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Final Written Examination and Operating Test Outlines (Folder 3)
	ML21088A064
Person / Time
Site:	Beaver Valley
Issue date:	02/18/2021
From:	David Silk; Operations Branch I
To:	; FirstEnergy Nuclear Operating Co
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ML19353E013	List: ML20016A415; ML21074A331; ML21074A332; ML21088A058; ML21088A059; ML21088A060; ML21088A062; ML21088A063; ML21088A064; ML21088A067; ML21088A069; ML21088A070;
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ES-301 Administrative Topics Outline Form ES-301-1 Facility: Beaver Valley Unit 1 Date of Examination: 3/1 thru 3/12, 2021 _

Examination Level RO SRO Operating Test Number: BV1LOT21 NRC __

Administrative Topic Type Describe activity to be performed Code*

(See Note)

Conduct of Operations D, R 2.1.43 (4.1)

(RO A 1.1)

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

JPM 1AD-016 Plot and Evaluate 1/M Data (RO)

Conduct of Operations M, R 2.1.23 (4.3)

Ability to perform specific system and integrated plant procedures (RO A 1.2) during all modes of plant operation.

JPM 1AD-007 Perform An Estimated Critical Position Calculation (RO) 2.2.37 (3.6)

Equipment Control N, R Ability to determine operability and/or availability of safety related (RO A 2) equipment.

JPM 1AD-053 Perform Accident Monitoring Instrument Channel Checks. (1OST-6.7) (RO)

Radiation Control D, R 2.3.11 (3.8)

(RO A 3) Ability to control radiation releases.

JPM 1AD-004 Perform Decay Tank Discharge Pre-Release Verification (Gas Discharge) (RO)

Emergency Plan NOT EVALUATED (RO A 4)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics (which would require all 5 items).

Type Codes & Criteria (C)ontrol Room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; < 4 for SROs & RO retakes)

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

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

NUREG-1021, Revision 11 BVPS Rev AG

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Beaver Valley Unit 1 Date of Examination: 3/1 thru 3/12, 2021 _

Examination Level RO SRO Operating Test Number: BV1LOT21 NRC __

Administrative Topic Type Describe activity to be performed Code*

(See Note)

Conduct of Operations N, R 2.1.34 (3.5)

Knowledge of primary and secondary plant chemistry limits.

(SRO A 1.1)

JPM 1AD-054 Determine Actions for Plant Chemistry Out Of Tolerance.

(SRO)

Conduct of Operations M, R 2.1.23 (4.4)

Ability to perform specific system and integrated plant procedures during all (SRO A 1.2) modes of plant operation.

JPM 1AD-011 Review an Estimated Critical Position Calculation (SRO)

Equipment Control D, R 2.2.17 (3.8)

(SRO A 2) Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

JPM 1AD-041 Perform A Risk Assessment [1FW-P-3A Maintenance]

(SRO)

Radiation Control D, R 2.3.11 (4.3)

(SRO A 3) Ability to control radiation releases.

JPM 1AD-042 Determine Compensatory Actions for RM-P-1GW-108 and O2A-1GW-110-1 being OOS (SRO)

Emergency Plan D, R 2.4.40 (4.5)

Knowledge of SRO responsibilities in emergency plan implementation.

(SRO A 4)

JPM 1AD-052 Classify an Emergency Event - Site Area Emergency (SRO)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics (which would require all 5 items).

Type Codes & Criteria (C)ontrol Room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; < 4 for SROs & RO retakes)

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

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

NUREG-1021, Revision 11 BVPS Rev AG

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Beaver Valley Unit 1 Date of Examination: 3/1 thru 3/12, 2021 __

Exam Level: RO SRO(I) SRO(U) Operating Test No.: BV1LOT21 NRC _

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

System / JPM Title Type Code* Safety Function S1 - Emergency Borate the RCS (1CR-511) A, D, S 1 S2 - Place Excess Letdown in Service (1CR-056) D, S 2 S3 - Isolate SI Accumulators During A LOCA (1CR-642) A, D, L, EN, 3 S

S4 - Start a Reactor Coolant Pump (1CR-513) A, D, L, S 4P S5 - Manually Actuate CIB (1CR-078) M, EN, L, S 5 S6 - Synchronize and Load No. 1 Diesel Generator (1CR-524) A, D, S 6 S7 - Feeding Steam Generators From Condensate System per FR-H.1 N, L, S 4S (uses ATT J) (1CR-190)

S8 - Set Up Steam Dumps In The Steam Pressure Mode (PT-1MS- A, N, S 7 464A fails) (1CR-691)

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

P1 - Reset the Emergency Diesel Generator After an Overspeed trip D, E 6 (1PL-002)

P2 - Place a Reactor Protection Channel in the Tripped Condition D, E 7 (1PL-009)

P3 - Reset the Terry Turbine Trip Throttle Valve (1PL-004) D, E, R 4S

@ All RO and SRO 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 / SRO-I / SRO-U (A)lternate Path 4-6 /4-6 /2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / > 1 (control room system)

(L)ow-power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3/ 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator NUREG-1021, Revision 11 BVPS Rev AG

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Beaver Valley Unit 1 Date of Examination: 3/1 thru 3/12, 2021 __

Exam Level: RO SRO(I) SRO(U) Operating Test No.: BV1LOT21 NRC _

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

System / JPM Title Type Code* Safety Function S1 - Emergency Borate the RCS (1CR-511) A, D, S 1 S2 - Place Excess Letdown in Service (1CR-056) D, S 2 S3 - Isolate SI Accumulators During A LOCA (1CR-642) A, D, L, EN, 3 S

S4 - Start a Reactor Coolant Pump (1CR-513) A, D, L, S 4P S6 - Synchronize and Load No. 1 Diesel Generator (1CR-524) A, D, S 6 S7 - Feeding Steam Generators From Condensate System per FR-H.1 N, L, S 4S (uses ATT J) (1CR-190)

