ML070610437

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Draft - Outlines (Folder 2)
ML070610437
Person / Time
Site: Pilgrim
Issue date: 12/11/2006
From:
NRC Region 1
To:
Entergy Nuclear Operations
Sykes, Marvin D.
Shared Package
ML060800104 List:
References
Download: ML070610437 (32)
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Text

ES-401 BWR Examination Outline Form ES-407-1 Tier I.

Emergency Abnormal Plant Evolutions

2.

Plant Systems Facilitv:

Pilarim Date of Exam:

2/26/2007 Group K I K2 K3 K4 1 K5 I K6 AI A2 A3 [ A4 G* Total A2 G* Total 1

3 3

4 5

2 2

19 3

5 8

1 8

2 2

4 2

I 2

2 Tier 4

5 6

6 3

3 27 5

7 12 N/A 1

1 N/A Totals I 4 4

2 1 3 3

1 0

4 1

2 2 1 2 6 2

2 4

2 1

0 0 1 0 1

2 2

0 2

2 2 1 1 2 010 2

2 Tier 5

4 2 ( 3 4 3

2 4

3 4

4 I 38 2

4 6

Totals I

I I

I RO WA Category Points I SRO-Only Points 10

3. Generic Knowledge and 1

2 3

4 4

2 2

2 Abilities Categories 1

2 3

4 1

2 2

2 Note:

1.
2.

~~

Ensure that at least two topics from every WA category are sampled within each tier of the RO outline (Le., the Tier Totals in each WA category shall not be less than two). Refer to Section D. 1.c for additional guidance regarding SRO sampling.

specified in the table. The final point total for each group and tier may deviate by k 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.

1 The point total for each group and tier in the proposed outline must match that I 3.

Select topics from many systems and evolutions; avoid selecting more than two WA topics from a given system or evolution unless they relate to plant-specific priorities.

I 4.

I Systemdevolutions within each group are identified on the associated outline.

9.

The shaded areas are not applicable to the categoryhier.

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

The SRO WAS must also be linked to 10 CFR 55.43 or an SRO-level learning objective.

On the following pages, enter the WA 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; summarize all the SRO-only knowledge and non-A2 ability categories in the columns labeled K and A.

Use duplicate pages for RO and SRO-only exams.

For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form ES-401-3.

Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate KIA statements.

t NUREG-I021 1

ES-4d.i 295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 & 4 2.1.20 steps 295003 Partial or Complete Loss of AC I 6 X

Conduct of Operations: Ability to execute procedure Conduct of Operations: Ability to apply technical specifications for a system Conduct of Operations: Ability to perform specific modes of plant operation.

Emergency Procedures / Plan Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

Ability to determine and/or interpret the following as Drywell pressure Ability to determine andlor interpret the following as they apply to SCRAM Condition Present and Power Above APRM Downscale or Unknown: Containment conditionsIisolations Conduct of Operations: Ability to execute procedure Ability to determine andlor interpret the following as X

EA2.03 they apply to LOW SUPPRESSION POOL WATER LEVEL : Reactor pressure Ability to determine andlor interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Jet pump operability: Not-BWR-l&2 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER : Systems necessary to assure safe plant shutdown Knowledge of the interrelations between PARTIAL X

AK2.03 OR COMPLETE LOSS OF D.C. POWER and the following: D.C. bus loads Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP : Pressure effects on reactor level 2,1.12 295006 SCRAM I 1 X

2.1 2 3 system and integrated plant procedures during all 2'4.4 295021 Loss of Shutdown Cooling I 4 X

295024 High Drywell Pressure I 5 X

EA2.01 they apply to HIGH DRYWELL PRESSURE:

EA2,07 X

295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown I 1 2.1.20 steps.

