ML15169A646

Final Outlines (Folder 3)

ML15169A646
Person / Time
Site:	Peach Bottom
Issue date:	03/13/2015
From:	Exelon Generation Co
To:	Peter Presby Operations Branch I
Shared Package
ML14254A311	List: IR 05000277/2015301 ML14302A020 ML15077A098 ML15103A132 ML15131A255 ML15131A256 ML15169A646 ML15169A648 ML15169A649 ML15169A651 ML15169A652 ML15169A654 ML15169A656 ML15174A053
References
TAC U01911
Download: ML15169A646 (41)

Text

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

!LT 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

I 1

2 I

1 1

7 1

2 3

Plant Tier Evolutions Totals 4

5 5

5 4

4 27 3

7 10 1

2 3

2 2

2 3

2 2

2 26 3

2 5

.)

.)

2.

Plant 2

1 1

l 0

3 I

1 I

I I

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 I

2 3

4 1

2 3

4 Categories 10 7

3 3

2 2

2 1

2 2

Note:

I.

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 K/ !\\ 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. l.b of ES-401, for guidance regarding elimination of inappropriate K/ 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 1 and 2 from the shaded systems and KIA categories.

7.*

The generic (G) Kl As 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 K/A's

8.

On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (JR) 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, !Rs, and point totals(#) on Form ES-401-3. Limit SRO selections to Kl As that are linked to 10CFR55.43

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 EAPE #I Name Safety Function K1 K2 K3 A1 A2 G

KIA Topic(s) 2950 I h Cintrol Room 2.12_; Ability to perform >pccitic system 1\\handonmcnt ! 7 x

and inkgratcd plant proccdun:s during all modes uf plant opcraticHL AA1.0:1 - Ability to <lctwninc and/or 2<l5021J Inadvertent C'onlainmcn1 x

interpret the following ~b they apply to lsolation 5 INi\\DVLRI I.NT CONIAINMl.:r-;T ISOLATION :Rcach>r l'rt:ssnrc LA2.04 - Ability to determine and/or 295031 Reactor l.m1 Water l.c1d x

imerprct the following as they apply to

/ 2 REACTOR LO\\V WATER LLVLL:

Adequate core cooling 2950.30 Low Suppression Pool x

2.2.22 Knowledge of limiting conditins Water Level I 5 for operations and safety limits.

295019 Partial 1lr Tntal Loss of x

1.1.20 - Conduct nfOpcrali1lns: Ability to lnst. Air I 8 interpret and execute procedure steps.

2l))Q(J3 Partial tir Complete l Jiss 2.1.2 - Conduct of Operations: Knmliedg.c of 1\\C (1

x of opcrntcir rcsp\\\\nsit>ilitics <luring all modes of plam opcratiun.

29~0 IX Partial or Total l.uss o I' 1.1.19 - Conduct,1f Opaali\\\\11s i\\hilil~ to CC \\V ! 8 x

use plant computers to naluate system (lr cnmponent status.

AKl.02 - Knowledge of the operational 29500 I Partial or Complete Loss implications of the following concepts as of Forced Core Flow Circulation I x

they apply to PARTIAL OR COMPLETE 1&4 LOSS OF FORCED CORE FLOW CIRCULATION: Power/flow distribution AKI.OJ - 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 implications of the following concepts as 12 x

they apply to REACTOR LOW WATER LEVEL: Natural circulation: Plant-Specific 295037 SCRAM Conditions EK2.09 - Knowledge of the interrelations Present and Reactor Power Above between SCRAM CONDITION PRESENT APRM Downscale or Unknown I x

AND REACTOR POWER ABOVE l

APRM DOWNSCALE OR UNKNOWN and the following: Reactor water level AK2.02 - Knowledge of the 295005 Main Turbine Generator x

interrelations between MAIN TURBINE Trip/ 3 GENERATOR TRIP and the following:

Feedwater temperature EK2.09 - Knowledge of the interrelations between HIGH 295024 High Drywell Pressure I 5 x

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 of AC 16 to 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 x

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

Q#

4.4 76 3.9 77 4.8 78

• U 79 4.6 80 4.4 81 J.8 X2 3.3 I

3.5 40 3.8 41 4.0 75 2.9 43 2.9 44 3.7 45 3.5 46

ES-401 2

Form ES-401-1 IL T 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 Inst. Air I 8 x

to PARTIAL OR COMPLETE LOSS OF 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. 15 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 x

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

interpret the following as they apply to Temperature I 5 HIGH DRYWELL TEMPERATURE:

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

interpret the following as they apply to Rate/ 9 HIGH OFF-SITE RELEASE RATE:

Source of off-site release 295021 Loss of Shutdown x

2.1.20 - Ability to interpret and execute Cooling/ 4 procedure steps.

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.

AA1.01 -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 295016 Control Room x

interrelations between CONTROL Abandonment I 7 ROOM ABANDONMENT and the following: Control room HVAC K/A Category Totals:

3 4

3 4

3/2 3/5 Group Point Total:

Imp.

