Text

Final Outlines (Folder 3)
	ML14094A153
Person / Time
Site:	Pilgrim
Issue date:	03/25/2014
From:	D'Antonio J; Operations Branch I
To:	; Entergy Nuclear Generation Co
	Jackson D
Shared Package
ML13295A007	List: IR 05000293/2014301; ML14049A493; ML14058A006; ML14094A153; ML14094A156; ML14094A157; ML14094A158; ML14094A160; ML14094A164; ML14094A165; ML14141A188; ML15043A711; ML15152A527;
References
	TAC U01876
	Download: ML14094A153 (49)
	v • d • e

ES-401 BWR Examination Outline Form ES-401-1 Facility: Pilgrim Date of Exam:

2014 NRC RO KIA Category Points SRO-Only Points Tier Group K

K K

K A

A A

A G

Total A2 G*

Total 1

2 3

4 5

6 1

2 3

4

1.

1 3

3 3

,.y~ ~~;~ i 3 4

~~'\\'

4 20 3

4 7

Emergency

~.~~.;~ ~~~

k*~~.~

f~

2 2

1 1

~y 1

1 1

7 2

1 3

Plant Evaluations Tier

~:... ~*

~***

Totals 5

4 4.*.....

4 5

5 27 5

5 10 1

2 2

3 2

2 3

3 2

3 26 2

3 5

2.

2 2

0 1

2 1

1 1

12 0

1 2

3 Plant Systems Tier 4

2 3

4 4

3 3

4 4

3 4

38 3

5 8

Totals

3. Generic Knowledge & Abilities 1

2 3

4 10 1

2 3

4 7

3 3

2 2

1 2

2 2

Note

1.

Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each KIA category shall not be less than two).

2.

The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by 1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3.

Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.1.b of ES-401, for guidance regarding elimination of inappropriate KIA statements.

4.

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

5.

Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6.

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

7.*

The generic (G) KIAs in Tiers 1 and 2 shall be select~3d 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 KIA's

8.

On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9.

For Tier 3, select topics from Section 2 of the KIA Catalog, and enter the KIA numbers, descriptions, IRs, and point totals(#) on Form ES-401-3. Limit SR.O selections to KIAs that are linked to 10CFR55.43 Pilgrim 2014 NRC Written Outline.doc

ES-401 Form ES-401-1 Pilgrim BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function 295023 Refueling Accidents I 8 295030 Low Suppression Pool Water Level/ 5 295031 Reactor Low Water Level/2 295018 Partial or Total Loss of CCW I 8 295025 High Reactor Pressure /3 295019 Partial or Total Loss of lnst. Air I 8 295004 Partial or Total Loss of DC Pwr I 6 295023 Refueling Accidents I 8 X

Pilgrim 2014 NRC Written Outline.doc G I KJA Topic(s)

AA2.04-Ability to determine

...*. f\\+. *. and/or interpret the following

.£ as they apply to t~L.: REFUELING ACCIDENTS :

li~~

Occurrence of fuel handling

. /,,

accident f:':.~

EA2.04 -Ability to determine t*

and/or interpret the following as they apply to LOW

~*

Zt~' SUPPRESSION POOL rt*, WATER LEVEL: Drywell/

t' suppression chamber 1 ~'~.

differential pressure: Mark-t;~... 1&11

>tC EA2.02 -Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL:

Reactor power 2.1.19 - Conduct of Operations: Ability to use plant computers to evaluate system or component status.

2.2.44-Equipment Control:

Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives effect plant and system conditions.

2.1.32 - Conduct of Operations: Ability to explain and apply all system limits and precautions.

2.2.25 - Equipment Control:

Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

AK1.01 -Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS : Radiation exposure hazards 4.1 76 3.7 77 4.2 78 3.8 79 4.4 80 4.0 81 4.2 82 3.6 39

ES-401 Form ES-401-1 Pilgrim BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function G I KIA Topic(s)

,, EK1.02-Knowledge of the operational implications of 295028 High Drywall

.i the following concepts as X

they apply to HIGH 2.9 40 Temperature I 5

/

DRYWELL TEMPERATURE

[~ :.

Equipment environmental

~~

"~

qualification L.,,,,,.,

AK1.04-Knowledge of the

. operational implications of 295004 Partial or Total

- '}

~ <,,

the following concepts as Loss of DC Pwr I 6 X

p i they apply to PARTIAL OR 2.8 41 COMPLETE LOSS OF D.C.

... **t POWER : Effect of battery

.*.s(~

discharge rate on capacity AK2.01 -Knowledge of the 600000 Plant Fire On-site I interrelations between 8

X

'.*~~***

PLANT FIRE ON SITE and 2.6 42 the following: Sensors, detectors and valves AK2.09 - Knowledge of the interrelations between 295019 Partial or Total X

PARTIAL OR COMPLETE 3.3 43 Loss of lnst. Air I 8 LOSS OF INSTRUMENT AIR and the following:

Containment EK2.1 0 - Knowledge of the interrelations between HIGH 295025 High Reactor X

REACTOR PRESSURE and 2.9 44 Pressure 13 the following:

SPDSIERISICRIDSIGDS:

Plant-Specific EK3.04-Knowledge of the reasons for the following 295026 Suppression Pool responses as they apply to X

SUPPRESSION POOL 3.7 45 High Water Temp. I 5 I

. >~ HIGH WATER

. TEMPERATURE:SBLC

..,\\ '

injection

,,.:,~:.'