S8 - Set Up Steam Dumps In The Steam Pressure Mode (PT-1MS- A, N, S 7 464A fails) (1CR-691)

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

P1 - Reset the Emergency Diesel Generator After an Overspeed trip D, E 6 (1PL-002)

P2 - Place a Reactor Protection Channel in the Tripped Condition D, E 7 (1PL-009)

P3 - Reset the Terry Turbine Trip Throttle Valve (1PL-004) D, E, R 4S

@ All RO and SRO 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 / SRO-I / SRO-U (A)lternate Path 4-6 /4-6 /2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / > 1 (control room system)

(L)ow-power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3/ 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator NUREG-1021, Revision 11 BVPS Rev AG

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Beaver Valley Unit 1 Date of Examination: 3/1 thru 3/12, 2021 __

Exam Level: RO SRO(I) SRO(U) Operating Test No.: BV1LOT21 NRC _

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

System / JPM Title Type Code* Safety Function S1 - Emergency Borate the RCS (1CR-511) A, D, S 1 S3 - Isolate SI Accumulators During A LOCA (1CR-642) A, D, L, EN, 3 S

S8 - Set Up Steam Dumps In The Steam Pressure Mode (PT-1MS- A, N, S 7 464A fails) (1CR-691)

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

P1 - Reset the Emergency Diesel Generator After an Overspeed trip D, E 6 (1PL-002)

P3 - Reset the Terry Turbine Trip Throttle Valve (1PL-004) D, E, R 4S

@ All RO and SRO 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 / SRO-I / SRO-U (A)lternate Path 4-6 /4-6 /2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / > 1 (control room system)

(L)ow-power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3/ 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator NUREG-1021, Revision 11 BVPS Rev AG

Facility: BVPS Unit 1 Scenario No. 1 Op Test No.: 1LOT21 NRC Examiners: Candidates: SRO ATC BOP Initial IC 156 (15): 55% power, 500 MWe, MOL, Equ. XE Conditions, CB D @ 169 steps, RCS Conditions: boron - 1180 ppm.

Turnover: 1FW-P-1A was returned to service 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago, maintain 55% power for 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months confidence run of feedwater pump.

Critical Tasks: 1. CT-16 (E-1.C) Stop RCP's

2. CT-51 (FR-S.1.B) Start AFW pumps

3. CT-52 (FR-S.1.C) Initiate negative reactivity Event Malf. No. Event Type Event Description No.

(C,A) BOP, SRO 1 AUX13G (0 0) VS-F-4A spurious trip, requires BOP to start VS-F-4B.

(TS) SRO 2 XMT-CNS008A 0 (TS) SRO RWST Level transmitter, LT-1QS-100B fails low Turbine valve position limiter fails low, causes ~ 100 mw TUR15 (0 0) 27 10 (C,A) ALL 3 load reduction with malfunction of Rod speed at 5 SPM.

CRF06 5 (TS) SRO BOP to remove turbine from Limiter.

B SG Feedwater valve, FCV-1FW-488, begins oscillating, 4 FWM09B (7 0) 25 0 (C,A) BOP, SRO requiring BOP to manually control level.

5 (N) BOP, SRO Shutdown of 1FW-P-1A.

(R) ATC 6 IMF EPS18 (0 0) TRUE Loss of Main transformer due to Differential Overcurrent.

(C,A) BOP, SRO CRF12A Loss of C 4Kv bus with Failure of Automatic and manual 7 (M) ALL CRF12B Reactor trip from the control room requires entry into FR-S.1.

INH20 INH21 All Aux Feedwater pumps fail to automatically start, requires 8 (C) BOP, SRO INH35 BOP to start AFW pumps.

INH36 9 SIS08 (M) ALL 1200 gpm LOCA occurs on "A" Loop.

10 INH40 (C) ATC, SRO CH-P-1B fails to auto start on SI signal.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E-0 FR-S.1 E-0 E-1 ES-1.2

1L21N1 After taking the shift at 55% power with directions to maintain power, MOL, Leak Collection Exhaust fan, 1VS-F-4A, will trip, the crew will respond using the ARP which will direct the BOP to manually start 1VS-F-4B. The SRO will address applicable TS and License Requirements Manual.

The B RWST level channel, LT-1QS-100B, will then fail low, the SRO will address Tech Specs for the failed RWST level channel.

A malfunction will then occur with the turbine valve position limiter causing a 100 mw load rejection, with a Rod speed malfunction that causes control rods to auto insert at 5 SPM, requiring the ATC to manually insert rods or borate the RCS.

At the same time, a malfunction will occur with the B main feed regulating valve controller, FCV-1FW-488, causing oscillations in the B SG level. IAW AOP 1.4.1, the BOP will manually stabilize and control level.

After level is stabilized, the BOP will take actions to Recover the Governor Valves from the Limiter IAW 1OM-26.4.AK.

The SRO will enter AOP 1.35.2, Load Rejection, to stabilize the plant and address DNB technical specifications.

The load rejection will cause reactor power to decrease to <49% (P-9).

A report from the field will state that there is a significant oil leak on the A Main feedwater pump requiring the BOP to shutdown the pump.

There will then be a loss of the main transformer due to Differential overcurrent causing an immediate turbine trip. The crew will enter AOP 1.26.1 and begin reducing Rx power. The rod speed remains at 5 SPM when in AUTO, requiring the ATC to manually insert rods.

When Rx power lowers to <13% or the condenser steam dumps are placed in Steam Pressure mode, the C 4Kv bus will fail, resulting in a loss of the C RCP and the B Main feedwater pump, the SRO will direct the ATC to manually trip the reactor.

The ATC will unsuccessfully attempt to trip the reactor from BB-B and BB-A and begin manually inserting the control rods as auto rod insertion capability failed when power reduced to < 38%.

The SRO will enter FR-S.1 with the ATC and BOP performing the IOAs.