295028 High Drywell Temperature I 5 X

~

~

295030 Low Suppression Pool Water Level I 5 X

295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4

~~

AA1.03 295003 Partial or Complete Loss of AC I 6 X

295004 Partial or Total Loss of DC Pwr I 6

~~~

~

AK1.03 295005 Main Turbine Generator Trip I 3 X

L i

Pilgrim 2007 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group I 4,2 76 4.0 77 4.0 78 4.3 79 4.4 ao 4.2 81 4,2 82 3.9 83 3.,

39 4.4 4o 3.3 41 3.5 42

("

Form ES-41-1 NUREG-I021 2

('

Pilirirn 2007 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group I 43 -

44 -

45 -

46 -

48 f'

Form ES--tdl-1 295006 SCRAM I 1 295016 Control Room Abandonment I 7

295018 Partial or Total Loss of CCW I 8 29501 9 Partial or Total Loss of Inst. Air / 8 295021 Loss of Shutdown Cooling I 4 1

E/APE # / Name Safety Function I G I K1 1 K2 1 K3 I AI 1 A2 I Number 1 K/A Topic(s) 1 Imp. I Q# I X

X X

X X

AK2'05 AK3.03 3.1 Knowledge of the interrelations between SCRAM and the following: CRD mechanism Knowledge of the reasons for the following ABANDONMENT : Disabling control room controls responses as they apply to CONTROL ROOM 3.5 AK3.05 AK3.02 AKI.03 3.2 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Placing standby heat exchanger in service Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR :

Standby air compressor operation Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING : Adequate core cooling 3.5 295032 HI Secondary Containment Area Temps I 8 295024 High Drywell Pressure 15 3.9 X

EK2.01 X

EA1.12 295030 Low Suppression Pool Water Level I 5 295031 Reactor Low Water Level I 2 X

EK3.01 X

2.2.22 Knowledge of the interrelations between HI SECONDARY CONTAINMENT AREA TEMP and the following: Arealroom coolers

~~

___________~

Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE:

Suppression pool sway: Mark-lbll 295025 High Reactor Pressure I 3 Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE : Pressure effects on reactor power 3.9 51 295026 Suppression Pool High Water Temp. I 5 295028 High Drywell Temperature I 5 EA1.03 EA2.03 Ability to operate andlor monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Temperature monitoring Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE :

Reactor water level Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: Emergency depressurization Equipment Control Knowledge of limiting conditions for operations and safety limits.

3.8 -

3.4 54 -

55 NUREG-I 021 3

jr Pilgrim 2007 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier I Group I 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown / 1 295038 High Off-site Release Rate I 9

('

Form E S - ~ d l - l Emergency Procedures I Plan Knowledge symptom Ability to operate and/or monitor the following as they apply to high off-site release rate: Stack gas monitorina svstem.

2'4'6 based EOP mitigation strategies.

X EA1.Ol KIA Category Point Totals:

215 3

3 4

5 Z3 Group PointTotal:

Ability to operate and I or monitor the following as they apply to PLANT FIRE ON SITE: Fire alarm 1 918 3.0 I 58 1 NUREG-IO21 4

li" ES-40'1 X

X f

Form E S - d -1 2.4.30 AA2.01 2.2.25 EA2.03 2.4.49 AK2.03 Pilirim 2007 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier I Group 2 Emergency Procedures I Plan Knowledge of which events related to system operations/status should be reported to outside agencies.

Ability to determine andlor interpret the following as Drywell temperature they apply to HIGH DRYWELL TEMPERATURE :

,95010 High Drywell Pressure I 5 295012 High Drywell Temperature I 5 3.6 3.9 295029 High Suppression Pool Water Level I5 500000 High CTMT Hydrogen Conc. I 5 295007 High Reactor Pressure I 3 295009 Low Reactor Water Level I 2 295013 High Suppression Pool Temperature l 5 295014 Inadvertent Reactivity Addition I 1 295015 Incomplete SCRAM I 1 295020 Inadvertent Cont. Isolation I 5 & 7 295022 Loss of CRD Pumps I 1 295029 High Suppression Pool Water Level I 5 X -

AA1.01 t AK2.01 AA2.02 t AK3.03 AK3.01 t EKI.01 84 85 Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Ability to determine and /or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Combustible limits for drywell Emergency ProcedureslPlan: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls 3.7 3.8 -

4.0 Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following:

Recirculation system 3.1 I 59 Ability to operate andlor monitor the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE : Suppression pool cooling Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following: RPS Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM : Control rod position Knowledge of the reasons for the following responses as they apply to INADVERTENT CONTAINMENT ISOLATION: Drywelllcontainment temperature response 4.1 3.2 62 -

63 Knowledge of the reasons for the following responses as they apply to LOSS OF CRD PUMPS:

Reactor SCRAM Knowledge of the operational implications of the following concepts as they apply to HIGH SUPPRESSION POOL WATER LEVEL :

Containment integrity NUREG-I021 5

hl Q

3 8

7 m

0 I-4 - 4 T-hl 0

(*, i, ES-46;r 3.2 3.7 3.2 4.0 2.7 3.8 3.8 3.2 3.2 2.7 3.9 I'

Pilgrim 2007 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 88 89 90 91 1

2 3

4 5

6 7

('

Form E s - 4 ~

1-1 2'1'12 K2.03 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty range switch Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Conduct of Operations: Ability to apply technical specifications for a system.

Knowledge of electrical power supplies to the following: Initiation logic X

K2,01 K1'08 A2.03 Knowledge of electrical power supplies to the following: System valves: BWR-2,3,4 Knowledge of the physical connections andlor cause-effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: A.C. electrical power Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High train temperature K3.03

~

Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: Reactor temperatures (moderator, vessel, flange)

Ability to manually operate and/or monitor in the control room: Turbine soeed controls Knowledge of the physical connections andlor NUREG-1021 7

a3 al 0

r In r

x 2

c hl Y s a3 9

9 X

X X

X X

X X

X X

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0 0

m hl 2