3.5 47 4.1 48 3.8 49 3.1 50 3.2 51 4.1 52 4.1 53 4.6 54 3.8 55 3.9 56 33 57 2.9 58 I 20/7

ES-401 3

IL T 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#

FA2.0l - Ability lo determine and/or 295035 Sc.:onJ~r) Containment intcrprd lhc follm\\ ing as they appl) tP x

Sl:CONDAJ~Y CONTAINMENT JHGH 3.lJ 83 I I igh 1.Jifkrcntial Pressure!'.'>

DllTI RFNllAL PRESSURIC: Si:condary containment pressure: Plant-Specific 29:'017 I ligh Off-,itc Rekas.: Rate x

2.2.12 - Lquipmcnl Control: Knowledge of 4.1 84

/l) surveillance rr,iccdures.

2.1.23 ****Ability lo perform spccitk system 2'J'll 1 :i lnuimpk-tc' Sl "RA\\!1 ! I x

and inkgratcd plant procedures during all

,,.4 85 modes or rlant opcrutiun.

EKl.03 - Knowledge of the operational implications of the following concepts as 295032 High Secondary x

they apply to HIGH SECONDARY 3.5 59 Containment Arca Temperature I 5 CONTAINMENT AREA TEMPERATURE: Secondary containment leakage detection AK2.0l - Knowledge of the interrelations 295015 Incomplete SCRAM I I x

between INCOMPLETE SCRAM and the 3.8 60 following: CRD hydraulics 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 drvwell 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 AA2.0 I - 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 2.4.31 - Emergency Procedures I Plan:

x Knowledge of annunciator alarms, 4.2 64 Water Level I 5 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 K/A Category Totals:

1 1

2 1

1 /1 1/2 Group Point Total:

I 7/3

ES-401 System # I Name K

K K

1 2

3 215004 Source Range J'vlonitL1r 2(>2001 l ll'S 1 :\\CiUC) 2640011 l.DUs 241000 Reactor/Turbine Pressure Rcgul<iting System 259002 Reactor Water l.cvcl Control 259002 Reactor Water Level x

Control 209001 LPCS x

203000 RHR/LPCl: 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

\\2.01 -Ability to (a) predict the impacts of the following on the SOURCL Ri\\1\\CiL MONITOR (SRl'vlJ SYSTEM

and (b) based on x

those predictions, use procedures to wrrcc1. control, or mitigak th.c consequences of those almorrnal conditions or operations: PO\\wr supply degraded

\\2 0 l - Ability 1,1 (a) predict the impacts uflhc following on the l ININ I El<Rl

Pl!\\BU.: PO\\\\TR SlJPPLY (A.C./D.C.): and (b) x based on those predictinns. use procedures 10 correcl. contrnl. or mitigate the consequences or those abnormal conditions L>r operations:

Under vollage x

2.4. l 1-Knowlcdge of ahnormal rnndition procedures.

2.1.7 -Ability tu cvaluak plant performance and make Llperational; x

judgmcnts based on operating characteristics, reactor behavior, and instrument interpretation.

/\\2.02 - Ability to (a) predict the impacts of the following on the Rf ACTOR W:\\TFR UVFI.

CONTROL SYSTFVI; and (b) x based <lll thuse predictions. use procedures to correct, c1>ntrol, \\lf mitigate the ccn1scquenccs,,f those abnormal conditi11ns or orcruticllls:

Luss 01* any numhcr or rcact<>r tccdwakr lluv, inputs Kl.05 - 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 svstem 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.'!

86 2.8 8'j,

4.2 88 4.7 S9 3.4 90 3.6 39 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 IRM 211000 SLC 209001 LPCS 4

Form ES-401-1 IL T 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

(DIESEUJET) design feature(s) and/or interlocks which provide for the following: Automatic startup K4.01 - Knowledge of AUTOMATIC DEPRESSURIZATION 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 K5.02 - Knowledge of the operational implications of the x

following concepts as they apply to A.C. ELECTRICAL DISTRIBUTION: Breaker Control 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. oower A1.04 -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 IO 3.8 II 3.2 12 3.7 13

ES-401 System # I Name K

K K

1 2

3 218000 ADS 206000 IIPCI 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: A.C. 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 SGTS 217000 RCIC x

KIA 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

K5.03 - 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 followin~:i: RPS A2.06 - 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: Valve Closure 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 2.9 25 3.6 26 I

26/5

ES-401 System #I Name K

K K

1 2

3 21 ~001 TrnVcTsing In-core l'ruhc 2t12ti02 l\\ccirculdliun Flm\\

Clllltrul 23300() h:c! [)01Ji C,H1l 111g1t:Jca11up 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 - ;\\bility lo (D) predict the impocls of the following on the TRA VLRSING IN-CORI PROB!