When Rx power is <3% and after Emergency boration flow is established in FR-S.1, if the crew previously dispatched an operator to locally trip the reactor, the reactor will be locally tripped. The ATC will verify reactor power is <5% after which the SRO will return to E-0.

3 minutes after the Rx is locally tripped, a 1200 gpm LOCA will occur on the A loop resulting in an automatic SI, the B charging/HHSI pump will fail to automatically start on the SI signal.

Additionally, during the ATWS condition, auto start is inhibited for Auxiliary feed water pumps.

After returning to E-0, the SRO will determine that the RCS is not intact and transition to E-1.

The crew will shutdown the A and B RCPs due to meeting RCP trip criteria.

The scenario will be terminated at the lead evaluators discretion after the crew transitions to ES-1.2.

Expected procedure flow path is E-0 FR-S.1 E-0 E-1 ES-1.2.

Facility: BVPS Unit 1 Scenario No. 2 Op Test No.: 1LOT21 NRC Examiners: Candidates: SRO ATC BOP Initial IC-161 (10): 100% power, BOL, Equ. XE Conditions, CB D @ 225 steps, Conditions: RCS boron - 1280 ppm.

Turnover: Maintain 100% power.

Critical Tasks: 1. CT-2 (E-0.D) Manually actuate 1 train of Safety Injection

2. CT-43 (FR-H.1.A) Establish feed flow to SG before Feed and Bleed is required

3. CT-17 (E-2.A) Isolates faulted SG.

Event Malf. No. Event Type Event Description No.

(C,A) ATC, SRO 1 CCW3A Trip of running CCR pump, with Autostart failure of standby.

(TS) SRO (I,A) ATC, SRO Master Pressure controller drifts to 85%, requires ATC to 2 PRS12 85 45 (TS) SRO manually control RCS pressure.

(I,A) ATC, SRO 3 NIS03D N44 failed high, control rods auto insert. (AOP 1.1.3)

(TS) SRO 4 (N) BOP, SRO N44 removal from service. (AOP 1.2.1C)

(R ) ATC Turbine Bearing failure - requires load reduction, 5 TUR03E 8 30 (C,A) BOP, SRO (Management directed at 2%/minute IAW AOP 1.51.1) 6 TUR03E 15 60 (C) ATC, SRO Turbine High vibrations - requires Rx trip.

Steam Header Break in turbine building occurs on Rx trip -

7 MSS02A,B,C 2.5E6 (M) ALL requires Safety Inj.

SIS10A 8 SIS10B (C) ATC, SRO Automatic SI fails to actuate - requires manual actuation.

9 VLV-MSS18 100 (C) BOP, SRO "C" SG Mainsteam line isolation valve fails to auto close.

10 EPS19 1 (C) BOP, SRO Exciter circuit breaker fails to auto open on Rx trip.

FWM11C (4 30) 11 FWM11A (0 0)

(M) ALL Loss of all Aux Feedwater flow, requires entry into FR-H.1 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E-0 FR-H.1 E-0 ES-1.1.

1L21N2 After taking the shift at 100% power, BOL. The A CCR pump will trip due to a faulty breaker with a failure of the B to auto start. The crew will enter AOP 1.15.1. The ATC will manually start the B CCR pump, the crew will dispatch an operator to place the C pump in service on the AE 4kv bus. The SRO will address Technical Specifications.

The Master Pressure controller will then fail to 85% causing the PRZR spray valves to open. IAW AOP 1.4.1, the ATC will close the spray valves to stabilize RCS pressure. The ATC will manually control PRZR pressure for the remainder of the scenario. The SRO will address Technical Specifications for DNB.

Power Range Nuclear instrument, N-44 will then fail high causing the control rods to automatically insert. The crew will perform the Immediate Operator Actions for AOP 1.1.3, Unexpected Control Rod Movement. The ATC will identify the N-44 failure and place the rods in manual. The SRO will then transition to AOP 1.2.1C, Power Range Channel Malfunction, and direct the BOP to remove the failed channel from service. The SRO will address Tech Specs for the failed instrument.

A bearing will then begin failing on the turbine causing turbine vibrations to begin increasing, IAW the alarm response procedure, the crew will address the bearing degradation and perform a management directed power reduction at 2%/minute IAW AOP 1.51.1 to remove the turbine from service. After the reactor power has been reduced ~6%, the turbine bearing will fail causing vibrations to exceed 14 mils - requiring the turbine to be manually tripped.

Upon the Rx trip, a steam header break will occur in the turbine building. An automatic steamline isolation will occur, however the "C" Mainsteam line isolation valve will fail to automatically close requiring the BOP to manually close it. The fault will also result in an SI being required, however, Safety Injection will not automatically actuate, requiring manual actuation to initiate SI flow.

Aux Feedwater malfunctions will occur such that the turbine driven pump, 1FW-P-2 trips on start, 1FW-P-3A/B will not start.

The crew will enter E-0 on the reactor trip, and then enter FR-H-1 due to no available auxiliary feed water pumps.

During the immediate operator actions of E-0, the BOP will be required to manually open the exciter circuit breaker due to an auto open failure.

After Feedwater has been established using either the dedicated Feedwater pump, 1FW-P-4 or either Main feed pump, the crew will return to E-0, determine that SI is not required and transition to ES-1.1 to terminate SI flow.

The scenario will be terminated after the crew terminates SI flow in ES-1.1.

Expected procedure flow path is E-0 FR-H.1 E-0 ES-1.1.

Facility: BVPS Unit 1 Scenario No. 3 Op Test No.: 1LOT21 NRC Examiners: Candidates: SRO ATC BOP Initial IC 167(5): ~4.5% power, BOL, Xe increasing, CB D @ 110 steps, RCS boron - 1815 Conditions: ppm.

Turnover: Continue power increase IAW reactivity plan and commence turbine roll.

After taking the watch, place 1LO-M-9B inservice IAW 1OM-26.4.D, part D Critical Tasks: 1. CT-1 (E-0.A), Manually trip the Reactor.