~h i*..*

AK3.01 -Knowledge of the reasons for the following 295021 Loss of Shutdown X

~'~ii responses as they apply to 3.3 46 Cooling 14 I *.

LOSS OF SHUTDOWN I ~~;~~~.

COOLING: Raising reactor

. ". water level Pilgrim 2014 NRC Written Outline.doc

ES-401 Form ES-401-1 Pilgrim BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function I K1 I K2 I K3 I A 1 I A2 I G I KIA Topic(s)

EK3.05-Knowledge of the 295024 High Drywell

!J,,

reasons for the following X

~~;~

responses as they apply to 3.5 47 Pressure /5 HIGH DRYWELL PRESSURE : RPV flooding L,s'<' EA 1.07-Ability to operate

' r~r and/or monitor the following as they apply to SCRAM 295037 SCRAM Conditions CONDITION PRESENT Present and Reactor Power X

,,~~ 'Gi',,' AND REACTOR POWER 3.9 48 Above APRM Downscale or

,,'~i~, ABOVEAPRM Unknown /1 q

DOWNSCALE OR

~\\.

UNKNOWN : RMCS: Plant-

'****~...

Specific

,~~'

AA1.01 -Ability to operate and/or monitor the following 295016 Control Room "t

' '~~<~

Abandonment I 7 X

as they apply to CONTROL 3.8 49 ROOM ABANDONMENT :

[,',

RPS i>

AA 1.04 - Ability to operate and/or monitor the following 295003 Partial or Complete

'"' as they apply to PARTIAL Loss of AC /6 X

OR COMPLETE LOSS OF 3.6 50 A.C. POWER: D.C.

electrical distribution system EA2.04 - Ability to determine 295038 High Off-site and/or interpret the following X*

as they apply to HIGH OFF-4.1 51 Release Rate I 9 SITE RELEASE RATE :

,.,~,

Source of off-site release AA2.04 -Ability to determine and/or interpret the following 700000 Generator Voltage x'

as they apply to and Electric Grid GENERATOR VOLTAGE 3.6 52 Disturbances AND ELECTRIC GRID

'*~'

DISTURBANCES: VARs outside capability curve.

'\\*

~.~~c~:*

EA2.01 -Ability to determine l

and/or interpret the following 295030 Low Suppression n;~.

~*****

as they apply to LOW 4.1 53 Pool Water Level/ 5 if SUPPRESSION POOL

~~i' ? WATER LEVEL:

Suppression pool level Pilgrim 2014 NRC Written Outline.doc

ES-401 Form ES-401-1 Pilgrim BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KIA Topic(s)

(~,~~>

2.4.47 - Emergency Procedures I Plan: Ability to diagnose and recognize 295018 Partial or Total I

.~'~, trends in an accurate and 4.2 54 Loss of CCW I 8 I

timely manner utilizing the

',~~ appropriate control room

>~'

reference material.

','" s ; 2.4.21 -Emergency i,,

Procedures I Plan:

's,,

',~~~;;, Knowledge of the I

/

~. '_*-' parameters and logic used to assess the status of 295005 Main Turbine

~~~,;:,':;

safety functions, such as X,<;~

4.0 55 Generator Trip I 3 I

.~~~i;;; ',

reactivity control, core cooling and heat removal, reactor coolant system integrity, containment

. conditions, radioactivity release control, etc.

0',

'~

2.2.42-Equipment Control::

Ability to recognize system 295006 SCRAM I 1

)(

parameters that are entry-3.9 56

'~',,

level conditions for Technical Specifications.

EA2.03 - Ability to determine 295031 Reactor Low Water x

and/or interpret the following Level/2 as they apply to REACTOR 4.2 57 LOW WATER LEVEL:

Reactor pressure 2.2.25 - Equipment Control:

295001 Partial or Complete

>)'2"'~\\

Knowledge of bases in Loss of Forced Core Flow X'

:~

technical specifications for 3.2 58 Circulation I 1 & 4

,~,

limiting conditions for operations and safety limits.

KIA CategoryTotals 3

3 3

3 413 41~-

Group Point Total:

1 20/7 Pilgrim 2014 NRC Written Outline.doc

ES-401 Form ES-401-1 Pilgrim BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function 295020 Inadvertent Cont.

Isolation I 5 & 7 500000 High CTMT Hydrogen Cone. I 5 295007 High Reactor Pressure I 3 295014 Inadvertent Reactivity Addition I 1 295033 High Secondary Containment Area Radiation Levels I 9 295029 High Suppression Pool Water Level I 5 295012 High Drywell Temperature I 5 X

Pilgrim 2014 NRC Written Outline.doc X

X KIA Topic(s)

AA2.02-Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION:

Drywell/containment tem erature 2.4.8 - Emergency Procedures I Plan:

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

AA2.01 -Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE :

Reactor ressure AK1.01 - Knowledge of the operational implications of the following concepts as they apply to INADVERTENT REACTIVITY ADDITION :

Prom t critical EK2.02-Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS :

Process radiation monitoring s stem EK3.01 -Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL:

3.4 83 4.5 84 4.1 85 3.7 59 3.8 60 3.5 61 3.5 62

ES-401 Form ES-401-1 Pilgrim BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function G I KIA Topic(s)

EA2.02 -Ability to determine

~~~,: ~

and/or interpret the following

~c;'

as they apply to 295034 Secondary X

SECONDARY 3.7 63 Containment Ventilation CO NT AI NMENT High Radiation I 9

>' VENTILATION HIGH RADIATION : Cause of high

, I ~~ \\ radiation levels

[i~

2.2.14-Equipment Control:

295036 Secondary Knowledge of the process 3.9 64 Containment High for controlling equipment Sump/Area Water Level I 5 1~0~

conflg_uration or status.