• and ( b) based on those x

predictions. use procedures to correct, conlrol, nr mitigate lhc cons<:quences or th,isc abnormal ct>11di1ions "r operation-.: Failure w rdr:rcl to siliclu: (\\JPt-B\\VR J J x

22'10 1\\hilitv l<' apph Technical Spccilicat1mi> for a s~stcrn 2.2.22 - l:quip111c1ll ( ontrnl.

x Knowkdgc or limiting Clllldilions for <lfll:ratiuns and salcty 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 K2.0I - 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 K5.01 - Knowledge of the operational implications of the x

following concepts as they apply to SECONDARY CONTAINMENT: Vacuum breaker operation K5.01 - 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 during system operation Imp.

Q#

2.9

')I 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 259001 Reactor Feed water 202001 Recirculation K/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

FUELPOOLCOOLINGAND CLEAN-UP including: System indicatinq liqhts and alarms A4.02 - Ability to manually x

operate and/or monitor in the control room: Emergency in/notch override switch 2.1.31 -Ability to locate control room switches, controls, and x

indications, and to determine that they correctly reflect the desired plant lineup.

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 4.6 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 RO SRO-Only KIA#

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 Knowledge of how to conduct system 2.1.29 lineups, such as valves, breakers, switches, 4.1 67 etc.

2.1.3 Knowledge of shift or short-term relief 3.7 42 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.

2.2.22 Knowledge of limiting conditions for 4.0 68

2.

operations and safety limits.

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

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 reactivity.

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

3.

responsibilities, access to locked high-radiation areas, aligning filters, etc.

Radiation Control Ability to use radiation monitoring systems, 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 emergency conditions.

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

Form ES-401-3 I Subtotal 2

2

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

Form ES-401-3 Facility:

ILT 13-1 2015 NRC Exam Date:

03/23/15 RO SRO-Only Category KIA#

Topic Q#

Q#

IR IR 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 3.8 73 usage.

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 KIA RO 211 262001K5.01 This Kl A overlapped with the operating test. Replaced with Q # 10 KIA 262001 K5.02 RO 2 I 1 300000 I 2.4.41 EAL classification task for an RO question. Replaced with Q #21 KIA 300000 2.4.31.

RO 211 Could not a satisfactory question to the original K/ A of Q # 25 261000 I A2.04 impact of high moisture content and procedure to correct or mitigate. Replaced with KIA 261000 A2.06.

RO 2/2 290001 I K4.0l 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.

R02/2 Could not a satisfactory question linking the process Q #37 259001 2.1.19 computer to the feed water system. Replaced with Kl A 259001 2.1.31 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 2.1.20.

RO 1 /2 Could not write a satisfactory question linking Radiation Q #59 295032 EKl.02 release to High Secondary Containment area temperatures.

Replaced with KIA 295032 EKl.03.

RO 1I2 295034 I AAl.05 Too many similar KIA for Secondary Containment selected.

Q#62 Replaced with KIA 295002 AAl.05 R03 2.1.27 Could not a tier 3 question matching the system purpose.

Q #67 Replaced with KIA 2.1.29 R03 2.2.7 Not an RO task. Replaced with Kl A 2.2.22 Q #68 R03 2.2.20 Not an RO task. Replaced with Kl A 2.2.1 Q#69 SRO 1I1 295016 I AA2.05 Could not make a SRO question linking the Kl A. Replaced Q # 76 with KIA 295016 2.1.23.

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 2.2.22.

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.

ES-401 Record of Rejected K/A's Form ES-401-4 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 1:8:1 SRO D 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-241C) (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 ~

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 M,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

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 03/23/2015 Exam Level: RO ~ SRO-I 0 SRO-U 0 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 AA 1.06 (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) (PLOR-381CA)

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

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

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 Using 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 (3.8/3.9) - RPV Venting During Containment Flooding D, E, L, R 4

(PLOR-93P)

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

Instrument Nitrogen 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)

_:::2/_:::2/_:::1 (P)revious 2 exams 5_ 3 I 5_ 3 I 5_ 2 (randomly selected)

(R)CA

1 I ;

1 I _::: 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 0 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 AA 1.06 (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)(PLOR-381 CA)

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 Using ESW System (PLOR-270C) 2011 NRC Exam

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. 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 (3.8/3.9) - RPV Venting During Containment Flooding D, E, L, R 4

(PLOR-93P)

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

Instrument Nitrogen 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 [8J 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 AA 1.06 (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. 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.

j. 206000 K1.01 (3.8/3.9) - RPV Venting During Containment Flooding D, E, L, R 4

(PLOR-93P)

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

~9/~8/~4 (E)mergency or abnormal in-plant

?.1/?.11?.1 (EN)gineered safety feature

- I -

I

~ 1 (control room system)

(L)ow-Power I Shutdown

?.11?.1/?.1 (N)ew or (M)odified from bank including 1 (A)

?.21?.21?.1 (P)revious 2 exams

~ 3 I ~ 3 I ~ 2 (randomly selected)

(R)CA

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

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 2.4.4 l EAL classification task for an RO question. Replaced with Q #21 Kl A 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 K5.0l.

RO 2/2 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 AKI.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 2.1.20.

RO 1I2 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 2.1.23.

SRO 1I1 295004 I AA2.04 Sarne 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 2.2.22.

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.