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0 0

m hl 0

0 0

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2 0

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3 5 2

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C 0

0" 5

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ES-40.1 3.3 2.4.28 emergency response to sabotage.

2.,,23 Ability to perform specific and integrated plant 4,0 procedures during all modes of plant operation Conduct of Operations: Knowledge of purpose Knowledge of procedures relating to 204000 RWCU X

286000 Fire Protection X

Pilgrim 2007 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 Knowledge of the operational implications of and Reheat Steam system: Flow Indication Ability to predict and/or monitor changes in parameters associated with operating the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) controls including: Rod position: P-Spec(Not-BWR6) the following concepts as they apply to Main Form E S - 4 ~

1-1 2.8 3.2 204000 RWCU 21 5001 Traversing In-core Probe Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER CLEANUP SYSTEM : CRD hydraulics: Plant-Specific Ability to manually operate and/or monitor in the K6.09 2'7 X

X A4.03 control room: Isolation valves: Mark-l&II(Not-3.0 21 5002 RBM I

216000 Nuclear Boiler Inst.

Ability to monitor automatic operations of the X

A3.01 ROD BLOCK MONITOR SYSTEM including:

3.1 Four rod display: BWR-3,4,5 Equipment Control Knowledge of bases in X

2.2.25 technical specifications for limiting conditions 2.5 for operations and safety limits.

X 226001 RHFULPCI: CTMT Spray Mode Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI:

K1.I 1

CONTAINMENT SPRAY SYSTEM MODE and 2.8 the following: Component cooling water systems A3.09 X

241 000 Reactorrrurbine Pressure Regulator NUREG-1021 10 Ability to monitor automatic operations of the REACTOWURBINE PRESSURE REGULATING SYSTEM including:

ControVgovernor valve operation 3.3 256000 Reactor Condensate Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CONDENSATE SYSTEM controls including: Hotwell level 2.9 A,.04 X

i/

ES-4d'i 228000 Plant Ventilation WA Category Point Totals:

Pilgrim 2007 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 Ability to operate and/or monitor in the control 3.1 38 room: Start and Stop fans A4.01 2/2 1

0 0

0 1

2 2

O/O 2

2 Group PointTotal: I 1212

('.

Form E s - 4 ~

1-1 NUREG-I 021 11

Facility: I Category

1.

Conduct of Operations

2.

Equipment Control

3.

Radiation Control

4.

Emergency Procedures / Plan Tier 3 Point Total Pilgrim I

DateofExam: I 2/26/2007 WA# I Topic Knowledge of less than one hour technical specification action statements for systems.

SRO-Only IR I Q#

66 Ability to make accurate, clear and concise 2.9 2'1.18 I logs, records, status boards, and reports.

I 3.8 Ability to coordinate personnel activities outside the control mom.

2.1.8 I I Abilitv to obtain and intermet station

- 1 Knowledge of the parameters and logic used to assess the status of safety functions including: 1 Reactivity control 2.

Core cooling and heat removal 3. Reactor coolant system integrity 4. Containment conditions 5. Radioactivity release control.

Knowledge of annunciator response Procedures.

2.4'21 2'4'10 Subtotal 2.9 3.9 74 75 2

10 4.3 3.1 99 100 2

7 N UREG-102 1 12

Tier /

Randomly Group 2 1 2 111 2 1 1 Selected WA 201005 K5.02 295028 G2.1.27 26200 I 2 1 2 G2.1.30 272000 2 1 1 K6.01 206000 A4.1 I 211 2 1 2000 K5.01 295005 I

AK1.02

,,," I 295024 I G2.2.26 L I L 295005 AA.2.04 201 002 G2.1.I4

  • "2 04 0 0 0

I I A2.11 Reason for Rejection System does not exist at facility. Randomly reselected 239001 K5.05 Impossible to meet KA Topic requirement at SRO level. Randomly reselected G2.1.20 for APE.

Impossible to meet KA Topic requirement at SRO level. Randomly reselected G2.1,12 for system.

No effect between systems, either directly or indirectly. Randomly reselected K6.03 for system Component does not exist at facility. Randomly reselected A4.03 for system. Subsequently randomly selected A4.01 for same reason.

Concept does not apply to facility. Randomly reselected K5.02 for svstem

~

Action does not exist at facility. Randomly reselected A2.09 for topic No action for condition at facility. Randomly reselected A2.12 for topic Not applicable to facility. Randomly reselected AKI.03 for topic Impossible to meet topic requirement at SRO Level. Randomly reselected EA2.01 for topic Impossible to develop question applicable to facility. Randomly reselected G2.1.23 for topic Topic area not directly or indirectly pertinent to system. Randomly reselected 295007 2.4.49 4.0 (Q. #47) This moved a T I G I topic to a T I G2 topic Impossible to meet KA Topic requirement at SRO level. Randomly reselected G2. I

.20 (4.2) for APE. (Q. #76)

Impossible to meet KA requirement at SRO level. Randomly reselected 295003 G2.1.I2 (4.0) for system. (Q. #77)

Topic area not directly or indirectly pertinent to system. Randomly reselected G2.1.28 (3.2) for system. (Q. #27)

Double jeopardy with question #83. Randomly reselected 295037 EA2.07 (4.2) for system. (Q. #81)

No operational valid question could be written for the topic. Randomly reselected EA 1.01 for topic (Q. #57) This moved a T I K l to a T I A? topic Double jeopardy with Q. #19. Randomly reselected 295032 EK2.01 (3.5) for Q.##49 Operational valid SRO level question could not be written for topic. Randomly reselected G2.4.28 (3.3) (Q. #92) This moved a T2 A2 to a T2 G toDic NUREG-1021 13

NUREG-1021 14

Record of Rejected WAS I Form ES-401-4 I ES-401 A4.11 system 212000 Z I 1 n, A Concept does not apply to facility. Randomly reselected K5.02 for K5.01 system 217000 A2.17 261000 A2.14 295005 AKI.02 LI I Action does not exist at facility. Randomly reselected A2.09 for topic No action for condition at facility. Randomly reselected A2.12 for topic Not applicable to facility. Randomly reselected AK1.03 for topic 211 211 Ill NUREG-I021 12

ES-301 Ad mi n ist rat ive Topics Out I i ne Form ES-301-1 Facility:

Pilgrim Examination Level (circle one):

RO Conduct of Operations N

Conduct of Operations Equipment Control N

Radiation Control N

Emergency Plan Date of Examination:

2/26/2007 Operating Test Number:

NRC