2. CT-3 (E-0.E), Actuate minimum CNMT cooling

3. CT-17 (E-2.A), Isolates faulted SG.

Event Malf. No. Event Type Event Description No.

1 N/A (N) BOP, SRO S/U B EHC pump, 1LO-M-9B, IAW 1OM-26.4.D, part D (R) ATC 2 N/A Normal power increase to 15% IAW 1OM-52.4.B (N) SRO 3 X06i030L (C,A) ATC, SRO Blender fails to stop after dilution 4 XMT-CFW032A (TS) SRO "A" SG Level Channel 2 fails HIGH (TS only), LT-1FW-475 C Bypass feed regulating valve fails in Auto. Requires 5 FWM08C (C,A) BOP, SRO manual control (I,A) ATC, SRO LT-1RC-459 drifts high, requires ATC to place alternate 6 PRS06A (0 0) 100 60 (TS) SRO channel in service.

7 MSS01B (20 20) 1E7 (M) ALL Large Steam Fault inside CNMT on B SG.

8 CRF12A (C) ATC, SRO Auto Reactor trip failure, requires manual RX trip.

9 INH50 (C) ATC, SRO CIB auto actuation failure 10 VLV-MSS21,22,23 (C) BOP, SRO TV-1MS-111A,B,C fail to close on MSLI signal IOR X19o005S OFF Control room emergency ventilation system fails to actuate on 11 (C) BOP, SRO IOR X19o020S OFF CIB signal, requiring manual initiation.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E-0 E-2 ES-1.1

1L21N3 The crew will assume the shift at ~4.5% power with instructions to raise power in accordance with the reactivity plan and 1OM-52.4.B., Additionally, the turnover includes direction to place the "B" EHC pump, 1LO-M-9B, in service to support turbine roll upon taking the watch.

When the ATC performs a dilution IAW the reactivity plan, the dilution will not automatically shutoff requiring the ATC to manually stop the dilution.

After the power has raised to >5.5%, an A SG level transmitter, 1FW-LT-475, will fail high. The BOP will verify the SG level remains stable, the SRO will enter 1OM-24.4.IF attachment 1 and address TS for the failed channel. There are no control actions associated with this failure.

Additionally, after power has raised to >5.5%, C SG Bypass Feed regulating valve will fail while in Auto, failure will become evident as Rx power continues to be raised, the BOP will be required to identify the malfunction and take action to control FCV-1FW-499 in manual IAW AOP 1.4.1 Process Control Failure.

LT-1RC-459 will then drift high. The ATC will recognize the failure and respond IAW AOP 1.4.1 IOA's to remove the failed channel from service. The SRO will transition to the Instrument Failure procedure, 1OM-6.4.IF, attachment 1 and direct the ATC complete the removal of the 459 channel from service and then review applicable Technical Specifications in effect for the failed level transmitter.

A large steam fault will then occur on the B SG inside of CNMT, however the Rx will fail to automatically trip.

The ATC will trip the Rx and the crew will enter E-0 and transition to E-2 after diagnosing a faulted SG. The steam break is of a magnitude that will raise CNMT pressure above CIB actuation setpoint, however automatic CIB will not occur.

Additionally, the main steamline drain valves, TV-1MS-111A,B,C, fail to automatically close on the MSLI signal and the control room emergency ventilation system, CREVs, fails to automatically actuate on the CIB signal.

After the crew completes SG isolation IAW E-2, they will transition to ES-1.1.

The scenario will be terminated at the Lead Evaluators discretion after the crew has shutdown the Low Head SI pumps.

Expected procedure flow path is E-0 E-2 ES-1.1.

Facility: BVPS Unit 1 Scenario No. 4 Op Test No.: 1LOT21 NRC Examiners: Candidates: SRO ATC BOP Initial IC-170 (17): 75% power, MOL, stable Xe, CB D @ 189 steps, RCS boron - 1080 ppm.

Conditions:

Turnover: Maintain stable plant conditions.

Startup standby Turbine Plant River Water pump.

Critical Tasks: 1. CT-18 (E-3.A) Crew isolates ruptured SG

2. CT-19 (E-3.B) Crew establishes/maintains temperature

3. CT-34 (ECA-3.1.B) Crew initiates C/D to cold shutdown Event Malf. No. Event Type Event Description No.

1 N/A (N) BOP, SRO Startup standby TPRW pump (C,A) BOP, SRO 2 XMT-MSS021A 0 PT-1MS-446 fails LOW (TS) SRO (C,A) ATC, SRO 3 RCS02A (0 0) 45 35 gpm RCS leak, "A" Loop, unisolable (TS) SRO (R) ATC Management ordered plant S/D @ 5%/min due to RCS 4 N/A (C,A) BOP, SRO leakage.

Pressurizer pressure transmitter PT-1RC-444 fails high in PRS08D (C,A) ATC, SRO automatic, PORV, PCV-1RC-455C opens with reseat failure.

5 PRS03A (TS) SRO PORV block valve, MOV-1RC-535, fails to close. Requires manual Reactor trip.

6 X07i038L OFF (C) ATC, SRO BB-B Reactor trip switch failure 7 RCS03A (M) ALL Reactor trip causes a 650 gpm 1A SG tube rupture.

Reheat steam fails to auto isolate on trip - requires closing 8 MSS03, MSS04 (C) BOP, SRO MOV-1MS-100A and B.

SG BD isolation failure, requires manual alternate valve 9 VLV-SGB01,02,03 (C) BOP, SRO closure.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E-0 E-1 E-3 ECA-3.1

1L21N4 The crew will assume the shift at approximately 75% power with instructions to startup the standby Turbine Plant River Water pump, 1WR-P-6B and place the running pump, 1WR-P-6A in standby, IAW 1OM-30.4.N, Standby Turbine Plant River Water Pump Startup.

The non-selected Turbine First Stage pressure transmitter, PT-1MS-446 will fail low. IAW the instrument failure procedure, the crew will take action to transfer the condenser steam dump control to Steam Pressure mode. The SRO will address Tech Specs for the failed channel.

