RBCCW." The Crew should reduce power using procedure GP-9 "Fast Power Reduction". As a result of the loss of RBCCW the 'B' Recirculation Pump will experience a mechanical seal failure which is the source of a steam leak into the primary containment. The Crew should enter procedure OT-101 "Drywell High Pressure". Temperatures on the recirculation pump will rise requiring the Crew to remove the pump from service and they should enter procedure OT-112 "Unexpected/Unexplained Change in Core Flow". When primary containment pressure reaches 1.2 psig the Crew will shut down the reactor using procedure GP-4 "Scram". When the Crew places the mode switch in shut down the control rods will not insert due to a failure of the reactor mode switch. Depressing the manual scram pushbuttons will insert the control rods. (Critical Task; Shutdown the Reactor by placing the Reactor Mode Switch in "SHUTDOWN" prior to Drywell pressure exceeding the 2 psig scram setpoint.)

As the steam leak progresses the Crew should execute proceduresT-101 "RPV 2015 NRC Scenario #4 D-1 Rev 0

Appendix D Scenario Outline ES-D-1 Control" and T-102 "Primary Containment Control". The Crew should spray the primary containment using procedure T-204 "Initiation of Containment Sprays Using RHR". (Critical Task; Spray the Drywell in accordance with T-204, "Initiation of Containment Sprays using RHR", when conditions permit, but before Drywell Temperature exceeds 281°F.) A Drywell Chilled Water system to RBCCW system leak will develop allowing steam to leak into the RBCCW Room outside of the primary containment. The Crew will need to isolate the RBCCW system using procedure GP-8.B "PCIS Isolation - Groups 2 and 3".

(Critical Task; Isolate RBCCW from the Drywell in the Control Room.)

The scenario may be terminated when the reactor is shut down with RPV level is under control, Primary Containment sprays are in service, and the RBCCW leak is isolated.

Initial IC-14, 100% power Conditions Turnover Unit 2 is at 100% power.

There is a leak in the RBCCW system that requires the head tank to be filled every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The head tank was last filled 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ago.

The "B" ESW pump is in-service to do a flow evaluation of flow through the DIG heat exchangers. The test is expected to be completed within the hour.

Following turnover the PRO will be directed to place the "B" EHC pump in-service and secure the "A" EHC pump.

2015 NRC Scenario #4 D-1 Rev O

Appendix D Scenario Outline ES-D-1 Event Malfunction Event Event No.

No.

Type*

Description 1

See Scenario Guide N

PRO Swap EHC Pumps CRS 2

See Scenario Guide TS CRS INOP Core Spray pump discharge snubber 3

See Scenario Guide c

PRO Steam Packing Exhauster (SPE) fan trip I (re)start SPE fan CRS 4

See Scenario Guide c

PRO

'B' ESW Room flood I secure the 'B' ESW Pump (Tech Spec)

TS CRS 5

See Scenario Guide R

URO Loss of RBCCW I fast reactor power reduction (w/ recirc and CRS rods) 6 See Scenario Guide c

URO

'B' Recirculation Pump seal failure I Steam leak in primary CRS containment 7

See Scenario Guide I

URO Failure to automatically scram (manual scram pushbuttons CRS are required to scram the reactor) 8 See Scenario Guide M

ALL Drywell to RBCCW leak I Steam leak in RBCCW Room outside of primary containment

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec 2015 NRC Scenario #4 D-1 Rev 0
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