AK1.03-Knowledge of the

> operational implications of 295015 Incomplete X

r**,,*<' the following concepts as 3.8 65 SCRAM I 1 they apply to INCOMPLETE SCRAM: Reactivity effects KIA CategoryTotals 2

1 1

1 112. 1t1 Group Point Total:

I 7/3 Pilgrim 2014 NRC Written Outline.doc

ES-401 System #/Name 209001 LPCS 203000 RHRILPCI:

Injection Mode 215004 Source Range Monitor 215003 IRM 264000 EDGs Pilgrim 2014 NRC Written Outline.doc Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)

A2.05 -Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM

and (b) based on those X

predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or

,>' operations: Core spray line break A2.12-Ability to (a) predict the impacts of the following on the RHRILPCI: INJECTION MODE (PLANT SPECIFIC); and (b)

X based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump runout 2.4.30 - Emergency Procedures I Plan; Knowledge of events related to system operation I status that X

must be reported to I

internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.

,\\

2.2.38 - Equipment

' Control: Knowledge of

,.J conditions and

' limitations in the facility license.

(

['~/

2.2.40 - Equipment Control: Ability to apply technical specifications for a system.

3.6 86 2.7 87 4.1 88 4.5 89 4.7 90

ES-401 System #/Name 300000 Instrument Air X

262002 UPS X

(AC/DC) 203000 RHR/LPCI:

Injection Mode X

2150031RM X

211000 SLC X

215005 APRM I LPRM X

Pilgrim 2014 NRC Written Outline.doc Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)

K1.02-Knowledge of the connections and I or cause effect I,,

relationships between INSTRUMENT AIR

\\

SYSTEM and the

'*' following: Service air K1.19 - Knowledge of the physical connections and/or cause-effect relationships between I

UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) and the following: Power range

',, neutron monitoring system: Plant-Specific K2.02-Knowledge of electrical power supplies to the following: Valves K2.01 - Knowledge of electrical power supplies to the following: IRM channels/detectors I

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 K3.07-Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will

~,. have on following: Rod block monitor: Plant-Specific 2.7 1

2.9 2 2.5 3 2.5 4 4.3 5 3.2 6

ES-401 System #/Name 206000 HPCI 209001 LPCS 259002 Reactor Water Level Control 215004 Source Range Monitor 264000 EDGs Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)

K4.14-Knowledge of HIGH PRESSURE COOLANT INJECTION SYSTEM design X

feature(s) and/or

,, interlocks which provide for the following: Control oil to turbine speed controls: BWR-2,3,4 i*,\\5y K4.02-Knowledge of LOW PRESSURE CORE SPRAY SYSTEM X

design feature(s) and/or interlocks which provide for the following:

Prevents water hammer K5.01 -Knowledge of the operational implications of the following concepts as X

they apply to REACTOR WATER LEVEL CONTROL SYSTEM :

GEMAC/F oxboro/Bailey controller operation:

Plant-Specific K5.03-Knowledge of the operational implications of the following concepts as X

. they apply to SOURCE RANGE MONITOR (SRM) SYSTEM :

Changing detector

'c~L _Q_osition

~

K6.08-Knowledge of the effect that a loss or malfunction of the X

following will have on the

~*,:z'; EMERGENCY GENERATORS

. *n (DIESEL/JET) : A.C.

power 3.4 7 3.0 8

3.1 9

2.8 10 3.6 11 Pilgrim 2014 NRC Written Outline.doc

ES-401 System #/Name 217000 RCIC 239002 SRVs 205000 Shutdown Cooling 223002 PC IS/Nuclear Steam Supply Shutoff Pilgrim 2014 NRC Written Outline.doc Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)

K6.01 -Knowledge of the effect that a loss or l

malfunction of the X

following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC):

Electrical power A 1.08-Ability to predict and/or monitor changes in parameters associated with X

~ *.*. operating the RELIEF/SAFETY

~* ** VALVES controls including: Suppression c.< pool water temperature A 1.05-Ability to predict and/or monitor changes in parameters associated with operating the X

SHUTDOWN COOLING SYSTEM(RHR SHUTDOWN COOLING MODE) controls including: Reactor water level A2.1 0 -Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR X

STEAM SUPPLY SHUT-OFF ; and (b) based on those predictions, use procedures to correct,

~

control, or mitigate the f*.. ; consequences of those

' abnormal conditions or

- operations: Loss of coolant accidents 3.4 12 3.8 13 3.4 14 3.9 15

ES-401 System #/Name 262001 AC Electrical Distribution 400000 Component Cooling Water 261000 SGTS 212000 RPS 218000 ADS Pilgrim 2014 NRC Written Outline.doc Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)

A2.04 -Ability to (a)

I \\

predict the impacts of the following on the A. C.