A 35 gpm RCS leak will develop on the A loop, IAW AOP 1.6.7, the crew will raise charging and isolate letdown to evaluate and quantify the leakage. The SRO will evaluate and enter Technical Specifications.

After the crew has determined that the RCS leak is not isolable and restored letdown, Operations Management will direct the crew to perform a plant shutdown at 5% / minute IAW AOP 1.51.1, Unplanned Power Reduction.

After Rx power has reduced ~10%, PT-1RC-444 will fail high causing the spray valves and PORV 455C to open, The ATC will be required to manually close spray valves and PORV, PCV-1RC-455C, IAW AOP 1.4.1 immediate operator actions. The US will enter AOP 1.4.1 and then transition to 1OM-6.4.IF, Attachment 2 and determine applicable Tech Spec actions. PORV 455C will fail to completely reseat, the ATC will then attempt to close the MOV block valve, MOV-1RC-535, which will fail to close.

Due to the leaking PORV, the SRO will direct the ATC to manually trip the Rx. When the ATC attempts to trip the Rx from BB-B, the Trip switch will fail, the ATC will then trip the Rx using BB-A trip pushbutton.

The Rx trip will cause a 650 gpm SGTR to occur on the A SG.

The crew will progress thru E-0, the BOP will identify that Reheat steam and SG blowdown flowpaths both failed to automatically isolate on the trip and take appropriate manual actions.

The crew will transition to E-1 at the check the PORVs step of E-0, then transition to E-3 due to the indications of a SGTR.

The crew will progress thru E-3, after the crew completes the RCS cooldown, they will transition to ECA-3.1 due to the open PORV.

The scenario will be terminated at the Lead Evaluators discretion when the crew performs an RCS cooldown IAW ECA-3.1, recognizes 100F/hr cooldown limit and demonstrates actions to control the cooldown to within limits.

Expected procedure flow path is E-0 E-1 E-3 ECA-3.1.

aFacility: BVPS Unit 1 Scenario No. 5 Op Test No.: 1LOT21 NRC Examiners: Candidates: SRO ATC BOP Initial IC-171 (17): 64% power,613MWe, MOL, Equ. XE Conditions, CB D @ 185 steps, Conditions: RCS boron - 1100 ppm.

Turnover: Maintain current power level.

Startup 1SD-P-1B and Shutdown 1SD-P-1A IAW 1OM-23B.4.G.

Critical Tasks: 1. CT-6 (E-0.I) - Establish 1 train of HHSI

2. CT-10 (E-0.M) - Close PORV Block valve

3. CT-13 (E-0.Q) - Manually trip the turbine Event Malf. No. Event Type Event Description No.

1 N/A (N) BOP, SRO Startup 1SD-P-1B and S/D 1SD-P-1A 2 FWM11C (TS) SRO 1FW-P-2 trip Throttle valve failure.

IMF MSS16B (0 0) 0 (I,A) BOP, SRO A SG Ch 3 steam press transmitter, PT-1MS-475, fails low 3 30 (TS) SRO FWM01A (R) ATC Main feedwater pump trip, requires turbine load reduction and 4

(C,A) BOP, SRO manual rod insertion due to auto rod failure.

5 IMF RCS06A (10 0)

(C,A) ATC, SRO A RCP #2 seal failure, requires Rx/RCP trip.

100 300 IMF RCS02A (1 0) 300 gpm SBLOCA occurs on Rx trip, A 4kv bus trips on 6 300 (M) ALL OC, EDG-1 starts then trips on overspd - LOSS OF 1AE bus.

IMF TUR04B (5 0) 3 Incomplete turbine trip, GV-2 and TV-2 fail to close, requires 7 IMF TUR06B (0 0) (C) BOP, SRO BOP to manually trip turbine.

100 0 100 IMF PRS03B (2 0) PORV, PCV-1RC-455D fails open during Rx trip requires 8 90 10 0 (C) ATC, SRO ATC to close block valve.

IMF INH25 MOV-1SI-867D fails to auto open on SI signal, requires ATC (C) ATC, SRO 9 to manually open to align HHSI flowpath (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E-0 E-1 ES-1.1

1L21N5 After taking the shift, IAW 1OM-23B.4.G, the crew will startup the standby Heater Drain pump, 1SD-P-1B and shutdown 1SD-P-1A.

The turbine driven AFW pump Trip throttle valve will then fail, the SRO will address TS implications of AFW pump being OOS.

The A SG Channel 3 steam pressure transmitter, PT-1MS-475, will fail low, IAW AOP 1.4.1, the BOP will place the A MFRV in manual to control level, and the SRO will direct placing the alternate channel in service using 1OM-24.4.IF, Attachment 4. After FCV-1FW-478 is returned to AUTO, the SRO will address Technical Specifications.

The "A" Main feedwater pump will spuriously trip, the crew will enter AOP 1.24.1 and reduce power to less than 52%. The control rods will fail to auto insert requiring the ATC to manually insert control rods and borate for the power reduction.

During the power reduction, the A RCP seal leakoff flow will begin to decrease, the crew will respond to the Annunciator, A3-101, for Hi Seal Vent Pot Level via the ARP and then via AOP 1.6.8, Abnormal RCP Operation, the crew will continue to monitor seal leakoff flow as it decreases. Seal leakoff flow will decrease to < 0.8 gpm with an increase in radial bearing temperature which meets the immediate trip criteria in AOP 1.6.8. The ATC will manually trip the reactor, trip the A RCP and then close the spray valve.

As a result of the Rx trip, a 300 gpm SBLOCA will occur on the A loop.

The turbine will fail to completely trip on the reactor trip, GV-2 and TV-2 will fail to close requiring a manual turbine trip.

Additionally; PORV, PCV-1RC-455D will fail open causing a Safety Injection due to low PRZR pressure, requiring closure of the block valve. During the transfer to offsite power, the A 4kv bus will trip due to overcurrent; the #1 EDG will start then trip on overspeed.

