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

Facility: Salem Date of Examination: 1/22/2018 Examination Level: RO ~ SRO D Operating Test Number: 16-01

Administrative Topic (see Type Describe activity to be performed Note) Code*

Perform a manual AFD calculation IAW S2.0P-ST.NIS-0001

Conduct of Operations R,D [2.1.25, R0-3.9; Ability to interpret reference materials, such as graphs, curves, tables, etc.]

Conduct of Operations Perform control room log readings for 21-24 reactor coolant loops OTDT

R,N channel check IAW S2.0P-DL.ZZ-0003

[2.1.18, R0-3.6; Ability to make accurate, clear, and concise logs, records, status boards, and reportsl

Calculate a manual OPTR !AW S2.0P-ST.NIS-0002 ·

Equipment Control R,D [2.2.12, R0-3.7; Knowledge of surveillance procedures}

NIA

Radiation Control

Activate EROS during an ALERT !AW EP-SA-111-FS

Emergency Plan S,P [2.4.43 R0-3.2; Knowledge of emergency communication systems and techniques}

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

* Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(:, 3 for ROs; ::; 4 for SR Os and RO retakes) (N)ew or (M)odified from bank (2: 1) (P)revious 2 exams(:, 1, randomly selected)

ES-301 > · AdministrativeTopics Outline Forrh ES-301-1

Facility: Salem Date of Examination: 1/22/2018 Examination Level: RO D SRO 0 Operating Test Number: 16-01

Administrative Topic (see Type Describe activity to be performed Note) Code*

Review and approve the Non-Essential heat load requirements for

Conduct of Operations inoperable chiller(s) IAW S2.0P-SO.CH-0001

R,D [2.1.25, SR0-4.2: Ability to interpret reference materials, such as graphs, curves, tables, etc.]

Determine the Heat Stress requirements for two operators entering Containment JAW SA-AA-111

Conduct of Operations R,N [2.1.26, SR0-3.6: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxyqen and hydroqen)l

Review and approve a safety tagging request JAW OP-AA-109 and

Equipment Control R,D identify applicable Technical Specifications required

[2.2.13, SR0-4.3: Knowledge of tagging and clearance procedures]

Authorize a Radioactive Gas Release

Radiation Control R,D [2.3.6, SR0-3.8: Ability to approve a release permit]

Classify Emergency Event and complete the ICMF IAW EP-SA-111-

Emergency Pian 101 (Time Critical)

R,M .[2.4.41, SR0-4.6: Knowledge of the emergency action level thresholds and classifications]

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

* Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(::: 3 for ROs; s 4 for SR Os and RO retakes) (N)ew or (M)odified from bank(~ 1) (P)revious 2 exams(::: 1, randomly selected)

ES-301 Control Room/In-Plant Systems Outline FormES-301~2

Facility: Salem Date of Examination: 1/22/2018 Exam Level: RO lZl SRO-I D SRO-U D Operating Test Number: 16-01

. Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U

System/JPM Title Type Code* Safety

(KIA, Importance RO/SRO) Function

a. [001] TCAF a Loss of Control Rod Drive Vent Fan(s) (A2.01, 3.1/3.7) A,N,S 1

b. [004] Perform manual Makeup of the VCT JAW S2.0P-SO.CVC-0006 (A4.13, N,S 2 3.3/2.9)

c. [006] Isolate ECCS Accumulators JAW 2-EOP-TRIP-6 (A 1.13, 3.5/3.7) A,D,L,P,S 3

d. [002] Initiate Bleed and Feed using reactor head vents JAW 2-EOP-FRHS-1 A,D,E,L,P,S 4-Pri (A2.04, 4.3/4.6)

e. [045] Synchronize the main generator JAW S2.0P-SO.TRB-0001 (A4.02, D,S 4-Sec 2.7/2.6)

f. [022] Perform 22 CFCU surveillance test JAW S2.0P-ST.CBV-0003 (A1.04, D,EN,S 5 3.2/3.3)

g. [062] Transfer 4 KV Group Buses from SPT to APT JAW S2.0P-S0.4KV- A,D,P,S 6 0008 (A2.04, 3.1/3.4)

h. [008] Start 23 CCW Pump JAW 2-EOP-APPX-1 (A2.01, 3.3/3.6) A,E,EN,L,M,S 8

. In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. [063 / 055] Shed non-essential 125V DC loads JAW 2-EOP-LOPA-1, Checkoff D,E 6 Sheet 3 (A4.01, 2.8/3.1; EA 1.04, 3.5/3.9)

j. [012] Locally open ReactorTrip and Bypass Breakers JAW S1.0P-AB.CR- D,E,L,R 7 0001 Attachment 5 (A4.06, 4.3/4.3)

I

k. [068] Perform a radioactive liquid release and TCAF high radiation monitor D,R 9 alarm IAW S2.0P-SO.WL-0001 (A2.04, 3.3/3.3)

* All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

* Type Codes I Criteria for R /SR0-1/SRO-U

ES-301 Control Roomfln.:PlantSystems Outline Form,ES-301-2

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (R)CA (S)imulator

4-6/4--6 /2-3

::: 9/:5 8/::: 4 2:1/2:1/2:1 ;;: 1/~ 1/2: 1 (control room system) ;;: 1/2: 1/2: 1 ;;: 2/2: 2/2: 1 ::: 3/::: 3/::: 2 (randomly selected) ;;: 1/2: 1/2: 1

. ES-30.1 Control Room/In-Plant Systems 'Outline Form ES~301-2

Facility: Salem Date of Examination: 1/22/2018 Exam Level: RO D SRO-I [:g] SRO-U D Operating Test Number: ·16-01

* Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U

System/JPM Title Type Code* Safety

(KIA, Importance RO/SRO) Function

a. [001] TCAF a Loss of Control Rod Drive Vent Fan(s) (A2.01, 3.1/3.7) A,N,S 1

b. [004] Perform manual Makeup of the VCT IAW S2.0P-SO.CVC-0006 (A4.13, N,S 2 3.3/2.9)

c. [006] Isolate ECCS Accumulators IAW 2-EOP-TRIP-6 (A 1.13, 3.5/3.7) A,D,L,P,S 3

d. [002] Initiate Bleed and Feed using reactor head vents IAW 2-EOP-FRHS-1 A,D,E,L,P,S 4-Pri (A2.04, 4.3/4.6)

e. N/A

f. [022] Perform 22 CFCU surveillance test IAW S2.0P-ST.CBV-0003 (A 1.04, D,EN,S 5 3.2/3.3)

g. [062] Transfer 4 KV Group Buses from SPT to APT IAW S2.0P-S0.4KV- A,D,P,S 6 0008 (A2.04, 3.1/3.4)

h. [008] Start 23 CCW Pump IAW 2-EOP-APPX-1 (A2.01, 3.3/3.6) A,E,EN,L,M,S 8

* In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. [063 / 055] Shed non-essential 125V DC loads IAW 2-EOP-LOPA-1, Checkoff D,E 6 Sheet 3 (A4.01, 2.8/3.1; EA1 .04, 3.5/3.9)

j. [012] Locally open Reactor Trip and Bypass Breakers IAW S1.0P-AB.CR- D,E,L,R 7 0001 Attachment 5 (A4.06, 4.3/4.3)

k. [068] Perform a radioactive liquid release and TCAF high radiation monitor D,R 9 alarm IAW S2.0P-SO.WL-0001 (A2.04, 3.3/3.3)

* All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

I * Type Codes I Criteria for R /SR0-1/SRO-U 11

ES-.301. Control Room/In-Plant Sy$tems OuUine Form ES-301-2

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator

4-6/4-6 /2-3

:5 9/:5 8/:5 4 ?.1/?.1/?.1 ?. 1/?. 1/?. 1 (control room system) ?. 1 /?. 1 /?. 1 ?. 2/?. 2/?. 1 :5 3/:5 3/:5 2 (randomly selected) ?.1/?.1/?.1

\;• ! t;: i '-~ ' . Es,-3:01 Control Room/In-Plant Systems Outline Forrri'ES-301-2

Facility: Salem Date of Examination: 1/22/2018 Exam Level: RO D SRO-I D SRO-U 13:1 Operating Test Number: 16-01

. Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U

System/JPM Title Type Code* Safety

(KIA, Importance RO/SRO) Function

a. NIA

b. NIA

c. [006] Isolate ECCS Accumulators IAW 2-EOP-TRIP-6 (A1.13, 3.5/3.7) A,D,L,P,S 3

d. N/A

e. N/A

f. [022] Perform 22 CFCU surveillance test !AW S2.0P-ST.CBV-0003 (A1 .04, D,EN,S 5 3.2/3.3)

g. N/A

h. [008] Start 23 CCW Pump IAW 2-EOP-APPX-1 (A2.01, 3.3/3.6) A,E,EN,L,M,S 8

. In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. [063 / 055] Shed non-essential 125V DC loads !AW 2-EOP-LOPA-1, Checkoff D,E 6 Sheet 3 (A4.01, 2.8/3.1; EA 1.04, 3.5/3.9)

j. [012] Locally open Reactor Trip and Bypass Breakers !AW S1 .OP-AB.CR- D,E,L,R 7 0001 Attachment 5 (A4.06, 4.3/4.3)

k. N/A

* All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

I * Type Codes I Criteria for R /SR0-1/SRO-U I (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank ::; 9/:5 8/:, 4 (E)mergency or abnormal in-plant 2: 1/2: 1/2: 1 (EN)gineered safety feature 2: 1/2: 1/2: 1 (control room system) {L )ow-Power/Shutdown 2: 1/2: 1/2: 1 (N)ew or (M)odified from bank including 1 (A) 2: 2/2: 2/2: 1 (P)revious 2 exams ::; 3/:5 3/:5 2 (randomly selected) (R)CA 2: 1/2: 1/2: 1 (S)imulator

AppendixD. Scenario Outline Form l;S-D-1

Facility: __ Salem ___ Scenario No.: _1 (ESG-1) __ Op-Test No.: _16-01_

Examiners: Operators:

Initial Conditions: IC-233; 4% power, BOL; 21 Charging Pump is running, 21 SGFP is in service, power ascension in progress at 10%/hr.

Turnover: The following equipment is out of service: 23 CV Pump is CIT to repair the fluid drive linkage. Crew is directed to continue raising Rx power to 10% and enter Mode 1 using Main Steam Dumps and Control Rods.

Critical Tasks:

1. Stop 22 RCP and close 22CV104 valve within 3-5 minutes after stopping the RCP (site specific CT).

2. Control initial RCS cooldown so that transition from EOP-SGTR-1 does not occur (see WOG CT-19).

Event Malf. No. Event Event No. Type* Description

1 N/A ALL (R) power ascension at 10% per hour

2 PR0016A RO (I) PZR pressure controlling channel fails high SRO (1,TS)

3 SW0216A ALL (C) #2 SW bay leak (ramped over 2 minutes) SRO (C,TS)

4 RC007B RO (C) 22 RCP #1 seal degradation (ramped over 1 minute) SRO (C)

5 RCOO?B ALL (M) after the crew addresses the degraded 22 RCP seal; the #1 seal will completely fail requiring a reactor trip

6 SG0078D ALL (C) 24 SG tube rupture will occur following transition into EOP-TRIP-2; crew will initiate SI and re-enter EOP-TRIP-1

7 VL0026 PO (C) 24BF13 fails to close on a SI/ Phase A signal SRO (C)

ABs AB.PZR-1 --> AB.SW-3--> AB.RCP-1

EOPs TRIP-1 --> TRIP-2--> TRIP-1 --> SGTR-1

* (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Appen~Hx D S_cenario O_utline Form ES-0-1

Scenario No.: 1 (ESG-1)

Target Quantitative Attributes per Scenario (See Section D.5.d) Actual Attributes Event No.

1. Total malfunctions (5-8) 6 -2. Malfunctions after EOP entry (1-2) 2 6,7

3. Abnormal events (2-4) 3 2,3,4

4. Major transients (1-2) 1 5 TRIP-1

5. EOPs entered/requiring substantive actions ( 1-2) 3 TRIP-2 SGTR-1

6. Entry into a contingency EOP with substantive actions ( ;;: 1 per 0 scenario set)

-

7. Preidentified critical tasks (:::2) 2 Site CT, CT-19

8. Tech Specs exercised(;;: 2) 2 2,3

Appendix D Scenario Outline Form ES-0~1

Facility: --Salem --- Scenario No.: _2 (ESG-2) __ Op-Test No.: _16-01_

Examiners: Operators:

Initial Conditions: IC-231; 40% power, BOL; 23 Charging Pump is running, power ascension is on hold due to turbine valve testing issues.