ELECTRICAL I

DISTRIBUTION; and (b) based on those predictions, use X

I' procedures to correct, control, or mitigate the

.. consequences of those abnormal conditions or I> ?z' operations: Types of loads that, if

.:f deenergized, would degrade or hinder plant operation A3.01 -Ability to monitor automatic operations of the CCWS including:

X Setpoints on instrument signal levels for normal operations, warnings, I

and trips that are applicable to the CCWS A3.03 -Ability to monitor automatic operations of X

the STANDBY GAS TREATMENT SYSTEM including: Valve operation A4.05 -Ability to X I manually operate and/or monitor in the control room: Reactor power A4.07 -Ability to manually operate and/or X

monitor in the control i'

room: ADS valve acoustical monitor noise:

Plant-Specific 3.8 16 3.0 17 3.0 18 4.3 19 3.5 20

ES-401 System #/Name 262002 UPS (AC/DC) 400000 Component Cooling Water 215004 Source Range Monitor 262001 AC Electrical Distribution Pilgrim 2014 NRC Written Outline.doc Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s) 2.1.7-Conduct of Operations: Ability to I

evaluate plant performance and make

)(}

operational judgments based on operating

~~ '~~$

characteristics, reactor behavior, and instrument intei"QI"etation.

2.1.2 - Conduct of Operations: Knowledge

- x~

of operator

~0'~:z, < ' responsibilities during all modes of plant operation.

A2.05 -Ability to (a)

I >

predict the impacts of the following on the SOURCE RANGE MONITOR (SRM)

SYSTEM ; and (b) based X

on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic operation of detectors/system A3.04 -Ability to monitor automatic operations of the A.C. ELECTRICAL X

DISTRIBUTION including: Load sequencing 2.7 21 4.1 22 3.3 23 3.4 24

ES-401 System #/Name 215005 APRM I LPRM 2150031RM KIA Category Totals 2

2 2

Pilgrim 2014 NRC Written Outline.doc Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)

K5.06-Knowledge of the operational implications of the following concepts as they apply to AVERAGE X

!,...., POWER RANGE MONITOR/LOCAL

. ;J POWER RANGE MONITOR SYSTEM :

Assignment of LPRM's

,. ~. to specific APRM channels 2.4.31 -Emergency Procedures I Plan:

X Knowledge of annunciator alarms, indications, or response procedures.

2 3

2 2

312 3

2 313 Group Point Total:

2.5 25 4.2 26 1 26/5

ES-401 System #/Name 256000 Reactor Condensate 202002 Recirculation Flow Control 204000 RWCU 204000 RWCU X

226001 RHR/LPCI:

CONTAINMENT SPRAY SYSTEM X

MODE Pilgrim 2014 NRC Written Outline.doc Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 2 KIA Topic(s)

A2.02 -Ability to (a) predict the impacts of the following on the REACTOR CONDENSATE SYSTEM; and (b) based X*

on those predictions, use procedures to correct, control, or mitigate the i~

.*. consequences of those abnormal conditions or

~*,

operations: Valve closures I

'** 2.4. 11 - Emergency X

Procedures I Plan:

Knowledge of abnormal condition procedures.

2.4.12 - Emergency Procedures I Plan:

X Knowledge of general operating crew responsibilities during emerqency operations K 1. 14 - Knowledge of the physical connections and/or cause-effect relationships between REACTOR WATER CLEANUP SYSTEM and the following: Process sample svstem K1.08-Knowledge of the physical connections and/or cause-effect I

relationships between k

RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE and the following: Nuclear boiler

,' instrumentation 2.9 91 4.2 92 4.3 93 2.5 27 3.2 28

ES-401 System #/Name 215002 RBM X

Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 2 KIA Topic(s)

K3.01 -Knowledge of the effect that a loss or malfunction of the ROD BLOCK MONITOR SYSTEM will have on following: Reactor manual control system:

BWR-3,4,5 K4.07-Knowledge of

- - - ~:g REACTOR MANUAL CONTROL SYSTEM design feature(s) and/or 3.3 29

- - - ~..

201002 RMCS X

interlocks which provide 2.5 30 for the following: Timing of rod insert and withdrawal cycles (rod S.*

movement sequence timer)

K5.02-Knowledge of the operational

.. implications of the 201001 CRD X

following concepts as 2.6 31 Hydraulic they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM :

Flow indication K6.02-Knowledge of the effect that a loss or malfunction of the 214000 RPIS X

.. following will have on the 2.7 32 ROD POSITION INFORMATION SYSTEM : Position indication probe A 1.06 - Ability to predict and/or monitor changes in parameters 202001 associated with X

... operating the 2.5 33 Recirculation 3~,

RECIRCULATION SYSTEM controls

....

including: Recirculation pump motor amps Pilgrim 2014 NRC Written Outline.doc

ES-401 System #/Name 286000 Fire Protection 288000 Plant Ventilation 241000 Reactor/Turbine Pressure Regulator 219000 RHR/LPCI:

Torus/Pool Cooling Mode Form ES-401-1 Pilgrim BWR Examination Outline Plant Systems - Tier 2 Group 2 KIA Topic(s)

A2.12 -Ability to (a) i predict the impacts of the following on the FIRE PROTECTION SYSTEM ; and (b) based on those predictions, use X

procedures to correct,

~~i~

control, or mitigate the consequences of those abnormal conditions or f*~

operations: Low diesel

~' ;;, fuel supply: Plant-S_Qecific A3.01 -Ability to monitor

,' automatic operations of the PLANT X

VENTILATION SYSTEMS including:

Isolation/initiation signals A4.11 -Ability to X

manually operate and/or monitor in the control k':

room: Turbine speed f",

2.1.32 - Conduct of Operations: Ability to X

explain and apply all system limits and precautions.