~~~~~~~~

Describe activity to be performed JPM - Perform a Short Form Heat Balance KIA: 2.1.7 (3.7)

Ability to evaluate plant performance and make operational judgments based on operating characteristics I reactor behavior I and instrument interpretation.

JPM -Verify AOG Recombine Operation KIA: 2.2.13 (3.8)

Ability to obtain and interpret station reference materials such as graphs I monographs / and tables which contain performance data.

JPM: Prepare an SLC Pump tagging clearance KIA: 2.2.13 (3.0)

Knowledge of tagging and clearance procedures (3.6).

JPM: Determine Stay Time.

KIA: 2.3.2 (2.5)

Knowledge of facility ALARA program 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 (D)irect from bank (I 3 for ROs; 5 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (> I)

(P)revious 2 exams (I 1; randomly selected)

(S)imulator NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1

'L, A.1.a The candidate will perform a short form heat balance by collecting the required plant data from control room indications and performing a manual calculation. The critical task will be to accurately calculate reactor thermal power. This is a modified bank JPM.

A.l.b The candidate will be required to identify improper recombiner operation by collecting the required plant data from control room indications and utilizing the appropriate attachments of PNPS 2.4.141, Abnormal Recombiner Operation. Plant data will indicate recombiner outlet temperature is not excessively high; however, abnormal recombiner delta-T will be indicated. The critical task will be to determine the recombiner is overheated at a reduced power level. This is a new JPM.

A.2 The candidate will prepare an SLC Pump tagging clearance. The critical task will be to prepare a tagout that assures equipment and personal safety. This is a new JPM.

A.3 The candidate will determine stay time given a set of plant conditions. The candidate must have knowledge of the facility limits to perform the calculation. The critical task will be to determine the correct stay time. This is a new JPM.

NUREG-I 021, Revision 9

ES-301 Ad minis t ra tive Topics Out I i ne Form ES-301-1 Facility:

Pilgrim Examination Level (circle one):

SRO Date of Examination:

21 26107 Operating Test Number:

NRC Ad m i n istrative To pic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan Type Code*

M N

N N

M Describe activity to be performed JPM - Perform a Short Form Heat Balance WA: 2.1.7 (4.4)

Ability to evaluate plant performance and make operational judgments based on operating characteristics / reactor behavior / and instrument interpretation.

JPM - Verify AOG Recombiner Operation WA: 2.1.25 (3.1)

Ability to obtain and interpret station reference materials such as graphs / monographs / and tables which contain performance data.

JPM: Review an SLC Pump tagging clearance WA:

2.2.13 (3.8)

Knowledge of tagging and clearance procedures JPM: Determine Stay Time.

WA:

2.3.2 (2.9)

Knowledge of facility ALARA program JPM: Perform Dose Assessment Using DAPAR Software WA:

2.4.44 (4.0)

Knowledge of emergency plan protective action recommendations VOTE:

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.

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1

  • Type Codes & Criteria:

(C)ontrol room (D)irect from bank (I 3 for ROs; I 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (> 1)

(P)revious 2 exams (I 1 ; randomly selected)

(S)imulator A.l.a The candidate will perform a short form heat balance by collecting the required plant data from control room indications and performing a manual calculation. The critical task will be to calculate the correct reactor thermal power. This is a modified bank JPM.

A.l.b The candidate will be required to identify improper recombiner operation by collecting the required plant data from control room indications and utilizing the appropriate attachments of PNPS 2.4.141, Abnormal Recombiner Operation. Plant data will indicate recombiner outlet temperature is not excessively high; however, abnormal recombiner delta-T will be indicated. The critical task will be to determine the recombiner is overheated at a reduced power level. This is a new JPM.

A.2 A.3 A.4 The candidate will review an SLC Pump tagging clearance. The critical task will be to identify all tagging errors. This is a new JPM.

The candidate will determine stay time given a set of plant conditions. The candidate must have knowledge of the facility limits to perform the calculation. The critical task will be to determine the correct stay time. This is a new JPM.

The candidate will perform a dose assessment using the DAPAR Software. The critical task will be to accurately perform the dose assessment and determine the appropriate protective action recommendations. This is a modified JPM.

NUREG-I 021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Date of Examination:

2/26/2007 Operating Test No.:

Facility:

PNPS Exam Level (circle one):

RO / SRO(I) / SRO (U)

Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)

System / JPM Title 201003 Control Rod and Drive Mechanism (JPM-201-11) s-I Reactor Startup to Criticality 21 7000 Reactor Core Isolation Cooling System (JPM-217-03a)

Inject to RPV with RClC 262001 A.C. Electrical Distribution (JPM-262-10)

Restoration of Power to 41 60 VAC Bus A5 from SUT 206000 High Pressure Coolant Injection System (JPM-206-09) s -2 s -3 s -4 Operate HPCl for pressure Control 400000 Component Cooling Water System (JPM-200-31)

Recover RBCCW Loop B with an elevated Drywell Temperature 239001 Main and Reheat Steam System (JPM-200-05) 5-5 S-6 Respond to an MSlV Closure 261 000 Standby Gas Treatment System (JPM-229-012)

Manually Start SBGT and Vent the Torus 21 2000 Reactor Protection System (JPM-212-04)

Reset a Reactor Scram (RO only) 5-7 5-8 n-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

Type Code*

Safety Function 1

2 6

4 8

3 9

7 2 12000 Reactor Protection System (JPM-200-22)

D. E 7

'-1 Reactor Scram from Outside Control Room 239001 Main and Reheat Steam System (JPM-200-16)