Turnover: The following equipment is out of service: 23 Condensate pump and all Heater Drain pumps due to power level, 22 SW Pump CIT for pump packing replacement, 238 CW pump is CIT for traveling screen repair, 2A EOG is CIT for governor inspection and output breaker swap.

Critical Tasks:

1. Manually actuate SI before transition to EOPs; LOCA-1, FRTS-1, or TRIP-2 (see WOG CT-2).

2. Transfer to cold leg recirculation before RWST level drops below 1.2 feet (see WOG CT-36).

Event Malf. No. I Event Event No.

1 Type* Description

1 VL0420 PO (I) 21MS167 drifts from full open position SRO (I)

2 RC0014B RO (I) 22 RC Loop Tavg Channel fails high SRO (!,TS)

3 RC0002 RO (C) RCS leak inside containment (35 gpm ramped over 2 minutes) SRO (C,TS)

4 N/A ALL (R) commence unit shutdown as a result RCS leakage exceeding TS limits

after shutdown has commenced, 24 RC loop line breaks inside 5 RC0001D ALL (M) containment (Large Break LOCA) requiring a reactor trip and safety

injection

RP0274A both trains of SI fails to auto actuate and manual key switch operation 6 RP0274B RO (I) will fail to actuate on one train; the operator will be able to manually A501 SRO (I) actuate SI using the other train key switch. A701

7 DG02D ALL(!) 28 SEC fails to actuate on SI signal resulting in operators manually starting safeguards loads for 28 vital bus

ABs AB.ROD-3--> AB.RC-1 --> AB.LOAD-1

EOPs TRIP-1 --> LOCA-1 --> LOCA-3

* (N)ormal, (R)eactivity, ( I )nstrument, (C)omponent, (M)ajor

. Appendix:D . . Scenario Outline Form ES-D-1

Scenario No.: 2 (ESG-2)

Target Quantitative Attributes per Scenario (See Section D.5.d) Actual Attributes Event No.

1. Total malfunctions (5-8) 6 -2. Malfunctions after EOP entry (1-2) 2 6,7

3. Abnormal events (2-4) 3 1,2,3

4. Major transients (1-2) 1 5 TRIP-1

5. EOPs entered/requiring substantive actions (1-2) 3 LOCA-1 LOCA-3

6. Entry into a contingency EOP with substantive actions (.:: 1 per 0 scenario set) -

7. Pre identified critical tasks (.::2) 2 CT-2, CT-36

8. Tech Specs exercised(.:: 2) 2 2,3

, __ Appendix D Scenario Outline Form ES-0-1

Facility: __ Salem ___ Scenario No.: _ 3(ESG-5)_ Op-Test No.: _16-01 _

Examiners: Operators:

Initial Conditions: IC-232; 88% power, MOL; 23 Charging Pump is running.

Turnover: The following equipment is out of service: 21 Safety Injection (SI) Pump CIT for motor lube, 21 AFW Pump CIT for motor replacement. Downpower to 88% just completed to support Main Turbine valve testing later in the shift.

Critical Tasks:

1. Close PZR PORV block valve by completion of Step 9 in EOP-TRIP-2 that directs closing of PZR block valves (see WOG CT-10).

2. Establish condensate flow to SGs before bleed and feed is required [2 SG WR levels < 32 % or 37% Adverse] (see WOG CT-43).

Event Malf. No. Event Type* Event No. Description

1 PR0017A RO (I) PZR level controlling channel fails high SRO (!,TS)

2 VC0173B RO(C) 22 CFCU Trips SRO (C,TS)

3 BF0105A ALL (C) 21 SGFP trips

4 EH0327 PO (I) failure of Turbine Runback circuit following SGFP trip; operator can SRO (I) manually initiate load reduction.

5 N/A ALL (R) turbine load reduction to 66% as directed by AB.CN-0001

6 BF0105B ALL (M) 22 SGFP trips requiring manual reactor trip

7 AF0181B PO (C) 22 AFW pump trips following reactor trip SRO (C)

8 VL0297 RO(C) after transition into EOP-TRIP-2, PZR PORV 2PR1 valve fails SRO (C) partially open; operator can manually close PZR PORV block valve

9 AF0183 PO (C) after 2PR1 issue is addressed; 23 AFW pump trips resulting in no SRO (C) AFW flow to any SGs

ABs AB.CVC-1 -> AB.CN-1

EOPs TRIP-1 -> TRIP-2-> FRHS-1 -> TRIP-2

* (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Appendix 0 .. . :. :. Scenario Outline · Form ES-D-1.

Scenario No.: 3 (ESG-5)

Target Quantitative Attributes per Scenario (See Section D.5.d) Actual Attributes Event No.

1. Total malfunctions (5-8) 7 -2. Malfunctions after EOP entry (1-2) 3 7,8,9

3. Abnormal events (2-4) 3 1,3,4

4. Major transients (1-2) 1 6

5. EOPs entered/requiring substantive actions (1-2) 1 TRIP-2 6. Entry into a contingency EOP with substantive actions ( ::: 1 per 1 FRHS-1 scenario set) 7. Preidentified critical tasks (:::2) 2 CT-10, CT-43

8. Tech Specs exercised(::: 2) 2 1,2

ES-401 PWR Examination Outline Form ES-401-2

II - 'LOT 1601 SRO Date of Exam: January 2018

RO K/A Category Points SRO-Only Points

Tier Group K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 18 3 3 6 Emergency and Abnormal Plant 2 N/A N/A 9 2 2 4

Evolutions Tier Totals 27 5 5 10

1 28 3 2 5 2.