K4.04 - Knowledge of

' REACTOR FEEDWATER SYSTEM 3.1 34 3.8 35 3.1 36 3.8 37 259001 Reactor X

~ design feature(s) and/or 2.5 38 Feedwater interlocks which provide i

for the following:

Dispersal of feedwater in

. the reactor vessel KIA Category Totals 1

1 1

2 1

1 1

1/1 1

1 1/2 Group Point Total:

1 12/3 Pilgrim 2014 NRC Written Outline.doc

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

Form ES-401-3 Facility:

Pilgrim Date:

2014 NRC Category KA#

Topic RO SRO-Only IR Q#

IR Q#

2.1.20 Ability to interpret and execute 4.6 66 procedure steps.

Ability to locate control room switches, 2.1.31 controls, and indications, and to 4.6 67 determine that they correctly reflect the desired plant lineup.

1. Conduct of Ability to make accurate, clear and Operations 2.1.18 concise logs, records, status boards, 3.6 74 and reports.

2.1.25 Ability to interpret reference materials, 4.2 94 such as graphs, curves, tables, etc.

Subtotal 3

1 2.2.40 Ability to apply technical specifications 3.4 68 for a system.

2.2.6 Knowledge of the process for making 3.0 69 changes to procedures.

Ability to determine the expected plant configuration using design and 2.2.15 configuration control documentation, 3.9 75 such as drawings, line-ups, tag-outs,

2. Equipment etc.

Control 2.2.21 Knowledge of pre-and post-4.1 95 maintenance operability requirements.

Ability to track Technical Specification 2.2.23 limiting conditions for operations.

4.6 98 Subtotal 3

2 Pilgrim 2014 NRC Written Outline.doc

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

Form ES-401-3 2.3.4 Knowledge of radiation exposure limits 3.2 70 under normal or emergency conditions.

Knowledge of radiation monitoring systems, such as fixed radiation 2.3.15 monitors and alarms, portable survey 2.9 71 instruments, personnel monitoring equipment, etc.

3. Radiation Knowledge of Radiological Safety Control Procedures pertaining to licensed operator duties, such as response to 2.3.13 radiation monitor alarms, containment 3.8 96 entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

2.3.11 Ability to control radiation releases.

4.3 99 Subtotal 2

2 2.4.18 Knowledge of the specific bases for EOPs.

3.3 72 2.4.29 Knowledge of the emergency plan.

3.1 73 Knowledge of local auxiliary operator

4. Emergency 2.4.35 tasks during emergency and the 4.0 97 Procedures I resultant operational effects.

Plan Knowledge of RO tasks performed 2.4.34 outside the main control room during an 4.1 100 emergency and the resultant operational effects.

Subtotal 2

2 Tier 3 Point Total:

10 7

Pilgrim 2014 NRC Written Outline.doc

ES-401 Record of Rejected KIA's Form ES-401-4 Tier I Group Randomly Selected Reason for Rejection KA 295016 I AA1.09 1 I 1 replaced by 295016 I Facility does not have an isolation/emergency condenser AA1.05 600000 I AK2.03 It isn't possible to prepare a psychometrically sound question 1 I 1 replaced by 600000 I AK2.01 related to the subject KIA.

295036 I 2.2.22 There are no LCOs or safety limits that are directly impacted by 1 I 2 replaced by 295036 I 2.2.14 Secondary Containment High Sump/Area water level.

259002 I K5.07 2 I 1 replaced by 259002 I Facility does not have turbine driven feed pumps K5.01 215004 I K5.01 2 I 1 replaced by 215004 I Unacceptable overlap with question 23 K5.03 400000 I 2.2.39 There are no one hour Tech Spec action statements associated 211 replaced by 400000 I 2.1.2 with Component Cooling water at Pilgrim 263000 I 2.1.7 2 I 1 replaced by 262002 I Excessive topic overlap. 5 items related to DC power 2.1.7 3 I 1 G1 /2.1.21 replaced It isn't possible to prepare a psychometrically sound question by_G1 /2.1.18 related to the subject KIA.

203000 I K2.01 It isn't possible to prepare a psychometrically sound question 2 I 1 replaced by 203000 I related to the subject KIA. Any question developed directly to K2.02 this KA would be LOD = 1 204000 I 2.4.34 The subject KIA isn't relevant at the subject facility. There are no 2/2 replaced by 204000 I RO tasks performed outside the control room associated with 2.4.12 this system.

211000IK3.02 There is no direct interrelationship between the Standby Liquid 2 I 1 replaced by 211 000 Control System and the Core Spray Line Break Detection K3.01 System at Pilgrim.

272000 I K2.03 NRC feedback indicated that there was an oversampling of 2/2 replaced by 226001 K1.08 direct power supply questions.

295016 I AA 1.05 Originally proposed question was determined to be LOD1 1 I 1 replaced by 295016 following NRC review. Unable to generate another AA1.01 1 psychometrically sound question related to the original KIA.

Originally selected KIA was determined to be applicable to only 3

G.4.23 replaced by SRO candidates following NRC review. It was therefore G2.4.18 determined that it was not possible to develop a I psychometrically sound question for RO candidates.

Originally proposed question was determined to be unsat G2.2.11 replaced by following NRC review. Unable to generate another 3

psychometrically sound question related to the original KIA that G2.2.23 did not overlap the already administered Audit exam. The same KIA was on the Audit Exam.

3 G2.3.5 replaced by Unable to generate a psychometrically sound question related to G2.3.11 the original KIA that is suitable for SRO candidates.