Defeat MSlV Isolation Signals 201 001 Control Rod Drive Hydraulic System (JPM-201-03)

Shift CRD Flow Control Valves D, E 3

'-2 D, R 1

'-3 NUREG-I 021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-30 1 -2 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 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power I Shutdown (N)ew or (M)odified from bank including I(A)

(P)revious 2 exams (S)imulator

( W A s-1 s-2 s-3 s-4 5-5 S-6 Criteria for RO I SRO-I I SRO-U 4-6 14-6 I 2-3

~ 9 1 ~ 8 1 ~ 4 2 1 / r l / r 1 2 1 / 2 1 / > 1 2 2 1 2 2 1 2 1 2 1 / 2 1 / > I 3 / 5 3 /-s 2 (randomly selected)

NRC JPM Examination Summary Description The candidate will continue a reactor startup, withdrawing control rods on a slightly subcritical reactor including the performance of CRDM coupling checks. The alternate path requires that the candidate recognize indications of an uncoupled control rod and take actions in accordance with PNPS 2.4.1 1, Control Rod Positioning Malfunctions. This is a bank JPM in the CRD system - Reactivity Control Safety Function.

The candidate will place RClC in injection mode and raise reactor water level. The alternate path requires that the candidate recognize a failure of the cooling water supply valve to reposition automatically and take manual action to open the valve. This is a modified JPM in the RClC system - Reactor Water Inventory Control Safety Function.

The candidate will perform a dead bus transfer and transfer the A5 bus back to the startup transformer during a loss of Off-Site power. This is a bank JPM under the AC Electrical Distribution - Electrical Systems Safety Function.

The candidate will start HPCl in pressure control mode. The alternate path requires that the candidate recognize a failure of the HPCl flow controller low and place the controller in manual and raise flow in to establish HPCl operation in pressure control. This is a modified JPM in the HPCl system - Heat Removal From Reactor Core Safety Function.

The candidate will take action to restore RBCCW Loop B system flow following a non-LOCA event that resulted in a reactor scram and elevated drywell temperatures. The alternate path requires that the candidate recognize a breach in the RBCCW System piping inside the drywell and isolate RBCCW system flow. This is a modified JPM in the Reactor Building Component Cooling Water System - Plant Service Systems Safety Function.

The candidate is required to recognize an MSlV closure, enter the appropriate procedures, and open the steam line drains. This is a bank JPM under the Main and Reheat Steam System -

Reactor Pressure Control Safety Function.

NUREG-1021, Revision 9

ES-301 Control RoomAn-Plant Systems Outline Form ES-301-2 s-7 The candidate is required to place the "A" train of SBGT in service to vent the primary containment through the Torus. The alternate path requires that the candidate terminate venting when an alarm is received which requires termination of venting. This is a bank JPM in the Standby Gas Treatment System - Radioactivity Release Safety Function.

'LJ' S-8 The candidate is required to bypass SDlV scram, reset scram, wait for SDlV to drain, and return the bypass switch to normal. This is a bank JPM in the Reactor Protection System -

Instrumentation Safety Function.

P-I The candidate is required to initiate a reactor scram from outside of the control room. This is a bank JPM in the Reactor Protection System - Instrumentation Safety Function.

P-2 The candidate is required to defeat main steam isolation signals to facilitate venting the reactor pressure vessel. This is a bank JPM in the Main and Reheat Steam System - Reactor Pressure Control Safety Function.

The candidate is required to shift CRD Flow Control Valves. This is a bank JPM in the Control Rod Drive Hydraulic system - Reactivity Control Safety Function.

P-3 NUREG-1021, Revision 9

Malf. No.

Event Type*

Event Description 1 Appendix D Scenario Outline Form ES-D-1 1

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Facility:

Pilgrim

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Scenario No.:

1 OpTest No.:

2007 Examiners:

Operators:

Initial Conditions:

Power was lowered to 90% for control rod pattern adjustment which has been completed

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HPCl 00s for Aux Oil Pump Replacement - 14 Day LCO RBCCW Pumu A (P-202A) OOC - Tracking LCO IRM G is bypassed.

LPRM 36-1 3-B is bypassed Turnover:

A CRD Suction filter D/P is high. Replace filter IAW 2.2.87, Attachment 2.

Return power to 100% using recirc flow Event No.

I NIA N - RO, 1 Swap CRD Pumps for suction filter replacement I SRO 2

NIA 1 R - RO 1 Raise reactor power with Recirc Fwo9

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I - RO, SRO Master FWLC fails as is 3

4 RBCCW B pump trip with failure of standby pump to auto start 5

FW23 I-RO, SRO TS-SRO FWLC NR Channel B fails high FW21 B C-RO, SRO M - All

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Condensate pump B trips.

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ED06 Loss of all offsite power RClC Injection Mode Push Button Fails RClC auto initiation fails I10 C - BOP 8

PCOl M - All Recirc leak in Drywell leads to Emergency Depressurization on low RPV level

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C-BOP

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A CS Injection valve (25A) failure to auto open B CS Injection valve (25A) failure to auto open B RHR LPCl iniection valve failure to auto ooen 129B) 9 CS02A CSO2B 10 C-BOP 11 RH04B C-BOP (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

I Appendix D Scenario Outline Form ES-D-1 I Scenario I Description Power was reduced to 90% for a control rod pattern adjustment. Due to CRD Pump suction filter high DIP, the FSS will request the crew to swap CRD Pumps. After CRD pumps have been swapped, the crew will raise reactor power with recirc as directed in the shift turnover.