10 Plant 2 2 1 3 Systems

Tier Totals 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories

2 2 1 2

Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

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 outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

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

5. Absent a plant-specific priority, only those K/As 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 K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' I Rs 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 a category 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 K/A catalog and enter the K/A numbers, descriptions, I Rs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

* These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

** These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2 Form ES-401-2

IIES-401 PWR Examination Outline F_

Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (SRO)

E/APE # / Name/ Safety Function K1 K2 K3 A1 A2 G* KIA Topic(s} IR #

000007 (EPE 7; BW E02&E10; CE E02) Reactor Trip, Stabilization, Recovery/ 1

000008 (APE 8) Pressurizer Vapor Space X AA2.29 4.2 1 Accident/ 3

000009 (EPE 9} Small Break LOCA / 3

000011 (EPE 11) Laroe Break LOCA / 3 X G2.1.2 4.4 2

000015 (APE 15) Reactor Coolant Pump X AA2.10 3.7 4 Malfunctions / 4

000022 (APE 22) Loss of Reactor Coolant Makeup/ 2

000025 (APE 25) Loss of Residual Heat Removal Svstem / 4

000026 (APE 26) Loss of Component Coolino Water/ 8

000027 (APE 27) Pressurizer Pressure Control Svstem Malfunction / 3

000029 (EPE 29) Anticipated Transient Without Scram / 1

000038 (EPE 38) Steam Generator Tube Rupture/ 3

000040 (APE 40; BW E05; CE E05; W E12) Steam Line Rupture-Excessive Heat Transfer I 4

000054 (APE 54; CE E06) Loss of Main X G2.1.6 4.8 3 Feedwater /4

000055 (EPE 55) Station Blackout/ 6

000056 (APE 56} Loss of Offsite Power/ 6

000057 (APE 57) Loss of Vital AC Instrument Bus/ 6

000058 (APE 58) Loss of DC Power/ 6

000062 (APE 62) Loss of Nuclear Service Water I 4

000065 (APE 65) Loss of Instrument Air/ 8

000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6

W E04) LOCA Outside Containment/ 3

(W E 11) Loss of Emergency Coolant X G2.1.9 4.5 6 Recirculation / 4

(BW E04; W E05) Inadequate Heat X EA2.1 4.4 5 Transfer-Loss of Secondarv Heat Sink/ 4

KIA Category Totals: 3 3 Group Point Total: 6

ES-401 3 Form ES-401-2

ES-401 PWR Examination Outline Form ES-401-2 Emerqencv and Abnormal Plant Evolutions-Tier 1/Group 2 (SRO)

E/APE # / Name I Safety Function K1 K2 K3 A1 A2 G* KIA Topic(s) IR #

000001 (APE 1) Continuous Rod Withdrawal / 1

000003 (APE 3) Dropped Control Rod / 1

000005 (APE 5) Inoperable/Stuck Control Rod/ 1

000024 (APE 24) Emergency Boration / 1

000028 (APE 28) Pressurizer (PZR) Level Control X AA2.12 3.5 7 Malfunction / 2

000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7

000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation/ 7

000036 (APE 36; BW/A08) Fuel-Handling Incidents/ 8

000037 (APE 37) Steam Generator Tube Leak / 3

000051 (APE 51) Loss of Condenser Vacuum / 4

000059 (APE 59) Accidental Liquid Radwaste Release/ 9

000060 (APE 60) Accidental Gaseous Radwaste Release/ 9 X AA2.04 3.4 8

000061 (APE 61) Area Radiation Monitoring System Alarms /7

000067 (APE 67) Plant Fire On Site / 8

000068 (APE 68; BW A06) Control Room Evacuation / 8

000069 (APE 69; W E14) Loss of Containment Integrity/ 5 X G2.4.6 4.7 9

000074 (EPE 74; W E06 & E07) Inadequate Core Cooling I 4

000076 (APE 76) High Reactor Coolant Activity/ 9

000078 (APE 78*) RCS Leak / 3

(W E01 & E02) Rediagnosis & SI Termination/ 3

(W E13) Steam Generator Overpressure/ 4

(W E15) Containment Flooding/ 5

(W E16) High Containment Radiation /9 X G2.4.1 4.7 10

(BW A01) Plant Runback / 1

(BW A02 & A03) Loss of NNI-X/Y/7

(BW A04) Turbine Trip/ 4

(BW A05) Emergency Diesel Actuation/ 6

(BW A07) Flooding / 8

(BW E03) Inadequate Subcooling Margin / 4

(BW E08; W E03) LOCA Cooldown-Depressurization / 4

(BW E09; CE A13**; W E09 & E10) Natural Circulation/4

(BW E13 & E14) EOP Rules and Enclosures

(CE A11**; W E08) RCS Overcooling-Pressurized Thermal Shock/ 4

(CE A16) Excess RCS Leakage/ 2

(CE E09) Functional Recovery

(CE E13* Loss of Forced Circulation/LOOP/Blackout/ 4

: 2 2 Group Point Total:

ES-401 4 Form ES-401-2

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems-Tier 2/Group 1 (SRO)

System # I Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* KIA Topic(s) IR #

003 (SF4P RCP) Reactor Coolant Pump

004 (SF1; SF2 CVCS) Chemical and Volume Control

005 (SF4P RHR) Residual Heat X AA2.02 3.8 12

Removal

006 (SF2; SF3 ECCS) Emergency X AA2.10 3.9 11

Core Coolina

007 (SF5 PRTS) Pressurizer Relief/Quench Tank

008 (SF8 CCW) Component Cooling Water

010 (SF3 PZR PCS) Pressurizer Pressure Control

012 /SF? RPS) Reactor Protection

013 (SF2 ESFAS) Engineered Safety Features Actuation

022 (SF5 CCS) Containment CoolinQ

025 (SF5 ICE) Ice Condenser

026 (SF5 CSS) Containment Spray X G2.2.40 4.7 13

039 (SF4S MSS) Main and Reheat Steam

059 (SF4S MFW) Main Feedwater

061 (SF4S AFW) Auxiliarv/Emeroencv Feedwater

062 (SF6 ED AC) AC Electrical Distribution

063 (SF6 ED DC) DC Electrical Distribution

064 (SF6 EDG) Emergency Diesel Generator

073 (SF? PRM) Process Radiation X A2.02 3.2 14

Monitorinc:i

076 (SF4S SW) Service Water

078 (SF8 IAS) Instrument Air X G2.2.15 4.3 15

103 {SF5 CNT) Containment

053 (SF1; SF4P ICS*) Integrated Control

. - ..... - 3 2 Group Point Total: 5

ES-401 5 Form ES-401-2

ES-401 PWR Examination Outline Form ES-401-2

Plant Systems-Tier 2/Grou J 2 (SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* KIA Topic(s) IR #