Pilgrim 2014 NRC Written Outline.doc

ES-301 Control Room/In-Plant Systems Outline (REV 12/02/13 Form ES-301-2 Facility:

Pilgrim Date of Examination:

Exam Level (circle

~R0-1 D SRO-U D Operating Test one):

No.:

Control Room Systems@ (8 for RO; 7 for SR0-1; 2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

S1 201003 Control Rod and Drive Mechanism [A2.02 (3.7/3.8)]

S, D,P,A Control Rod Exercise S2 202001 Recirculation System [A4.08 (3.2/3.1 )]

S,D Startup an Idle Recirc Pump with Other Pump Running S3 217000 Reactor Core Isolation Cooling System [A4.03 (3.4/3.3)]

S,N,A,EN,L Manual Start of RCIC S4 241000 ReactorfTurbine Pressure Regulating System [A2.03

( 4.1 /4.2)]

S,D,A, L Establish a Reactor Pressure Band S5 295013 High Suppression Pool Water Temperature [AA1.01 (3.9/3.9)]

S, D,A Place RHR in Torus Cooling S6 264000 Emergency Generators [A3.04 3.1/3.1)]

S,D,A,EN EMERGENCY DIESEL GENERATOR SURVEILLANCE S7 261000 Standby Gas Treatment System [A4.07 (3.1/3.2)]

S,N,A Main Stack Dilution Flow Fan Trip S8 215005 Average Power Range Monitor/Local Power Range Monitor

[A4.04 (3.2/3.2)]

S,N Bypass a Failed LPRM In-Plant Systems@ (3 for RO; 3 for SR0-1; 3 or 2 for SRO-U)

P1 295018 Partial or Complete Loss of Component Cooling Water

[AA 1.01 (3.3/3.4 )]

D,P, R,E Cross-Tie RBCCW Cooling Loops Document in jmd\\AppData\\Locai\\Microsoft\\Windows\\Temporary Internet Files\\Content. Outlook\\RXZYV7PA \\NRC 75-day submittal.doc 2014 Control Room/In-Plant-NRC Exam 2/2014 N14-1 Safety Function 1

4 2

3 5

6 9

7 8

ES-301 Control Room/In-Plant Systems Outline (REV 12/02/13)

Form ES-301-2 P2 212000 Reactor Protection System [A2.01 (3. 7/3.9)]

Transfer of RPS Bus "A" from Alternate Power to the "A" RPS MG D

7 Set P3 295003 Partial or Complete Loss of AC Power [AA 1.03 ( 4.1/4.1 )]

D,E 6

Manual Transfer of B-6 All RO and SR0-1 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 SR0-1 I SRO-U (A)Iternate path 4-6 Required For RO, 6 Actual (C)ontrol room (D)irect from bank

s; 9 Required for RO, 8 Actual (E)mergency or abnormal in-plant

~ 1 Required for RO, 2 Actual (EN)gineered Safety Feature NIA for RO (L)ow-Power I Shutdown

>

1 Required for RO, 2 Actual (N)ew or (M)odified from bank including 1 (A)

~ 2 Required for RO, 3 Actual (P)revious 2 exams

s; 3 Randomly Selected for RO, 2 Actual (R)CA (S)imulator JPM S1 JPM S2 JPM S3

~ 1 Required for RO, 1 Actual JPM Summary This is a Bank JPM previously used on the 2011 NRC Exam. The operator will be told that the plant is at 100% power, and that the Weekly Control Rod exercising is to be performed in accordance with PNPS 8.3.2, Control Rod Exercise. The operator will be directed to commence the weekly control rod exercising in accordance with Section 8.1 of PNPS 8.3.2. When a coupling check is performed on a rod being withdrawn, the rod will go into an overtravel condition (Alternate Path). The operator will be expected to recouple the rod per off-normal procedure 2.4.11, Control Rod Positioning Malfunctions.

This is a Bank JPM. The operator will be told that the "A" Recirc pump tripped two days ago, that the reactor was placed into single loop operation, that the problem has been repaired and post work tested, and that preparations to restart the pump are underway. The operator will be provided with a set of initial plant conditions and a mostly completed OPER-19 (Reactor Recirculation Pump Start Checklist). The operator will be directed to start the "A" recirc pump lAW section 7.4, of PNPS 2.2.84 (Startup of Reactor Recirculation Pump(s) (Reactor Pressurized and/or at Power Conditions)), and to match speeds with the operating pump. The operator will be expected to start the "A" Recirc pump on the first attempt and then match speeds with the "B" Recirc pump pump.

This is a New JPM. The operator will be placed in a situation that requires the manual startup of Reactor Core Isolation Cooling System (RCIC). The operator will be directed to place RCIC in the Injection Mode and raise water level in Document in jmd\\AppData\\Locai\\Microsbft\\Windows\\Temporary Internet Files\\Content.Outlook\\RXZYV7PA\\NRC 75-day subrnittal.doc 2014 Control Room/In-Plant-NRC Exam

ES-301 JPM S4 JPM S5 JPM S6 Control Room/In-Plant Systems Outline (REV 12/02/13}

Form ES-301-2 accordance with PNPS 2.2.22.5, RCIC Injection and Pressure Control,, RCIC Injection. The operator will be expected to use Section 1.0, Injection, of Attachment 1 and discover that the RCIC System Injection Mode pushbutton fails to start the system (Alternate Path).

The operator will be expected to go to Section 2.0, Manual Injection, of Attachment 1 and manually start the RCIC System.