With the AE 4kv emergency bus de-energized, A "B" train SIS Injection valve, MOV-1SI-867D, will fail to auto open; therefore, with no Safety Injection flow, the crew will be required to manually open MOV-1SI-867D to restore HHSI flow.

The crew will enter E-0, transition to E-1 due to containment parameters and then to ES-1.1 to terminate Safety Injection.

The scenario will be terminated when normal charging flowpath is established in ES-1.1.

Expected procedure flow path is E-0 E-1 ES-1.1.

ES-401 PWR Examination Outline Form ES-401-2 Facility: Beaver Valley Unit 1 1LOT21 Date of Exam: 3/1 through 3/12/2021 RO K/A Category Points SRO ONLY Points Tier Group K K 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 3 3 3 18 Emergency 2

& 1 1 2 2 2 1 9 Abnormal Plant Tier Evolutions Totals 4 4 5 5 5 4 27 1

2.

3 2 2 2 3 3 3 3 1 3 3 28 Plant Systems 2 1 0 1 1 1 1 1 1 1 1 1 10 Tier Totals 4 2 3 3 4 4 4 4 2 4 4 38 1 2 3 4 10

3. Generic Knowledge and Abilities Category 3 3 2 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 IRs 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, IRs, 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.

NUREG-1021, Revision 11 RO Page 1 of 13 BVPS Rev. AG

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

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000007 (EPE 7; BW E02&E10; CE E02)

Reactor Trip, Stabilization, Recovery / 1 X EK1 Knowledge of the operational implications of 3.4 the following concepts as they apply to the reactor trip:

[Question 1]

EK1.02 Shutdown margin (CFR 41.8 / 41.10 / 45.3) 000008 (APE 8) Pressurizer Vapor Space X AK1. Knowledge of the operational implications of 3.2 Accident / 3 the following concepts as they apply to a Pressurizer Vapor Space Accident:

[Question 2] AK1.01 Thermodynamics and flow characteristics of open or leaking valve (CFR 41.8 / 41.10 / 45.3) 000009 (EPE 9) Small Break LOCA / 3 X 2.4.45 Ability to prioritize and interpret the 4.1 significance of each annunciator or alarm.

[Question 3] (CFR: 41.10 / 43.5 / 45.3 / 45.12) 000015 (APE 15) Reactor Coolant Pump X AA1. Ability to operate and / or monitor the 3.0 Malfunctions / 4 following as they apply to the Reactor Coolant

[Question 4] Pump Malfunctions (Loss of RC Flow):

AA1.08 S/G LCS (CFR 41.7 / 45.5 / 45.6) 000022 (APE 22) Loss of Reactor Coolant X AA2. Ability to determine and interpret the following 3.2 Makeup / 2 as they apply to the Loss of Reactor Coolant

[Question 5] Makeup:

AA2.02 Charging pump problems (CFR 43.5/ 45.13 000025 (APE 25) Loss of Residual Heat X AK1. Knowledge of the operational implications of 3.9 Removal System / 4 the following concepts as they apply to Loss of

[Question 6] Residual Heat Removal System:

AK1.01 Loss of RHRS during all modes of operation (CFR 41.8 / 41.10 / 45.3) 000026 (APE 26) Loss of Component X AK3. Knowledge of the reasons for the following 3.6 Cooling Water / 8 responses as they apply to the Loss of Component

[Question 7] Cooling Water:

AK3.02 The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFAS (CFR 41.5,41.10 / 45.6 / 45.13) 000027 (APE 27) Pressurizer Pressure X AK2. Knowledge of the interrelations between the 2.6 Control System Malfunction / 3 Pressurizer Pressure Control Malfunctions and the

[Question 8] following:

AK2.03 Controllers and positioners (CFR 41.7 / 45.7)

NUREG-1021, Revision 11 RO Page 2 of 13 BVPS Rev. AG

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1(RO) Continued E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000029 (EPE 29) Anticipated Transient X EK3 Knowledge of the reasons for the following 4.2 Without Scram / 1 responses as the apply to the ATWS:

[Question 9] EK3.11 Initiating emergency boration (CFR 41.5 / 41.10 / 45.6 / 45.13) 000038 (EPE 38) Steam Generator Tube X EA1 Ability to operate and monitor the following as 4.0 Rupture / 3 they apply to a SGTR:

[Question 10] EA1.30 Safety injection and containment isolation systems (CFR 41.7 / 45.5 / 45.6) 000040 (APE 40; BW E05; CE E05; W E12) X AK2. Knowledge of the interrelations between the 2.6 Steam Line RuptureExcessive Heat Steam Line Rupture and the following:

Transfer / 4 AK2.01 Valves

[Question 11]

(CFR 41.7 / 45.7) 000054 (APE 54; CE E06) Loss of Main X 2.4.21 Knowledge of the parameters and logic used 4.0 Feedwater /4 to assess the status of safety functions, such as

[Question 12] reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5 / 45.12) 000055 (EPE 55) Station Blackout / 6 X EA1 Ability to operate and monitor the following as 3.5

[Question 13] they apply to a Station Blackout:

EA1.04 Reduction of loads on the battery (CFR 41.7 / 45.5 / 45.6) 000062 (APE 62) Loss of Nuclear Service X 2.1.27 Knowledge of system purpose and/or 3.9 Water / 4 function.