001 (SF1 CRDS) Control Rod Drive

002 (SF2; SF4P RCS) Reactor Coolant

011 (SF2 PZR LCS) Pressurizer Level Control

014 (SF1 RPI) Rod Position Indication

015 (SF? NI) Nuclear Instrumentation

016 (SF? NNI) Nonnuclear Instrumentation

017 (SF7 ITM) In-Core Temperature Monitor

027 (SF5 CIRS) Containment Iodine Removal

028 (SF5 HRPS) Hydrogen Recombiner and Purge Control

029 (SF8 CPS) Containment Purge

033 (SF8 SFPCS) Spent Fuel Pool X G.2.1.20 4.6 17 Cooling

034 (SF8 FHS) Fuel-Handling Equipment

035 (SF 4P SG) Steam Generator

041 (SF4S SDS) Steam Dump/Turbine Bypass Control

045 (SF 4S MTG) Main Turbine X A2.12 2.8 16 Generator

055 (SF4S CARS) Condenser Air Removal

056 (SF4S CDS) Condensate

068 (SF9 LRS) Liauid Radwaste

071 (SF9 WGS) Waste Gas Disposal

072 (SF? ARM) Area Radiation Monitoring

075 (SF8 CW) Circulating Water

079 (SF8 SAS**) Station Air X A2.01 3.2 18

086 Fire Protection

050 (SF 9 CRV*) Control Room Ventilation

IK/A Cate2o!i'. Point Totals: I I I I I I I I 2 I I I 1 I Group Point Total: I I 3 I

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

Facility: Salem ILOT 16-01 Date of Exam: January 2018

Category KIA# Topic RO SRO-only

IR # IR #

2.1.4 Knowledge of individual licensed operator responsibilities related 3.8 19 to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55, etc

2.1.7 Ability to evaluate plant performance and make operational 4.7 20 judgments based on operating characteristics, reactor behavior,

1. Conduct of and instrument interpretation

Operations 2.1.

2.1.

2.1.

2.1.

Subtotal 2

2.2.25 Knowledge of the bases in Technical Specifications for limiting 4.5 21 conditions for operations and safety limits

2.2.39 Knowledge of less than or equal to one hour Technical 4.5 22 Soecification action statements for svstems

2. Equipment 2.2. Control

2.2.

2.2.

2.2.

Subtotal 2

2.3.6 Ability to aoorove release permits 3.8 24

2.3.

2.3. 3. Radiation Control 2.3.

2.3.

2.3.

Subtotal 1

2.4.21 Knowledge of the parameters and logic used to assess the 4.6 23 status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

4. 2.4.28 Knowledge of procedures relating to a security event (non- 4.1 25

Emergency safeguards information).

Procedures/ 2.4. Plan

2.4.

2.4.

2.4.

Subtotal 2

Tier 3 Point Total 10 7

ES-401 PWR Examination Outline Form ES-401-2

Facility: Salem ILOT 1601 FINAL RO Date of Exam: January 2018

RO KIA Category Points SRO-Only Points

Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 3 3 3 3 3 3 18 6 Emergency and Abnormal Plant 2 2 2 1 N/A 2 1 N/A 1 9 4

Evolutions Tier Totals 5 5 4 5 4 4 27 10

1 3 2 3 3 2 2 3 3 2 2 3 28 5 2.

1 1 1 1 0 1 1 1 1 1 1 10 Plant 2 3 Systems

Tier Totals 4 3 4 4 2 3 4 4 3 3 4 38 8

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 7 Categories

10 3 2 3 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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the "Tier Totals" in each KIA category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it is replaced by a KIA from another Tier 3 category.)

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 outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.

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

5. Absent a plant-specific priority, only those KIAs 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 selected from Section 2 of the KIA catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIAs.

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' I Rs 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 a category 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 Kl A catalog and enter the Kl A numbers, descriptions, I Rs, and point totals (#) on Form ES-401-3. Limit SRO selections to Kl As that are linked to 10 CFR 55.43.

G* Generic KIAs

* These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the KIA catalog.

** These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan.

ES-401 2 Form ES-401-2

ES-401 PWR Examination Outline F~ Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO)