SRO 3 2.3.6 Could not develop a SRO question. Replaced with 2.3.13.

Q #98

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 3% power during a reactor startup.

Following shift turnover, the PRO will secure the Drywell Purge lineup.

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.

The URO will continue the startup by raising reactor power to > 4% 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".

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 the mispositioned Control Rod, the steam supply valve for the 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. The CRS might direct the PRO to swap Air Ejectors using SO 8A.6.A-2, "Placing the Standby SJAE In Service and Placing the In-Service SJAE in Standby".

Once main condenser vacuum is normal there will be a spurious start of the "A" loop of Core Spray. Following the spurious start a leak will occur in the "A" Core Spray pump discharge piping. The leak will continue until the Crew secures the pumps and isolated the suction for the "A" Core Spray pump. The CRS should reference Technical Specifications for required actions with the "A" Core Spray pump inoperable.

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 and RFP speed.

2015 NRC Scenario #1 D-1 Rev 1

Appendix D Scenario Outline ES-D-1 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 A TWS condition. The control rods will fully insert and the A TWS 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 Drywell 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 3% power Conditions Turnover Unit 2 startup is in progress.

Drywell purge needs to be secured. The extra RO will begin inerting Containment shortly after turnover.

Reactor Power is approximately 3% with direction to continue to raise Reactor power with control rods using GP-2-2 2015 NRC Scenario #1 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 Secure the Drywell Purge Lineup CRS 2

See Scenario Guide TS CRS Failure of a Primary Containment isolation valve 3

See Scenario Guide R

URO Raise reactor power by withdrawing control rods until 2 main CRS turbine bypass valves are open with reactor pressure at 915 psig 4

See Scenario Guide c

URO A control rod becomes mispositioned, requiring execution of CRS ON-122 "Mispositioned Control Rod" 5

See Scenario Guide c

PRO Steam supply valve for in-service Steam Jet Air Ejector fails CRS closed I lowering main condenser vacuum 6

See Scenario Guide c

PRO "A" Core Spray loop spurious start. "A" Core Spray suction TS CRS line break/flooding (Tech Spec) 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 1

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

NRC Examiners Scenario Summary Operator _______ CRS (SRO)

______ URO (ATC)

______ PRO (BOP)

The scenario begins with the reactor at 100% power with the 'B' Electrohydraulic Control (EHC) pump and the E-332 breaker are blocked out of service for scheduled maintenance.

After taking the shift, the Crew will perform the Master Trip Solenoid Valve Routine Test RT-0-01 D-402-2.

Shortly after this, the E-4 diesel generator will inadvertently start, requiring the Crew to shut down the E-4 diesel generator from the main control room and apply Technical Specifications for an inoperable diesel generator.

Following the diesel generator inoperability there will be a loss of the 250 VDC bus that supplies power to the RCIC system. The Crew must recognize the loss of DC power and apply Technical Specifications for the inoperable DC power supply.

Following the 250 VDC bus inoperability, the in-service Turbine Building Closed Cooling Water (TBCCW) pump trips on overload and the standby TBCCW pump fails to automatically start. The Crew should place the standby pump in service by placing its control switch to start and monitor the system pressure and temperature.

Shortly after the TBCCW system is restored the Crew should recognize and respond to lowering main condenser vacuum caused by air in-leakage. The Crew will be able to stabilize the plant with the existing air in-leakage by entering OT-106 "Condenser Low Vacuum" and reducing reactor power in accordance with GP-9-2 "Fast Power Reduction". During the power reduction a reactor feed pump will not automatically respond requiring the Crew to take manual control and reduce feed pump flow to avoid a main turbine trip on high reactor water level.

Following the power reduction, a high vibration condition for the main turbine will occur, requiring the Crew to scram the reactor and trip the main turbine. A CRD hydraulic malfunction will result in an ATWS, requiring the Crew to execute T-101 "RPV Control" and T-117 "Level/Power Control."

A failure of the only available EHC pump will cause the turbine bypass valves to close, requiring the Crew to utilize SRVs for reactor pressure control. The Crew should 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 Switches", T-220 "Driving Control Rods During Failure to Scram", T-246, "Maximizing CRD Flow to the Reactor Vessel", "Initiating Standby Liquid Control before Torus temperature 2015 NRC Scenario #2 D-1 Rev 2

Appendix D Initial Conditions Turnover Scenario Outline ES-D-1 exceeds 110°F"). The scenario may be terminated when the Crew has control of RPV power and level using T-240 "Termination and Prevention of Injection into the RPV" and the Crew is inserting control rods. (Critical Task; Before violating the Heat Capacity Temperature Limit (HCTL) curve,, perform T-240 "Terminating and Preventing Injection Into the RPV" to protect Primary Containment until: Reactor power is below 4% or RPV level reaches -172 inches or All SRVs remain closed and Drywell pressure is below 2 psig.)

IC-14, 100% power Reactor power is 100% power.

'B' Electrohydraulic Control (EHC) pump blocked out of service for scheduled maintenance.

E-332 breaker is blocked out of service for scheduled maintenance.