This is a modified Bank JPM. The operator will be told that the the plant is shutdown following a manual Reactor scram, that all control rods are inserted, that the EPR is controlling Reactor pressure at =940 psig, and that another operator is controlling Reactor water level in the normal band. The operator will be directed to take control of Reactor pressure with the Bypass Valve Opening Jack (BVOJ) and establish a pressure band between 800 and 950 psig in the lower part of the band in accordance with Section 4.2, RPV Cooldown and Depressurization Under Emergency Conditions, of PNPS 5.3.35.2, Operations Emergency and Transient Response Strategies. The operator will be expected to use the BVOJ to establish and maintain a Reactor pressure band between 800 to 950 psig.

When this pressure is established a new pressure band will be directed. During the establishment of this reactor pressure one Bypass Valve will fail fully open (Alternate Path). The operator will be expected to respond to this event using Steps 4.3.2 of PNPS 2.4-47, Turbine Control System Malfunctions.

The use of the BVOJ Control Switch and the Vacuum Trip Monitor (Trip#2)

Pushbutton will not close the Bypass Valve, and the operator will be required to close the MSIVs to limit the cooldown.

This is a Bank JPM. The operator will be told that the plant is at 100% power, Torus temperature is >80°F, and that RBCCW cooling has been maximized by another operator. The operator will be directed to maximize Torus cooling by initially placing the "A" RHR loop in torus cooling in accordance with Attachment 14, Maximize Torus Cooling Checklist, of PNPS 5.3.35.1, Transient Response Hardcards for Operating Crews. The operator will be expected to align the RHR System for Torus Cooling. When the "A" RHR Pump is started the operator will observe indications that there is potential blockage of the ECCS Suction Strainers (Alternate Path).

The operator will be expected to address PNPS 2.2.19.5, RHR Modes of Operation for Transients, Attachment 8, Mitigative Actions for Potential ECCS Suction Strainer Blockage, and determine that the indications are associated with blockage of the ECCS Suction Strainers.

This is a Bank JPM. The operator will be told that the reactor is at power with all house loads aligned to the Unit Aux Transformer. The monthly operability run of the "A" EDG is in progress. The Operator parallelthe EDG and commence the EDG operability run.

As KW loading is increased, the EDG governor will become unstable as indicated by KW swings on the engine (Alternative Path).

The operator is expected to recognize indications of unstable operation and lAW the precautions of procedure 8.9.1 unload the engine and open its output breaker.

JPM S7 This is a New JPM. The operator will directed to respond to a C7 Mise Alarm on a Control Room back panel. The operator will discover that the "A" Main Stack Dilution Flow Fan has tripped and diagnose that the "B" Fan failed to start. lAW Document in jmd\\AppData\\Locai\\Microsoft\\Windows\\Temporary Internet Files\\Content.Outlook\\RXZYV7PA\\NRC 75-day submittal.doc 2014 Control Room/In-Plant-NRC Exam

ES-301 JPM S8 JPM P1 JPM P2 JPM P3 Control Room/In-Plant Systems Outline (REV 12/02/13)

Form ES-301-2 the ARP, the operator is expected to start the "B" fan. Shortly thereafter the "B" Fan will also trip and the operator is expected to enter PNPS 2.4.45, Loss/Reduction of Main Stack Dilution Flow at Stack Building (Alternate Path).

The operator will be expected to execute the actions of PNPS 2.4.45, and manually start one train of Standby Gas Treatment in accordance with PNPS 2.2-50, Standby Gas Treatment.

This is a New JPM. The operator will be told that with the plant operating at power an LPRM failure occurred; and that the crew is implementing PNPS 2.4.38, LPRM Failure. The operator will be directed to bypassLPRM 20-29B by performing Attachment 4 of PNPS 2.2.66, Local Power Range Monitoring Systems. The operator will be expected to complete Steps 1-1 0 of Attachment 4 of PNPS 2.2.66.

This is a Bank JPM previously used on the 2009 NRC Exam. The operator will be told that the plant was operating at 100% power when bus A5 locked out due to a ground fault and as a result, the RBCCW loop 'A' pumps have been lost.

The operator will also be told that Off-Normal procedure 2.4.A5, Loss of Electrical Bus A5, has been entered. The operator will be directed to Cross-tie Reactor Building Closed Cooling Water with the 'B' loop supplying, in accordance with PNPS 2.4.42, Loss of RBCCW, Attachment 5, RBCCW Loop Crosstie. The operator will be expected to cross-tie the 'A' and 'B' Reactor Building Closed Cooling Water loops.

This is a Bank JPM. The operator will be told that RPS "A" was transferred to its alternate power supply to support repairs to the "A" RPS MG set, that repairs have been completed and the "A" RPS MG set has been restarted, and that RPS "A" needs to be transferred back to its normal supply.

The operator will be directed to transfer RPS bus "A" back to "A" RPS MG Set in accordance with PNPS 2.2. 79, Reactor Protection System. The operator will be expected to complete Section 7.1.4[1] of PNPS 2.2. 79.

This is a Bank JPM. The operator will be told that with the plant operating at 100% power, a momentary loss of B 1 has occurred, that B-6 failed to transfer to Bus B-2, and that PNPS 2.4.B6, Loss of Bus B6, has been entered.

The operator will be directed to transfer B6 to B1 in accordance with PNPS 2.4.B6.

The operator will be expected to transfer B6 to B1 in accordance with Section 4.2.1 of PNPS 2.4.B6.