E/APE # / Name I Safety Function K1 K2 K3 A1 A2 G* KIA Tooic(s) IR #

000007 (EPE 7; BW E02&E10; CE E02) X EK2.02 2.6 49

Reactor Trio, Stabilization, Recovery / 1

000008 (APE 8) Pressurizer Vapor Space Accident/ 3

000009 IEPE 9) Small Break LOCA / 3

000011 IEPE 11) Large Break LOCA / 3 X EK1.01 4.1 50

000015 (APE 15) Reactor Coolant Pump X G2.2.42 3.9 51

Malfunctions / 4

000022 (APE 22) Loss of Reactor Coolant X AA2.01 3.2 52

Makeuo / 2

000025 (APE 25) Loss of Residual Heat X AK3.01 3.1 53

Removal Svstem / 4

000026 (APE 26) Loss of Component X AA2.01 3.1 54

Coolino Water/ 8

000027 (APE 27) Pressurizer Pressure X AK2.03 2.6 55

Control Svstem Malfunction / 3

000029 (EPE 29) Anticipated Transient X EA1.01 3.4 56

Without Scram / 1

000038 (EPE 38) Steam Generator Tube X EK3.09 4.1 57

Rupture/ 3

000040 (APE 40; BW E05; CE E05; W E12) X EK1.1 3.4 66

Steam Line Rupture-Excessive Heat Transfer 14

000054 (APE 54; CE E06) Loss of Main X G2.4.49 3.8 58

Feedwater /4

000055 IEPE 55) Station Blackout/ 6 X EK3.02 4.3 59

000056 (APE 56) Loss of Offsite Power I 6 X AA2.54 2.9 60

000057 (APE 57) Loss of Vital AC Instrument Bus / 6

000058 (APE 58) Loss of DC Power/ 6

000062 (APE 62) Loss of Nuclear Service Water 14

000065 (APE 65) Loss of Instrument Air/ 8 X G2.4.47 4.2 61

000077 (APE 77) Generator Voltage and X AA1.02 3.8 62

Electric Grid Disturbances / 6

W E04\ LOCA Outside Containment/ 3 X EA1.2 3.6 63

(W E 11) Loss of Emergency Coolant X EK1.1 3.7 65

Recirculation / 4

(BW E04; W E05) Inadequate Heat X EK2.2 3.9 64

Transfer-Loss of Secondary Heat Sink/ 4

.. - - - . : 3 3 3 3 3 3 Group Point Total: 18

ES-401 3 Form ES-401-2

ES-401 PWR Examination Outline Form ES-401-2

Emergency and Abnormal Plant Evolutions-Tier 1/Group 2 (RO}

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* KIA Topic(s) IR #

000001 (APE 1) Continuous Rod Withdrawal / 1

000003 (APE 3) Dropped Control Rod / 1

000005 (APE 5) Inoperable/Stuck Control Rod / 1

000024 (APE 24) Emergency Boration / 1 X G2.4.34 4.2 67

000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2

000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7

000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7

000036 (APE 36; BW/A08) Fuel-Handling Incidents/ 8

000037 (APE 37) Steam Generator Tube Leak / 3 X AA1.07 3.1 68

000051 (APE 51) Loss of Condenser Vacuum / 4

000059 (APE 59) Accidental Liquid Radwaste Release / 9

000060 (APE 60) Accidental Gaseous Radwaste Release / 9

000061 (APE 61) Area Radiation Monitoring System Alarms /7

000067 (APE 67) Plant Fire On Site / 8

000068 (APE 68; BW A06) Control Room Evacuation I 8

000069 (APE 69; W E14) Loss of Containment Integrity/ 5 X EK3.2 3.1 73

000074 (EPE 74; W E06 & E07) Inadequate Core Cooling I X EK1.3 3.7 71 4

000076 (APE 76) High Reactor Coolant Activity/ 9 X AA2.02 2.8 69

000078 (APE 78*) RCS Leak / 3

(W E01 & E02) Rediagnosis & SI Termination/ 3

(W E13) Steam Generator Overpressure/ 4 X EK2.2 3.0 72

(W E15) Containment Flooding/ 5 X EA1.2 2.7 74

(W E16) High Containment Radiation /9 X EK1.2 2.7 75

(BW A01) Plant Runback / 1

(BW A02 & A03) Loss of NNI-X/Y/7

(BW A04) Turbine Trip/ 4

(BW A05) Emergency Diesel Actuation / 6

(BW A07) Flooding / 8

(BW E03) Inadequate Subcooling Margin / 4

(BW E08; W E03) LOCA Cooldown-Depressurization / 4 X EK2.1 3.6 70

(BW E09; CE A13**; W E09 & E10) Natural Circulation/4

(BW E13 & E14) EOP Rules and Enclosures

(CE A 11 **; W E08) RCS Overcooling-Pressurized Thermal Shock/ 4

(CE A 16) Excess RCS Leakage / 2

(CE E09) Functional Recovery

(CE E13*) Loss of Forced Circulation/LOOP/Blackout/ 4

IK/A Category Point Totals: I 2 I 2 I 1 I 2 I 1 I 1 I Group Point Total: I I 9 I

ES-401 4 Form ES-401-2

IES-401 PWR Examination Outline F-.. , -- ..... -

Plant Svstems Tier 2/Grouo 1 (RO\

Svstem # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* KIA Tooicls) IR #

003 (SF4P RCP) Reactor Coolant X K3.01 3.7 1

Pumo

004 (SF1; SF2 CVCS) Chemical and X K6.20 2.5 2

Volume Control

005 (SF4P RHR) Residual Heat X K1.09 3.6 3

Removal

006 (SF2; SF3 ECCS) Emergency X X K6.05 3.0 4

Core Coolina EA4.09 4.1 5

007 (SF5 PRTS) Pressurizer X K4.01 2.6 6

Relief/Quench Tank

008 (SF8 CCW) Component Cooling X A2.04 3.3 7

Water

010 (SF3 PZR PCS) Pressurizer X X K5.01 3.5 8

Pressure Control A3.02 3.6 9

012 (SF7 RPS) Reactor Protection X K4.05 2.7 10

013 (SF2 ESFAS) Engineered X K1.02 3.2 11

Safetv Features Actuation

022 ISF5 CCS\ Containment Coolina X A1.03 3.2 12

025 /SF5 ICE) Ice Condenser

026 (SF5 CSS) Containment Spray X X G2.1.27 3.9 13

K2.01 3.4 14

039 (SF4S MSS) Main and Reheat X X G2.2.2 4.6 15

Steam K5.01 2.9 16

059 (SF4S MFW) Main Feedwater X A1.03 2.7 17

061 (SF4S AFW) X A2.06 2.7 18

Auxiliarv/Emeraencv Feedwater

062 (SF6 ED AC) AC Electrical X A1.03 2.5 19

Distribution

063 (SF6 ED DC) DC Electrical X K2.01 2.9 20

Distribution

064 (SF6 EDG) Emergency Diesel X X G2.2.22 4.0 21

Generator K3.02 4.2 22

073 (SF7 PRM) Process Radiation X K4.01 4.0 23

Monitoring

076 (SF4S SW) Service Water X A2.02 2.7 24

078 (SF8 IAS) Instrument Air X X K3.01 3.1 25

A4.01 3.1 26

103 (SF5 CNT) Containment X X K1.01 3.6 27

A3.01 3.9 28

053 (SF1; SF4P ICS*) Integrated Control

- - . - tals: 3 2 3 3 2 2 3 3 2 2 3 Grouo Point Total: 28

ES-401 5 Form ES-401-2

ES-401 PWR Examination Outline Form ES-401-2

Plant Systems-Tier 2/Group 2 {RO)