When raising power with recirc, the FWLC Master Controller will fail as is requiring the crew to enter procedure 2.4.49, Feedwater Malfunctions and place the Master FWLC Controller in manual.

After level is stable with the FWLC Master Controller in manual, RBCCW pump B will trip and the standby pump will fail to auto start requiring operator action to start the standby RBCCW pump.

After Tech Specs are evaluated for RBCCW, FWLC Narrow Range Channel B fails high, requiring the crew to swap to Narrow Range Channel A and evaluate Tech Specs.

When Tech Specs have been referenced for the narrow range transmitter failure, Condensate Pump B will trip resulting in a trip of the B RFP and an automatic Recirc runback when RPV drops to 19 inches. Because the FWLC Master Controller is in manual, the crew must closely monitor and control RPV level during the Recirc runback.

When conditions have stabilized following the Condensate Pump trip and recirc runback, a loss of all offsite power will occur resulting in a loss of all feedwater and a reactor scram. The crew will enter and execute EOP-I. Following the scram, RClC will fail to initiate automatically, and the RClC Injection Mode Push Button will also fail, requiring the crew to manually align RClC and inject to the vessel. The crew will also maximize injection with the available CRD pump.

Once conditions have stabilized post scram, a recirc leak will develop in the drywell and drywell pressure will rise, requiring EOP-3 entry and EOP-1 re-entry. Torus cooling, torus spray, and drywell spray will be directed in accordance with EOP-3. As required by EOP-1, the crew will maximize flow with available high pressure injection systems; however, the recirc leak in Drywell will continue to worsen and vessel level will lower until Emergency Depressurization and injection with low pressure ECCS systems is required. When all SRVs have been opened and the low RPV pressure valve permissives are received, both Core Spray injection valves and the B LPCl injection valve will fail to open automatically requiring manual operator action to open the valves and recover vessel level.

Appendix D NUREG 1021 Revision 9

I Appendix D Scenario Outline Form ES-D-1 I 1

2 3

Facility:

Pilgrim Scenario No.:

2 OpTest No.:

2007 Examiners:

Operators:

NIA R - R O Pull rods to continue power ascension N - SRO Transfer RMS to run NM21 E I - RO, SRO APRM E fails downscale TS-SRO RD02 C - RO, In service CRD flow control valve fails closed SRO Initial Conditions:

Power is 2.5% with a startup in progress 2.2.96, Attachment 15, preset checks are done on all feedpumps Two Condensate pumps in service. Running additional condensate pump to wear new bearing in 4

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7 IRM G is bypassed RDO22-C-RO, Control rod 22-23 drifts inward 23 SRO TS - SRO NIA N - BOP RFP C intermittent TBCCW leak, place RFP B in service RWCU Pump A RBCCW Temp High, pump fails to auto trip I - BOP, SRO PC02 M -All RWCU leak leading to scram RM07 Turnover:

Continue startup. At 5% power, transfer RMS to run and resume pulling control rods.

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10 RH04B

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Malf. No. I EventType* 1 C - BOP SRV B fails to open Event Description

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8 I RP14A I I-RO I Manual scram failure. ARI required.

9 I RM07 I M - A l l I RWCU leak leads to Emergency Depressurization Appendix D NUREG 1021 Revision 9

1 Appendix D Scenario Outline FormES-D-I I ii W

Scenario 2 Description The crew will take the watch with a reactor startup in progress. They will withdraw control rods, continuing the startup until 5% power is obtained and the reactor mode switch is placed in run.

After the mode switch is in run, APRM E will fail downscale. The APRM will be bypassed, and after Tech Specs have been referenced, the startup will continue.

When the startup continues, the crew will discover that the in-service CRD Flow Control Valve has failed closed requiring action to remove the failed FCV from service and place the standby CRD Flow Control Valve in service.

After placing the standby CRD Flow Control Valve in service, the startup will continue until control rod 22-23 drifts inward. The crew will take actions in accordance with PNPS 2.4.1 1, Control Rod Positioning Malfunctions and address Tech Specs for the inoperable control rod. The startup will be halted.

While waiting for troubleshooting and reactor maneuvering plans, an intermittent TBCCW leak will develop on RFP C requiring the crew to place RFP B in service and secure RFP C in accordance with PNPS 2.2.96, Condensate and Feedwater.

After RFP B in service, an RBCCW high temperature condition will develop on the A RWCU Pump. RWCU pump A will fail to automatically trip and the crew will take action in accordance with the ARP to manually stop the pump.

When RWCU Pump A is stopped, a leak will develop on the RWCU pump and temperatures will rise in the RWCU pump room requiring entry into EOP-4. RWCU area temperatures will continue to rise until a manual scram is required, and the crew will enter and execute EOP-I. When manual scram is attempted, the scram push buttons and reactor mode switch will fail to initiate rod movement; however, all control rods will insert when the ARI pushbuttons are depressed.

Following control rod insertion, RWCU area temperatures will continue to rise until conditions are degraded in two of the areas specified in EOP-4 and Emergency Depressurization is required. The crew will execute EOP-17; however, one SRV will fail to open requiring the crew to utilize Alternate RPV Depressurization Systems (SRV Remote SD Panel) to augment emergency depressurization.