[Question 14] (CFR: 41.7) 000065 (APE 65) Loss of Instrument Air / 8 X AA2. Ability to determine and interpret the following 2.9

[Question 15] as they apply to the Loss of Instrument Air:

AA2.08 Failure modes of air-operated equipment (CFR: 43.5 / 45.13) 000077 (APE 77) Generator Voltage and X AK3. Knowledge of the reasons for the following 3.6 Electric Grid Disturbances / 6 responses as they apply to Generator Voltage and

[Question 16] Electric Grid Disturbances:

AK3.02 Actions contained in abnormal operating procedure for voltage and grid disturbances (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)

NUREG-1021, Revision 11 RO Page 3 of 13 BVPS Rev. AG

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1(RO) Continued E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 (BW E04; W E05) Inadequate Heat X EK2. Knowledge of the interrelations between the 3.9 TransferLoss of Secondary Heat Sink / 4 (Loss of Secondary Heat Sink) and the following:

[Question 17] EK2.2 Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

(CFR: 41.7 / 45.7)

(W E11) Loss of Emergency Coolant X EA2. Ability to determine and interpret the following 3.4 Recirculation / 4 as they apply to the (Loss of Emergency Coolant

[Question 18] Recirculation)

EA2.2 Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

(CFR: 43.5 / 45.13) 3 3 3 3 3 3 K/A Category Point Totals: Group Point Total: 18 NUREG-1021, Revision 11 RO Page 4 of 13 BVPS Rev. AG

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

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000036 (APE 36; BW/A08) Fuel-Handling X AA1. Ability to operate and / or monitor the following 3.5 Incidents / 8 as they apply to the Fuel Handling Incidents:

[Question 19] AA1.03 Reactor building containment evacuation alarm enable switch (CFR 41.7 / 45.5 / 45.6) 000037 (APE 37) Steam Generator Tube X AA2. Ability to determine and interpret the following 3.3 Leak / 3 as they apply to the Steam Generator Tube Leak:

[Question 20] AA2.12 Flow rate of leak (CFR: 43.5 / 45.13) 000059 (APE 59) Accidental Liquid X AK1. Knowledge of the operational implications of 2.7 Radwaste Release / 9 the following concepts as they apply to Accidental Liquid Radwaste Release:

[Question 21]

AK1.01 Types of radiation, their units of intensity and the location of the sources of radiation in a nuclear power plant (CFR 41.8 / 41.10 / 45.3) 000069 (APE 69; W E14) Loss of X AK2. Knowledge of the interrelations between the 2.8 Containment Integrity / 5 Loss of Containment Integrity and the following:

[Question 22] AK2.03 Personnel access hatch and emergency access hatch (CFR 41.7 / 45.7) 000074 (EPE 74; W E06 & E07) X 2.2.44 Ability to interpret control room indications to 4.2 Inadequate Core Cooling / 4 verify the status and operation of a system, and understand how operator actions and directives

[Question 23] affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45.12)

(W E02) SI Termination / 3 X EA2. Ability to determine and interpret the following 3.3 as they apply to the (SI Termination)

[Question 24]

EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(CFR: 43.5 / 45.13)

EK3. Knowledge of the reasons for the following (W E03) LOCA Cooldown X responses as they apply to the (LOCA Cooldown 3.5 Depressurization / 4 and Depressurization)

[Question 25]

EK3.4 RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

(CFR: 41.5 / 41.10, 45.6 / 45.13)

NUREG-1021, Revision 11 RO Page 5 of 13 BVPS Rev. AG

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2(RO) Continued E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 (W E13) Steam Generator Overpressure / 4 X EA1. Ability to operate and / or monitor the following 3.0 as they apply to the (Steam Generator

[Question 26] Overpressure)

EA1.2 Operating behavior characteristics of the facility.

(CFR: 41.7 / 45.5 / 45.6)

(W E16) High Containment Radiation /9 X EK3. Knowledge of the reasons for the following 2.9 responses as they apply to the (High Containment

[Question 27] Radiation)

EK3.1 Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

(CFR: 41.5 / 41.10, 45.6, 45.13) 1 1 2 2 2 1 K/A Category Point Totals: Group Point Total: 9 NUREG-1021, Revision 11 RO Page 6 of 13 BVPS Rev. AG

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

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 (SF4P RCP) Reactor X K1 Knowledge of the physical connections 2.7 Coolant Pump and/or cause-effect relationships between the RCPS and the following systems:

[Question 28]

K1.08 Containment isolation (CFR: 41.2 to 41.9 / 45.7 to 45.8) 003 (SF4P RCP) Reactor X K6 Knowledge of the effect of a loss or 2.7 Coolant Pump malfunction on the following will have on the RCPS:

[Question 29]

K6.02 RCP seals and seal water supply (CFR: 41.7 / 45/5) 004 (SF1; SF2 CVCS) Chemical X K5 Knowledge of the operational implications 3.2 and Volume Control of the following concepts as they apply to the CVCS:

[Question 30]

K5.44 Pressure response in PZR during in-and-out surge (CFR: 41.5/45.7) 005 (SF4P RHR) Residual Heat X K4 Knowledge of RHRS design feature(s) 3.1 Removal and/or interlock(s) which provide or the following:

[Question 31]

K4.08 Lineup for "piggy-back" mode with high-pressure injection (CFR: 41.7) 006 (SF2; SF3 ECCS) X A2 Ability to (a) predict the impacts of the 3.9 Emergency Core Cooling following malfunctions or operations on the ECCS; and (b) based on those predictions,

[Question 32] use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.02 Loss of flow path (CFR: 41.5 / 45.5) 006 (SF2; SF3 ECCS) X K2 Knowledge of bus power supplies to the 2.5 Emergency Core Cooling following:

[Question 33] K2.02 Valve operators for accumulators (CFR: 41.7) 007 (SF5 PRTS) Pressurizer X K4 Knowledge of PRTS design feature(s) 2.6 Relief/Quench Tank and/or interlock(s) which provide for the following:

[Question 34]

K4.01 Quench tank cooling (CFR: 41.7) 008 (SF8 CCW) Component X A3 Ability to monitor automatic operation of the 3.2 Cooling Water CCWS, including:

[Question 35] A3.01 Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS (CFR: 41.7 / 45.5) 010 (SF3 PZR PCS) Pressurizer X K5 Knowledge of the operational implications 2.6 Pressure Control of the following concepts as the apply to the PZR PCS:

[Question 36]

K5.02 Constant enthalpy expansion through a valve (CFR: 41.5 / 45.7)