Svstem # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* KIA Topic{s) IR #

001 (SF1 CRDS) Control Rod Drive

002 (SF2; SF4P RCS) Reactor Coolant

011 (SF2 PZR LCS) Pressurizer X K2.01 3.1 29 Level Control

014 (SF1 RPI) Rod Position X A4.02 3.4 30 Indication

015 (SF? NI) Nuclear X A3.05 2.6 31 Instrumentation

016 (SF? NNI) Nonnuclear Instrumentation

017 (SF? ITM) In-Core Temperature Monitor

027 (SF5 CIRS) Containment Iodine Removal

028 (SF5 HRPS) Hydrogen Recombiner and Puroe Control

029 {SF8 CPS} Containment Purge X A2.01 2.9 32

033 (SF8 SFPCS) Spent Fuel Pool X K4.01 2.9 33 Cooling

034 (SF8 FHS) Fuel-Handling X A1.02 2.9 34 Equipment

035 (SF 4P SG) Steam Generator

041 (SF4S SDS) Steam X K6.03 2.7 35 Dump/Turbine Bvpass Control

045 (SF 4S MTG) Main Turbine Generator

055 (SF4S CARS) Condenser Air X K3.01 2.5 36 Removal

056 (SF4S CDS) Condensate X G2.1.30 4.4 37

068 (SF9 LRS) Liquid Radwaste X K1.02 2.5 38

071 (SF9 WGS) Waste Gas Disposal

072 (SF? ARM) Area Radiation Monitorina

075 {SF8 CW) Circulating Water

079 {SF8 SAS**) Station Air

086 Fire Protection

050 (SF 9 CRV*) Control Room Ventilation

KIA Category Point Totals: 1 1 1 1 0 1 1 1 1 1 1 Group Point Total: 10

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

Facility: Salem ILOT 16-01 Date of Exam: January 2018

Category K/A# Topic RO SRO-only

IR # IR #

2.1.18 Ability to make accurate, clear, and concise logs, records, status 3.6 39 boards, and reoorts

2.1.28 Knowledge of the purpose and function of major system 4.1 40 comoonents and controls

2.1.43 Ability to use procedures to determine the effects on reactivity of 4.1 41 1 . Conduct of plant changes, such as reactor coolant system temperature, Operations secondary plant, fuel depletion, etc.

2.1.

2.1.

2.1.

Subtotal 3

2.2.21 Knowledge of pre- and post-maintenance operability 2.9 42 reauirements

2.2.38 Knowledae of conditions and limitations in the facilitv license 3.6 43 2. Equipment 2.2. Control

2.2.

2.2.

Subtotal 2

2.3.7 Ability to comply with radiation work permit requirements during 3.6 44 normal or abnormal conditions

2.3.13 Knowledge of radiological safety procedures pertaining to 3.4 45 licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation I areas, aligning

3. Radiation filters, etc

Control 2.3.14 Knowledge of radiation or contamination hazards that may arise 3.4 46 durina normal, I abnormal, or emeraencv conditions or activities

2.3.

2.3.

2.3.

Subtotal 3

2.4.23 Knowledge of the bases for prioritizing emergency procedure 3.6 47 imolementation during emergency operations

2.4.47 Ability to diagnose and recognize trends in an accurate and 4.2 48 timely manner utilizing the appropriate control room reference

4. material Emergency

2.4. Procedures/ Plan 2.4.

2.4.

2.4.

Subtotal 2

Tier 3 Point Total 10 7

ES-401 Record of Rejected K/ As Form ES-401-4

Tier/ Randomly Reason for Rejection Group Selected KIA

R056 029 Anticipated Kl A: Ability to operate and monitor the following as they apply to a Transient Without ATWS: Operating switch for charging pump recirculation valve

1/ 1 Scram (ATWS)-EAl.07 CV139 & CV140 are the Charging Pump recirculation valves. EOP-

FRSM-1 (RESPONSE TO NUCLEAR POWER GENERATION) makes no reference to these valves. Consequently, to write a question about operating the switches for these two valves in accordance with EOP-FRSM-1 is not possible.

Replaced with 029 Anticipated Transient Without Scram (ATWS)-EAl.01

Ability to operate and monitor the following as they apply to a ATWS: Charging pumps

R057 EPE: 038 Steam Kl A: Knowledge of the reasons for the following responses as the apply to Generator Tube the SGTR:RCS loop isolation values

1/1 Rupture (SGTR)-EK3.07 Salem does not have the capability to isolate any RCS loops.

Consequently, the SGTR EOPs do not address this Kl A. Need to resample.

Replaced with 038 Steam Generator Tube Rupture (SGTR)-EK3.09

Knowledge of the reasons for the following responses as the apply to the SGTR: Criteria for securing/throttling ECCS

R058 APE: 054 Loss of Kl A: Ability to control radiation releases Main Feedwater

1/1 (MFW)-G2.3.11 S2.0P-AB.CN-0001 (MAIN FEEDWATER/CONDENSATE SYSTEM ABNORMALITY) does not have any guidance on controlling / monitoring radiation releases.

Replaced with 054 Loss of Main Feedwater (MFW)-G2.4.49

Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

R067 APE: 024 Emergency Kl A: Knowledge of procedures relating to a security event (non-safeguards

1/2 Boration-G2.4.28 information).

There is no direct tie between Emergency Boration procedural guidance and security procedures.

Replaced with APE: 024 Emergency Boration -G2.4.34

Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects

RO 13 026 Containment Kl A: Ability to comply with radiation work permit requirements during Spray System (CSS)- normal or abnormal conditions

2/1 G2.3.7 Complying with RWPs is a generic task that is not unique to Containment Spray. Any attempt to write a question relating RWP compliance and Containment Spray would be a "tack-on"

Replaced with 026 Containment Spray System (CSS)-G2.1.27

Knowledge of system purpose and/or function.

ES-401 Record of Rejected K/ As Form ES-401-4

R021 064 Emergency Kl A: Knowledge of emergency communications systems and techniques

2/1 Diesel Generators (ED/G)-02.4.43 Emergency communication techniques are generic tasks that are not

unique to EDGs. Any attempt to write a question relating Emergency communication techniques and EDGs would be a "tack-on"

Replaced with 064 Emergency Diesel Generators (ED/G)-G2.2.22

Knowledge oflimiting conditions for operations and safety limits

R030 014 Rod Position Kl A: Ability to manually operate and/or monitor in the control room:

2/2 Indication System Primary coil voltage measurement (RPIS)-A4.03

Salem does not have the ability to operate and/or monitor RPIS primary coil voltage in the MCR

Replaced with 014 Rod Position Indication System (RPIS)-A4.02

Ability to manually operate and/or monitor in the control room: Control rod mode-select switch

R037 056 Condensate Kl A: Ability to track Technical Specification limiting conditions for

2/2 System-02.2.23 operations

The Condensate System does not have any safety function and there are no LCOs concerning the Condensate System.