Appendix D NUREG 1021 Revision 9

1 Appendix D Scenario Outline Form ES-D-1 ]

1 Malf. No.

-acility:

Pilgrim Scenario No.:

3 OpTest No.:

2007 ixaminers:

Operators:

RD05A

~ RR13A RR13B N - BOP, SRO C - RO, SRO Remove Seawater pump 6 from service for emergency backwash A RRP Pump Motor Vibration High I - RO, SRO TS - SRO A IRM fails downscale

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C - RO, SRO R - RO, SRO A RRP Inner Seal Failure. A RRP Outer Seal Failure Insert control rods to exit the exclusion region C - BOP, SRO TS-SRO RCIC Steam Leak, failure to auto isolate.

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M - All C - RO C - R O Main Turbine Trip, ATWS.

First squib valve fails to open when fired RWCU M080 fails to isolate automatically nitial Conditions:

Power is 50%

iPCI 00s for Aux Oil Pump Replacement - 14 Day LCO iBCCW Pump A (P-202A) OOC, Tracking LCO

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RM G is bypassed.

-PRM 36-13-B is bypassed.

rurnover:

iemove Seawater Dump B from service and perform emergency backwash EventType* I Event Description Event No.

1 NIA 2

RRI jA 3

NM12A C-RO, SRO I B CRD Pump Trip 4

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NIA 7

TC06 8

TCO 1 9

10 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

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I Appendix D Scenario Outline Form ES-D-1 1 Scenario 3 Description After taking the watch, as directed in the shift turnover, the crew will remove Seawater pump B from service for emergency backwash of the inlet water boxes.

After backwash is in progress, the crew will respond to a Recirc Pump Motor Vibration High alarm. Recirc Pump speed will be lowered as directed by the ARP. When Recirc pump speed has been reduced a small amount, the Recirc Pump Motor Vibration Monitor will reset.

After the Recirc Pump Motor Vibration High alarm is reset, the A IRM will fail downscale. The crew will take actions in accordance with the ARP, and with IRM G previously inoperable and bypassed, a Tech Spec LCO will be entered.

After Tech Specs have been evaluated, the B CRD Pump will trip. The crew will take actions per PNPS 2.4.4, Loss of CRD Pumps, including immediate action to verify reactor pressure is greater than 950 psig. When directed, the RO will place the standby CRD Pump in service.

When the standby CRD Pump is in service, the A Recirc Pump inner and outer seals will fail in sequence. Initially, the inner seal will fail requiring action to monitor seal status and drywell conditions per the ARP as well as entry into PNPS 2.4.22. After a brief period of time, the outer seal will fail. When the crew determines a catastrophic seal failure has occurred, the A Recirc Pump will be tripped and isolated.

After the A Recirc Pump is tripped, the crew will estimate total core flow, plot the location on Power/Flow Map, and take actions in accordance with PNPS 2.4.17, Recirc Pump Trip and PNPS 2.4.165, Thermal Hydraulic Instabilities. Control rods will be manually inserted to exit the exclusion region. The Reactor Operator is expected to closely monitor LPRM indications for indications of thermal hydraulic instability.

When control rods have been inserted sufficiently, a steam leak will develop in the RCIC steam line, and RCIC will fail to isolate automatically. The crew is expected to take action to isolate RCIC manually; additionally, the CRS will enter and direct actions per EOP-4 which will direct RClC isolation if action has not been taken previously. When RCIC is isolated (and with HPCI already inoperable), the CRS will evaluate Tech Specs and enter a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO.

After Tech Specs have been addressed, the main turbine will trip and an ATVVS will result due to hydraulic lock of the scram discharge volume. The CRS is expected to enter and direct actions per EOP-02, including SLC initiation. When the SLC Actuate switch is placed in the SYS A or SYS B position, the selected SLC pump will start; however, the associated squib valve will fail to open as indicated by high SLC pump discharge pressure and zero flow. The reactor operator is expected to select the alternate train and inject SLC. Additionally, the RWCU return isolation valve (M080) will fail to auto isolate, requiring manual operator action to isolate the RWCU system. ATWS actions will continue per EOP-02 until all rods have been fully inserted; EOP-01 has been entered, and reactor level is in the normal band.

Appendix D NUREG 1021 Revision 9

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3 I Appendix D Scenario Outline Form ES-D-1 I NIA R - RO Reduce power with Recirc NM17 I - RO, SRO LPRM 36-45-B fails upscale 36-45-8 TS - SRO Facility:

Pilgrim Scenario No.:

4 OpTest No.:

2007 6

7 Examiners:

Operators:

MS14D M - All RH04B C-BOP Initial Conditions:

Power is 100%

9 HPCI 00s for Aux Oil Pump Replacement - 14 Day LCO RBCCW Pump A (P-202A) OOC, Tracking LCO PC22 M - All D SRV tail pipe fails leading to Emergency Depressurization LPRM 36-13-8 is bypassed IRM G is bypassed.

Turnover:

Shift TBCCW Pumps for maintenance vibration Test Lower Dower for control rod pattern adiustment Event No. I Malf. No. I Event Type* I Event Description 1

I NIA 1 N - BOP I Shift TBCCW Pumps for maintenance vibration Test

-c RR20A I BOP, SRO TS - SRO I - RO, SRO TS - SRO EPR Pressure Oscillations B Recirc Flow Controller fails upscale D SRV fails open, Manual reactor scram PASS H2102 Sample valve fails to Isolate CV91 I 8

1 I C - BOP I RBCCW to A RHR HX inlet valve fails shut Appendix D NUREG 1021 Revision 9

w I Appendix D Scenario Outline Form ES-D-1 I Scenario 4 Description After taking the watch, as directed in the shift turnover, the crew will shift TBCCW pumps. After the TBCCW pumps have been swapped, the crew will proceed with a planned power reduction using Recirc flow.