Perform the Master Trip Solenoid Valve Routine Test RT-0-01 D-402-2.

2015 NRC Scenario #2 D-1 Rev 2

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

No.

Type*

Description 1

See Scenario Guide N

PRO Perform the master trip solenoid valve routine test CRS 2

See Scenario Guide I

PRO E4 diesel generator spurious start I diesel generator manual TS CRS shutdown (Tech Spec) 3 See Scenario Guide TS CRS Loss of 250 VDC bus I RCIC becomes unavailable (Tech Spec) 4 See Scenario Guide c

PRO In service Turbine Building Closed Cooling Water (TBCCW)

CRS pump trips on overload I Failure of standby TBCCW pump to automatically start 5

See Scenario Guide R

URO Main condenser air in-leakage causes lowering condenser vacuum I GP-9 fast power reduction (with Recirc) 6 See Scenario Guide c

URO Reactor feed pump does not respond to lowering power change I must place in manual to control 7

See Scenario Guide c

URO Main turbine high vibration I reactor scram CRS 8

See Scenario Guide M

ALL A TWS (hydraulic) 9 See Scenario Guide c

PRO Remaining EHC pump trips causing loss of main turbine bypass valves I control reactor pressure with HPCI and/or SRVs

• (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec 2015 NRC Scenario #2 D-1Rev2

Appendix D Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #4 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 and 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, a high temperature condition will occur on the in-service RWCU pump. With the standby pump out of service the Crew will be required to remove the RWCU system from service.

After the RWCU system is removed from 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 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 depressing the Manual Scram Pushbuttons.)

The steam leak worsens. The Crew should execute proceduresT-101 "RPV Control" and T-102 "Primary Containment Control". The Crew should spray the primary containment using procedure T-204 "Initiation of Containment Sprays 2015 NRC Scenario #4 D-1 Rev 2

Appendix D Scenario Outline ES-D-1 Using RHR". (Critical Task; Spray the Drywell to (restore and) maintain Drywell Bulk Average Temperature below 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.

"B" RWCU pump is out of service.

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 D/G 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 2

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 RWCU pump motor high winding temperature, secure CRS RWCU.

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 c

URO Failure of the Recirc suction valve "M0-2-02-0438" to close CRS 8

See Scenario Guide I

URO Failure to automatically scram (manual scram pushbuttons CRS are required to scram the reactor) 9 See Scenario Guide M

ALL Steam leak worsens 10 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 2

Peach Bottom 2015 NRC Initial License Exam Changes Made to JPMs Since Initial Submittal JPM Change(s)

CR-a This JPM was heavily modified to make the Examinee believe that the JPM task is to Reactor Operator lockup the Scoop Tube for the 2A Reactor Recirculation Pump M-G Set. Once the Actions on a Recirc JPM begins, then the Examinee receives alarm indication and cueing that is Pump Trip -Alt Path, expected to drive the Examinee to trip the 2A Reactor Recirculation Pump M-G Set.

THI Exists with Once tripped, Thermal Hydraulic Instability will occur driving the Examinee to scram Inoperable OPRM the reactor. Specific changes include:

System

1. Task Conditions/Prerequisites cue sheet was updated to have the Examinee prepare to lockup the Scoop Tube for the 2A Reactor Recirculation Pump M-G Set.

2.

Added new steps #2, #3, and #4 (ALL CRITICAL) related to the Examinee having to determine that the 2A RRP must be removed from service due to high motor bearing temperatures.

3.

Bolded cue in steps 9 and 13 since it is required for JPM to progress.

CR-b This JPM was heavily modified to remove all references to shutting down the HPCI Startup HPCI in CST to system using normal operating procedure SO 23.1-2 "HPCI System Shutdown" since CST Mode - Alt path, we are expecting the candidate to shutdown the system using either 1) the Alarm Turbine Exhaust Response Card for the abnormal condition (immediately trip the turbine) or 2) the Diaphragm High Rapid Response Card that is directed in the initiating cue. Specific changes include:

Pressure

1.

Removed normal operating procedure SO 23.1-2 "HPCI System Shutdown" from the list of references.

2.

Added two new Task Condition /Prerequisites to state that RPV level is normal and steady, and that primary containment pressure is also normal and steady. This is information that the Candidate will need to determine that there are NO HPCI initiation signals present.

3.

Deleted extra period in name for M0-2-23-24 in steps #2 and #3.

4.

Added HPCI system flow controller FIC-2-23-108 operation to step 12.

5.

Added new NOTE after step 13 to describe 2 possible procedural flow paths for securing HPCI.

6.

Steps 14 through 27 are now the steps out of RRC 23.1-2, Section F "HPCI Shutdown with NO Initiation Signal Present" that are expected to be used by the Candidate CR-c

1.

Bolded cue in steps 16 and 19 since it is required for JPM to progress.

Manual Startup of

2.

Added "and down slow" to RPV level trend in Task LPCI for Injection -Alt Conditions/Prerequisites.

Path, Injection Valve

3.

Corrected misaligned Initiating Cue on Cue Sheet Trips

4.