Replaced with 056 Condensate System -G2.1.30

Ability to locate and operate components, including local controls

SR02 EPE: 011 Large Kl A: Knowledge of the refueling process

1/1 Break LOCA-02.1.14

There is no direct tie between the LOCA Series EOPs and refueling. Any attempt to write a refueling question during a LOCA would be a "tack-on"

Replaced with 011 Large Break LOCA-G2.1.2

Knowledge of operator responsibilities during all modes of plant operation

SR03 APE: 054 Loss of Kl A: Knowledge of the process for managing maintenance activities Main Feedwater during power operations, such as risk assessments, work prioritization, and

1/1 (MFW)-02.2.17 coordination with the transmission system operator

There is no direct tie between the S2.0P-AB.CN-0001 (MAIN FEEDWATER/CONDENSATE SYSTEM ABNORMALITY) and maintenance activities. Any attempt to write a maintenance / work control question during a Loss of Main Feedwater would be a "tack-on".

Replaced with 054 Loss of Main Feedwater (MFW)-G2.1.6

Ability to manage the control room crew during plant transients

ES-401

SR04

1/1

SR06

1/1

SROlO

1/2

SRO 12

2/1

Record of Rejected K/ As Form ES-401-4

APE: 058 Loss of DC Kl A: Ability to determine and interpret the following as they apply to the Power-AA2.0l Loss of DC Power: That a loss of de power has occurred; verification that

substitute power sources have come on line

W/El 1 Loss of Emergency Coolant Recirculation-G2.3.l l

W/El6 High Containment Radiation-G2. l .14

007 Pressurizer Relief Tank/Quench Tank System (PRTS)­A2.07

Salem does not an AB or EOP that deals with malfunctions within the DC power system. All malfunctions within the DC power system are addressed using Alarm Response Procedures. Additionally, the majority of ARPs are too similar to write a discerning question. "Check if a DC breaker is tripped and refer to Protective Circuit Breaker / Breaker reset and reclosure policy. Initiate a notification to correct the cause of the alarm".

Need to resample including new procedure.

Replaced with 015/017 Reactor Coolant Pump (RCP) Malfunctions­AA2.10

Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): When to secure RCPs on loss of cooling or seal injection

Kl A: Ability to control radiation releases

There is no direct tie between the EOP-LOCA-5 (LOSS OF EMERGENCY RECIRCULATION) and controlling radiation releases. In fact, EOP-LOCA-5 is trying to prevent core damage and minimizing radiation being released to the public.

Replaced with W/Ell Loss of Emergency Coolant Recirculation­G2.1.9

Ability to direct personnel activities inside the control room

Kl A: Knowledge of criteria or conditions that require plant-wide announcements, such I as pump starts, reactor trips, mode changes, etc

There is no direct tie between the FRCE-3 (RESPONSE TO HIGH CONTAINMENT RADIATION) and making plant-wide announcements. Any attempt to write a plant-wide announcement requirement question during implementation ofFRCE-3 would be a "tack-on". Need to resample.

Replaced with W/E16 High Containment Radiation-G2.4.1

Knowledge ofEOP entry conditions and immediate action steps

KIA: Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Recirculating quench tank.

Unable t write a discerning SRO question dealing with the PRT

Need to resample including new system.

Replaced with 005 (SF4P RHR) Residual Heat Removal- AA2.02

Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Leakage ofreactor coolant from RHR into closed cooling water system or into reactor building atmosphere

ES-401 Record of Rejected K/As Form ES-401-4

SRO 13 022 Containment K/ A: Ability to use procedures related to shift staffing, such as minimum

2/1 Cooling-G2.1.5 crew complement, overtime limitations, etc.

There is no direct tie between Containment Cooling System and the ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations. Any attempt to write a staffing question as related to containment cooling would be a "tack-on".

Additionally, different aspects of the CFCUs (which provide containment cooling) have been tested on R026, R027, R039, R040, R051 and R054. Testing another different aspect of the CFCUs at the SRO level seems to be an oversampling of containment cooling.

Need to resample including new system.

Replaced with 025 Containment Spray-G2.2.40

Ability to apply Technical Specifications for a system

SRO 14 073 Process Radiation KIA: Ability to (a) predict the impacts of the following malfunctions or

2/1 Monitoring (PRM) operations on the PRM system; and (b) based on those predictions, use System-A2.03 procedures to correct, control, or mitigate the consequences of those

malfunctions or operations: Calibration drift.

Unable to write a discerning question that complies with SRO-Only guidance of NUREG-1021. Operations Procedures concerning Radiation monitors address detector failure but really don't address calibration drift. Need to resample.

073 Process Radiation Monitoring (PRM) System-A2.02

Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure

SRO 17 068 Liquid Radwaste K/ A: Knowledge of operator response to loss of all annunciators.

2/2 System (LRS)-

A loss of all annunciators is a generic malfunction that is not specifically G2.4.32 related to the Liquid Radwaste System. Any attempt to write a loss of all annunciators question related to the Liquid Radwaste System would be a "tack-on".

Additionally, SRO Q8 and SRO Q24 are questions that have already been sampled on this exam that also relate to radiation releases. Writing a third SRO question concerning radiation released is an oversampling.

Need to resample including new system.

Replaced with 033 Spent Fuel Pool Cooling-G.2.1.20

Ability to interpret and execute procedure steps

ES-401 Record of Rejected K/ As Form ES-401-4

SR023 G2.3.5 K/ A: Ability to use radiation monitoring systems, such as fixed radiation

3 monitors and alarms, portable survey instruments, personnel monitoring equipment, etc ..

Throughout the entire RO and SRO exam, 7 questions have tested some type of fixed radiation monitor (RO 52, RO 75, RO 7, RO 23, SRO 8, SRO 24, and SRO 14). This is an over sampling resulting in not being able to develop an additional SRO question without overlap.

Need to resample to generic Emergency Procedures I Plan (2.4)

Replaced with G2.4.21

Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.