Document in jmd\\AppData\\Locai\\Microsoft\\Windows\\Temporary Internet Files\\Content.Outlook\\RXZYV7PA\\NRC 75-day submittal.doc 2014 Control Room/In-Plant-NRC Exam

ES-301 Control Room/In-Plant Systems Outline (REV 12102113 Form ES-301-2 Facility:

Pilgrim Date of Examination:

Exam Level (circle RO D~RO-U D Operating Test one):

No.:

Control Room Systems@ (8 for RO; 7 for SR0-1; 2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

S1 201003 Control Rod and Drive Mechanism [A2.02 (3. 7 /3.8}]

S, D,P,A Control Rod Exercise S2 Not Used for SR0-1 S3 217000 Reactor Core Isolation Cooling System [A4.03 (3.4/3.3}]

S,N,A, EN,L Manual Start of RCIC S4 241000 Reactor/Turbine Pressure Regulating System [A2.03 (4.1/4.2)]

S, D, A, L Establish a Reactor Pressure Band S5 295013 High Suppression Pool Water Temperature [AA1.01 (3.9/3.9}]

S,D,A Place RHR in Torus Cooling S6 264000 Emergency Generators [A3.04 3.1/3.1 )]

S, D,A,EN EMERGENCY DIESEL GENERATOR SURVEILLANCE S7 261000 Standby Gas Treatment System [A4.07 (3.1/3.2}]

S,N,A Main Stack Dilution Flow Fan Trip S8 215005 Average Power Range Monitor/Local Power Range Monitor

[A4.04 (3.2/3.2)]

S,N Bypass a Failed LPRM In-Plant Systems@ (3 for RO; 3 for SR0-1; 3 or 2 for SRO-U)

P1 295018 Partial or Complete Loss of Component Cooling Water

[AA 1.01 (3.3/3.4 )]

D,P, R, E Cross-Tie RBCCW Cooling Loops Document in jmd\\AppData\\Locai\\Microsoft\\Windows\\Temporary Internet Files\\Content.Outlook\\RXZYV?PA\\NRC 75-day submittal.doc 2014 Control Room/In-Plant-NRC Exam 2/2014 N14-1 Safety Function 1

2 3

5 6

9 7

8

ES-301 Control Room/In-Plant Systems Outline (REV 12/02/13 Form ES-301-2 P2 212000 Reactor Protection System [A2.01 (3.7/3.9)]

Transfer of RPS Bus "A" from Alternate Power to the "A" RPS MG D

7 Set P3 295003 Partial or Complete Loss of AC Power [AA 1.03 ( 4.1/4.1 )]

D, E 6

Manual Transfer of B-6 All RO and SR0-1 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 SR0-1 I SRO-U (A)Iternate path 4-6 Required For SR0-1, 6 Actual (C)ontrol room (D)irect from bank

8 Required for SR0-1, 7 Actual (E)mergency or abnormal in-plant

1 Required for SR0-1, 2 Actual (EN)gineered Safety Feature NIA for SR0-1 (L)ow-Power I Shutdown

1 Required for SR0-1, 2 Actual (N)ew or (M)odified from bank including 1 (A)

2 Required for SR0-1, 3 Actual (P)revious 2 exams

3 Randomly Selected for SR0-1, 2 Actual (R)CA (S )imulator JPM S1 JPM S2 JPM S3

1 Required for SR0-1, 1 Actual JPM Summary This is a Bank JPM previously used on the 2011 NRC Exam. The operator will be told that the plant is at 1 00% power, and that the Weekly Control Rod exercising is to be performed in accordance with PNPS 8.3.2, Control Rod Exercise. The operator will be directed to commence the weekly control rod exercising in accordance with Section 8.1 of PNPS 8.3.2. When a coupling check is performed on a rod being withdrawn, the rod will go into an overtravel condition (Alternate Path). The operator will be expected to recouple the rod per off-normal procedure 2.4.11, Control Rod Positioning Malfunctions.

Not Used For SR0-1 This is a New JPM. The operator will be placed in a situation that requires the manual startup of Reactor Core Isolation Cooling System (RCIC). The operator will be directed to place RCIC in the Injection Mode and raise water level in accordance with PNPS 2.2.22.5, RCIC Injection and Pressure Control,, RCIC Injection. The operator will be expected to use Section 1.0, Injection, of Attachment 1 and discover that the RCIC System Injection Mode pushbutton fails to start the system (Alternate Path).

The operator will be expected to go to Section 2.0, Manual Injection, of Attachment 1 and manually start the RCIC System.

JPM S4 This is a modified Bank JPM.

The operator will be told that the the plant is shutdown following a manual Reactor scram, that all control rods are inserted, Document in jmd\\AppData\\Locai\\Microsoft\\Windows\\Temporary Internet Files\\Content.Outlook\\RXZYV7PA\\NRC 75-day submittal.doc 2014 Control Room/In-Plant-NRC Exam

ES-301 JPM S5 JPM S6 JPM S7 Control Room/In-Plant Systems Outline (REV 12/02/13 Form ES-301-2 that the EPR is controlling Reactor pressure at :::::940 psig, and that another operator is controlling Reactor water level in the normal band.

The operator will be directed to take control of Reactor pressure with the Bypass Valve Opening Jack (BVOJ) and establish a pressure band between 800 and 950 psig in the lower part of the band in accordance with Section 4.2, RPV Cooldown and Depressurization Under Emergency Conditions, of PNPS 5.3.35.2, Operations Emergency and Transient Response Strategies. The operator will be expected to use the BVOJ to establish and maintain a Reactor pressure band between 800 to 950 psig.