All motor operated valve number designators were updated with formal valve number (i.e. M0-25A is now M0-2-10-25A).

JPM CR-d Drywell Venting via the 2" Vent - Alt Path CR-e Load EDG to 500 KW -

Alt Path, Differential Ground CR-f Rod Worth Minimizer Initialization CR-g ECW System Makeup to Emergency Cooling Tower Using ESW System CR-h Manually Place SBGT on Equipment Cell Exhaust IP-i Alternate RPV Injection Using the SBLC Test Tank Change(s)

1.

Initial Condition setup to support a Drywell pressure of 1 psig.

2.

Modified Simulator Operator setup page to raise indication on both the A and B Main Stack Gas Recorders to reflect rising Main Stack high radiation levels during venting.

1.

Fixed numbering mistakes on page 3 of 9.

2.

Modified Simulator Operator setup page to bring in additional alarm for E-42 bus fault at the same time EOG fault alarm is received.

1.

Added /(Ensure that all RWM alarm messages are cleared on the RWM screen prior to JPM starting" to the Tools and Equipment section on page 3 of 6.

2.

Balded cue in steps 2, 10, and 18 since it is required for JPM to progress.

3.

Changed /(Initialization" to /(Initialize" in the cue and Standard for step #3.

4.

Changed /(FULLCORE" to all capitals in step #5.

1.

Added NOTE before steps 14 and 26 since these steps will require coordination with the Simulator Operator to open and close breakers locally.

2.

Balded cue in steps 3, 4, 5, 11, 13, 14, 21, 22, 24 and 26 since it is required for JPM to progress.

1.

Balded cue in steps 2 and 20 since it is required for JPM to progress

1.

Added a fifth Task Condition /Prerequisite to state that a loss of power event is NOT in progress

2.

Added the following Note at the beginning of the JPM:

a.

IF this is the first in-plant JPM then the license candidate is required to go to the EOP Tool Locker. For subsequent in-plant JPMs it is not necessary to have the candidate go to the Tool Locker for JPMs that require material from the locker. Describing where the locker is located, how to access key to unlock it, and what procedure packages/material would be obtained is sufficient for subsequent JP Ms.

3.

Revised step #3 to address that the required 50' section of hose is in a nearby locker

4.

Reworded step #5 to match step #4.

5.

Balded cue in step #11 since it is required for JPM to progress.

6.

Balded cue in step #17 since it is required for JPM to progress.

JPM IP-j RPV Venting During Containment Flooding IP-k Loss of RBCCW -

Plant Actions for the Instrument Nitrogen System RO-COO #1 Manually Calculate Drywell Bulk Average Temperature with Failed Points Change(s)

1. Added the following Note at the beginning of the JPM:

a.

IF this is the first in-plant JPM then the license candidate is required to go to the EOP Tool Locker. For subsequent in-plant JPMs it is not necessary to have the candidate go to the Tool Locker for JPMs that require material from the locker. Describing where the locker is located, how to access key to unlock it, and what procedure packages/material would be obtained is sufficient for subsequent JPMs

2.

Updated standard in steps 5,6,10, and 11 to better address that a telephone must be used for communication while in the Cable Spreading Room.

3.

Balded cue in steps 5,6,10 and 11 since it is required for JPM to progress.

4.

Added a balded cue in steps 12 since it is required for JPM to progress.

1.

2.

Added the following Note right before step #4:

"For the next 2 steps, after the examinee identifies the location of the control station for the A0-4230B on the TIP Room Roof, have them demonstrate required actions back on the control station for the A0-4230A since it is more accessible."

Balded cue in step 10 since it is required for JPM to progress.

No changes needed.

SRO-COO#l Resolution of Thermal Limits SRO-C00#2 Evaluate Overtime Work Request SRO-EC Review a Temporary Procedure Change - Change of Intent SRO-RC Perform Primary Containment Purge I Vent Isolation Valve Cumulative Log

1.

Corrected spelling of "Monicore" in title on cover page.

2.

Top of page 3 - added to print the 3D Pl edit on green paper (Unit 3).

3.

Revised cue to identify Unit 3 and have Examinee "document all actions/notifications".

4.

Added applicable GP-13 "Resolution of Thermal Limits" step numbers into JPM steps# 4,, 5, 6, and 8.

5.

Added TS LCO 3.2.2 wording in step #7

6.

Added Examiners Note prior to step #7 stating that the other unit (Unit 2) Tech Spec LCO 3.2.2 applicability value is~ 23% RTP following EPU".

1.

Added OP-PB-101-111-1002, "Peach Bottom Operations Overtime Guidelines" to Tools and Equipment" and "Reference" sections.

2.

Revised "Task Conditions/Prerequisites and Cue Sheet" was completely revised with the following:

a.

Added present date/time reference

b.

Changed RO 1 to RO#l

c.

2/6, 2/7, and 2/8 work hours shifted

d.

The following new Cue Sheet questions were inserted:

i.