While the planned power reduction is underway, LPRM 36-45-B will fail upscale, and the CRS will enter and direct actions per PNPS 2.4.38, LPRM Failure. The crew will bypass the failed LPRM and verify that APRM AGAFs and thermal limits are in spec. The crew will also determine that the affected APRM has less than 2 LPRM inputs in a level, making the affected APRM inoperable per TS 3.1, Table 3.1.

After Tech Specs have been addressed, the EPR will begin to oscillate, and the CRS will direct actions per PNPS 2.4.37, Turbine Control System Malfunctions. The crew will take control with the MPR, and the EPR power control switch will be placed to off, stabilizing reactor pressure and power. With the EPR removed from service, the plant will enter an administrative LCO requiring both pressure regulators be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the plant be c 25% CTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

After the administrative LCO has been addressed, the B Recirc flow controller will fail upscale resulting in an increase in core flow and reactor power. When the flow controller failure has been diagnosed, the crew will initiate a scoop lockup per PNPS 2.4.20, Reactor Recirculation System Speed or Flow Control System Malfunction. The crew will then take actions per PNPS 2.4.19, Recirculation Pump MG Set Scoop Tube Lockup, including an evaluation of Recirc pump speeds against the Tech Spec 3.6.F limits. The CRS should also identify and brief the crew on the need to trip the B Recirc pump in the event of a reactor SCRAM.

When the required actions have been directed for the Scoop Tube Lockup, the D SRV will indicate open and Torus temperature will rise. The CRS will enter and direct actions per PNPS 2.4.29, Stuck Open Safety Relief Valve; however, the D SRV will fail to close, requiring a manual reactor scram. Following the scram, the RO should trip the B Recirc pump. The CRS will enter and direct EOP-01, and the BOP operator should identify a failure of PASS H2/02 Sample valve CV91 to isolate, requiring operator action to close the valve. With the SRV still open, Torus cooling will be initiated. The RBCCW inlet to the A RHR HX inlet valve will fail shut requiring additional operator action to lineup RBCCW. When Torus temperature rises to 80°F, EOP-03 will be entered.

After Torus cooling has been placed in service, the D SRV tail pipe will fail, resulting in rising Torus and Drywell pressure, and EOP-03 and EOP-01 will be re-entered on high drywell pressure. Torus and drywell spray will be initiated as Torus bottom pressure continues to rise; however, with the broken SRV tail pipe, Torus bottom pressure will continue to rise, and emergency depressurization will be required prior to exceeding the limits of the Pressure Suppression Pressure curve.

Appendix D NUREG 1021 Revision 9

u' Pilgrim Nuclear Power Station 2007 NRC Initial License Written Examination Written Examination Outline Methodology The written examination outline was developed using a proprietary electronic random outline generator developed by Western Technical Services, Inc.

The software was designed to provide a written examination outline in accordance with the criteria contained in NUREG 1021, Revision 9.

The application was developed using Visual Basic code, relying on a true random function based on the PC system clock. The random generator selects topics in a Microsoft Access Database containing Revision 2 of the BWR K&A catalogue. The selected data is then written to a separate data table. The process for selection of topics is similar to the guidance in ES-401, Attachment 1.

The attached outline report and plant specific suppression profile (not used for PILGRIM.

Suppressed topics are listed on attached page) report are written directly from the data tables created by the software. Electronic copies of the data tables are on file.

The process used to develop the outlines is as follows:

0 For Tier 1 and Tier 2 generic items, only the items required to be included in accordance with ES-401, Attachment 2 are included in the generation process.

0 The PILGRIM plant suppression profile lists all suppressed topics, either at the Topic level (System/EPE) or at the statement level. These items were suppressed prior to the electronic generation process.

0 Outline is generated for all topics with KA importance 22.5.

0 25 SRO topics are randomly selected from Tier 1 AA2 and required generic items, Tier 2 A2 and required generic items, and Tier 3 generic items (All with ties to 10CFR55.43). 75 RO topics are randomly selected to complete the outline, 100 topics total.

0 The exam report generated lists the topic (Question) number in the far right column. RO topics are numbered 1-75, and SRO topics are numbered 76-100.

The SRO topics are written in red ink for ease of identification.

Items that are rejected after the initial generation process are automatically placed on the rejected items page. The software tracks whether items are added manually or by random generation, and a report of outline modification may be generated.

0 Disposition of any item randomly selected but not included in the outline is documented and included.

Pilgrim Nuclear Power Station 2007 NRC Initial License Written Examination Written Examination Outline Methodology The following topics were suppressed because they either do not exist or the function is not performed at Pilgrim:

Emergency and Abnormal Plant Evolutions -Tier l/Group 1 (RO / SRO) 295027 High Containment Temperature / 5 Emergency and Abnormal Plant Evolutions - Tier VGroup 2 (RO / SRO) 29501 1 High Containment Temp / 5 Plant Systems -Tier 2IGroup 1 (RO / SRO) 207000 Isolation (Emergency) Condenser 209002 HPCS Plant Systems -Tier 2lGroup 2 (RO I SRO) 201004 RSCS 239003 MSlV Leakage Control

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