Determine whether or not RO #1 is able to cover the required shift.

ii. If applicable, document all work hour limits that would be exceeded if RO #1 works on Saturday 2/16.

iii. Determine whether or not RO #1 has already violated any work hour limits.

iv. If applicable, document all work hour limits that have already been exceeded.

1.

Revised wording in Task Standard to say "it has been identified that the proposed temporary change results in change of intent".

2.

Task Condition/Prerequisite now states: A Temporary Change has been prepared for ST-R-003-495-2, "CRD Scram Insertion Timing of Selected Control Rods During Hydro". Steps 5.8, 6.17, and Data Sheet 3 have been modified."

3.

Added new note at beginning of JPM: "Provide marked up procedure change AND a copy of procedure AD-PB-101-1003 to the Examinee."

4.

Added that either decision choice in step #2 is acceptable.

5.

The Temporary Change Control Form will be printed on white paper.

6.

The word "Criteria" will be removed from the Temporary Change Control Form.

1.

Changed last sentence of Initiating Cue to "Document all errors on procedure copy, if applicable."

JPM Change(s)

RO-COO #2 This JPM setup was modified to have the Examinee do a Scram margin check on APRM #1 Perform an APRM (used to be APRM #4) as part of performing ST-0-001-200-2, "Turbine Stop Valve Closure Scram Margin Check and EOC-RPT Functional" in order to determine if there is enough scram margin to continue on with the test. APRM #1 was selected since the procedure will require a scram margin check on it first.

1.

Added the following to Section B "Tools and Equipment"

a.

"A copy of ST-0-001-200-2, Rev. 29, "Turbine Stop Valve Closure and EOC-RPT Functional", with procedure steps marked up and completed up to and including step 6.1.1. The Examinee will continue the surveillance at step 6.1.2."

2.

Cue Sheet now states:

a.

Unit 2 is at full power (approximately 100%).

b.

ST-0-001-200-2, "Turbine Stop Valve Closure and EOC-RPT Functional", is in progress.

c.

All APRM Channels are operable.

d.

INITIATING CUE:

The Control Room Supervisor directs you to perform steps 6.1.2 and 6.1.3 of ST-0-001-200-2, "Turbine Stop Valve Closure and EOC-RPT Functional". Inform the Control Room Supervisor when the steps are completed and whether the surveillance can be continued.

3.

Step #10 - Changed calculated Scram Margin value from 17.7% to 16.1%.

4.

Added the following note after step #10 to ensure that only one scram margin check is performed:

At this time the Evaluator should inform the Examinee that the remaining 3 APRM Scram Margin checks have been completed with the same results obtained.

5.

Balded cue in steps 1, 2, and 13 since it is required for JPM to progress.

RO-EC

1.

Deleted the Note from step # 8.

Determine Status of

2.

Deleted the word "loud" in the description of the SV humming noise from both the Instrument Nitrogen Task/Prerequisites and the Initiating Cue.

Compressor

3.

Deleted the drawing references from the Initiating Cue.

Discharge SV Using

4.

Balded cue in step #8 since it is required for JPM to progress.

Station P&IDs RO-EP

1.

Added "Proprietary - Not for Public Release" on the top of the cover page.

ERO Response

2.

Added to Tools and Equipment Section: "Optional - a pager that will receive a call Augmentation Using out from Scenario 11 to verify that the call out was successfully completed.

the Everbridge Call Contact the Site EP Coordinator.

Out System

3.

Changed Initiating Cue to read "Obtain a PEER CHECK from NRC Examiner.......

II

4.

Note after step #6 changed as follows:

a.

"Evaluator" to "NRC Examiner"

b.

Removed "true {valid)" from peer check description

c.

Underlined "to minimize risk of initiating an actual ERO callout."

JPM Change(s)

SRO-EP

l. Added the following to the Tools and Equipment section:

Make EAL

a.

"Ensure there are multiple copies of EP binders available that contain all Classification &

the procedures listed below in the References Section C."

State /Local

b.

Copy of PMS Met Data screen that displays a wind direction of 75 Notifications for degrees and a wind speed of 5 mph.

ALERT - Inability

2.

Initiating Cue changed to:

to Maintain Cold

a.

As the Emergency Director:

Shutdown

i.

Make the EAL classification. Inform the Proctor when you have made a declaration.

ii. Complete the State/Local Event Notification form. Inform the Proctor when you have completed the State/local Event Notification Form.

3.

Changed step #7 Standard to ALERT Classification.

Scenario 1 Changes Based on NRC Comments Event 1 {Secure the Drywell Purge Line up) o Added detail on how to secure Drywell purge including:

Procedure number and title

• Valve nomenclature

• Step numbers Expected alarms received during the procedure Event 2 {Failure of a Primary Containment Isolation valve) o Moved Event 6 to Event 2 and changed the valve from A0-2505, "Drywell Air Purge Inlet Valve" to A0-2506, "INBD 18" Vent Valve".

o Event 2 initiated when the crew begins to secure SBGT.

Event 3 {Control Rod Withdrawal) o Moved from event 2 to 3.

Event 4 {Mispostioned Control Rod) o Moved from event 3 to 4.

o Moved the mispositioned rod to rod 42-35.

o Event 3 will be complete before Event 4 occurs. This allows the lead examiner to move from event 4 to event 5 without requiring the crew to withdraw more Control Rods.

o Add the specific ON-122 steps into the scripting.

Event 5 {Loss of Main Condenser Vacuum) o Moved from event 4 to 5.

o Included all alarms that are received during the initial event not just the one the crew needs to reference.

o Included the steps from SO 8A.6.A-2, "Placing the Standby SJAE in Service and Placing the in Service SJAE in Standby" incase the crew does not use OT-106.

Event 6 {Spurious start of A and C Core Spray pumps and a leak at the discharge of the A Core Spray pump) o Moved from event 5 to 6.

o Added cooling tower trouble alarms.

o Added use of the keylock switch to isolate the Core Spray suction.

o Added monitoring of Torus level.

o Added steps to rack out Core spray pump breaker.

Event 7 {Startup level control valve fails closed) o Added ARC 201 H-1 Feedwater Field Instrument Trouble Event 8 {Steam leak in the Drywell) o Added information about chiller and pump starts to maximize drywell cooling.

o Included potential entry into T-117, "Level Power Control" o Added detail on how to place HPCI into short term shutdown.

Event 9 {ATWS}

o Balded steps that are related to the Critical Task.

o Added detail on how to initiate ARI.

Event 10 {RHR pump in Drywell Spray Trips) o Balded steps that are related to the Critical Task.

o Added detail on how to secure RHR following the RHR pump trip.

o Assed detail on how to recommence Containment Sprays using the other RHR pump in the same loop or using the other RHR loop.

Scenario 2 Changes Based on NRC Comments Added blocking tag for the E-332 breaker to enhance the Tech spec call.

Included the 1/0 overrides to block the E-332 breaker.

Event 1 (Master Trip Solenoid valves routine test) o No changes Event 2 (E4 Diesel Generator spurious start) o Moved the note to the Simulator Operator Directions section.

Event 3 (RCIC 250 VDC bus failure) o Add Tech Spec 3.5.3 (RCIC) and Tee Spec 3.6.1 (PCIV) references.

o Add a step indicating that the Crew may test annunciators.

Event 4 (TBCCW pump trips with failure of standby pump to start) o No changes.

Event 5 (Condenser Air In-leakage/fast power reduction w/ Recirc) o Included ARC 03 E-3 for "2 Unit Off Gas Recombiner Trouble" alarm.

Event 6 (A RFP will not respond to changing Reactor Power) o No changes.

Event 7 (Main Turbine high vibration I Reactor Scram) o Added an option step to scram instead of a GP-9 power reduction.

Event 8 (ATWS) o Added information about when to restore instrument nitrogen.

o Added information if the Recirc pumps trip on the 13 KV fast transfer or ARI initiation.

o Balded Critical task steps.

Event 9 (Loss of EHC I Loss of RPV pressure control) o No changes.

Changed the termination criteria to include having Torus cooling in-service and RPV pressure control with SRVs.

Scenario 4 Changes Based on NRC Comments Fixed the critical task for using the scram pushbuttons.

Defeated ARI to support the critical task for using the scram pushbuttons.

Event 1 (Placing the "B" EHC pump n-service) o Added computer point as an option to verify pressure for Pl-4403.

o Added alarm 205 K-3, "EHC Standby Pump not in Auto".

o Added a step to provide discharge pressure of 1555 psig if asked as an Equipment Operator.

Event 2 (Core Spray Snubber INOP) o No changes Event 3 (A RWCU Pump High Winding Temperature) o Added that the Crew may close M0-2-12-15, "RWCU Inboard lsol" but it is not required.

Event 4 ("B" ESW Room Flooding) o Reformatted so the actions directed from the Room Flooding ARC are clear.

Event 5 (Loss of RBCCW Cooling) o Added ARC 214 A-5 and 214 F-5.

o Added step for Nitrogen compressor trouble if necessary.

o Added a step to direct monitoring of Recirc pump temperatures.

Event 6 (Failed Seal on the "B" Recirc Pump) o Added ARC reference for seal failures.

o Added detail for how to maximize drywell cooling.

o Changed guidance to trip the Recirc pump instead of secure the Recirc pump.

o Removed the step to reduce the speed to 30% since the Recirc pump will be tripped.

Event 7 (Suction valve fails to close) o Added this as a separate event Event 8 (Failure of the Auto Scram and Reactor Mode Switch) o Balded Critical task step.

o Added detail that while the mode switch is in Shutdown it did not cause the Scram.

Event 9 (Steam leak worsens) o Added ON-120 entry condition.

o Listed the entry conditions for T-102.

o Balded Critical tasks.

o Added detail on how to trip HPCI.

o Added detail on how to spray the Torus.

o Added detail on how to spray the Drywell.

o Added steps if the Crew decides to do a T-112 blowdown at 281°F.

Event 10 (Drywell to RBCCW Leak) o Balded Critical tasks.