NUREG-1021, Revision 11 RO Page 7 of 13 BVPS Rev. AG

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1(RO) Continued System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 010 (SF3 PZR PCS) Pressurizer X K6 Knowledge of the effect of a loss or 3.2 Pressure Control malfunction of the following will have on the PZR PCS:

[Question 37]

K6.03 PZR sprays and heaters (CFR: 41.7 / 45.7) 012 (SF7 RPS) Reactor X A4 Ability to manually operate and/or monitor 3.6 Protection in the control room:

[Question 38] A4.03 Channel blocks and bypasses (CFR: 41.7 / 45.5 to 45.8) 013 (SF2 ESFAS) Engineered X K1 Knowledge of the physical connections 3.3 Safety Features Actuation and/or cause effect relationships between the ESFAS and the following systems:

[Question 39]

K1.11 CVCS (CFR: 41.2 to 41.9 / 45.7 to 45.8) 013 (SF2 ESFAS) Engineered X K5 Knowledge of the operational implications 2.9 Safety Features Actuation of the following concepts as they apply to the ESFAS:

[Question 40]

K5.02 Safety system logic and reliability (CFR: 41.5 / 45.7) 022 (SF5 CCS) Containment X K1 Knowledge of the physical connections 2.9 Cooling and/or cause effect relationships between the CCS and the following systems:

[Question 41]

K1.04 Chilled water (CFR: 41.2 to 41.9 / 45.7 to 45.8) 026 (SF5 CSS) Containment X K3 Knowledge of the effect that a loss or 4.2 Spray malfunction of the CSS will have on the following:

[Question 42]

K3.02 Recirculation spray system (CFR: 41.7 / 45.6) 039 (SF4S MSS) Main and X A1 Ability to predict and/or monitor changes in 2.5 Reheat Steam parameters (to prevent exceeding design limits) associated with operating the MRSS

[Question 43] controls including:

A1.09 Main steam line radiation monitors (CFR: 41.5 / 45.5) 039 (SF4S MSS) Main and X A4 Ability to manually operate and/or monitor 3.8 Reheat Steam in the control room:

[Question 44] A4.04 Emergency feedwater pump turbines (CFR: 41.7 / 45.5 to 45.8)

NUREG-1021, Revision 11 RO Page 8 of 13 BVPS Rev. AG

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1(RO) Continued System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 059 (SF4S MFW) Main X A4 Ability to manually operate and monitor in 3.1 Feedwater the control room:

[Question 45] A4.11 Recovery from automatic feedwater isolation (CFR: 41.7 / 45.5 to 45.8) 061 (SF4S AFW) X K2 Knowledge of bus power supplies to the 3.7 Auxiliary/Emergency Feedwater following:

[Question 46] K2.02 AFW electric drive pumps (CFR: 41.7) 062 (SF6 ED AC) AC Electrical X A2 Ability to (a) predict the impacts of the 2.7 Distribution following malfunctions or operations on the ac distribution system; and (b) based on those

[Question 47] predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.09 Consequences of exceeding current limitations (CFR: 41.5 / 43.5 / 45.3 / 45.13) 063 (SF6 ED DC) DC Electrical X A2 Ability to (a) predict the impacts of the 2.5 Distribution following malfunctions or operations on the DC electrical systems; and (b) based on those

[Question 48] predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01 Grounds (CFR: 41.5 / 43.5 / 45.3 / 45.13) 063 (SF6 ED DC) DC Electrical X 2.2.44 Ability to interpret control room 4.2 Distribution indications to verify the status and operation of a system, and understand how operator

[Question 49] actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45.12) 064 (SF6 EDG) Emergency X K6 Knowledge of the effect of a loss or 2.7 Diesel Generator malfunction of the following will have on the ED/G system:

[Question 50]

K6.07 Air receivers (CFR: 41.7 / 45.7) 073 (SF7 PRM) Process X A1 Ability to predict and/or monitor changes in 3.2 Radiation Monitoring parameters (to prevent exceeding design limits) associated with operating the PRM

[Question 51] system controls including:

A1.01 Radiation levels (CFR: 41.5 / 45.7)

NUREG-1021, Revision 11 RO Page 9 of 13 BVPS Rev. AG

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1(RO) Continued System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 076 (SF4S SW) Service Water X 2.2.25 Knowledge of the bases in Technical 3.2 Specifications for limiting conditions for

[Question 52] operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) 078 (SF8 IAS) Instrument Air X K3 Knowledge of the effect that a loss or 3.4 malfunction of the IAS will have on the

[Question 53] following:

K3.02 Systems having pneumatic valves and controls (CFR: 41.7 / 45.6) 103 (SF5 CNT) Containment X A1 Ability to predict and/or monitor changes in 3.7 parameters (to prevent exceeding design

[Question 54] limits) associated with operating the containment system controls including:

A1.01 Containment pressure, temperature, and humidity (CFR: 41.5 / 45.5) 103 (SF5 CNT) Containment X 2.2.25 Knowledge of the bases in Technical 3.2 Specifications for limiting conditions for

[Question 55] operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) 3 2 2 2 3 3 3 3 1 3 3 K/A Category Point Totals: Group Point Total: 28 NUREG-1021, Revision 11 RO Page 10 of 13 BVPS Rev. AG

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

System # / Name K K K K K K A A A A G K/A Topic(s)

IR #

1 2 3 4 5 6 1 2 3 4 001 (SF1 CRDS) Control Rod X A2 Ability to (a) predict the impacts of the 3.8 Drive following malfunction or operations on the CRDS- and (b) based on those predictions, use

[Question 56] procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.02 Loss of power source to reactor trip breakers (CFR: 41.5/43.5/45.3/45.13) 015 (SF7 NI) Nuclear X K1 Knowledge of the physical connections 3.4 Instrumentation and/or cause effect relationships between the NIS and the following systems:

[Question 57]

K1.02 Vital ac systems (CFR: 41.2 to 41.9 / 45.7 to 45.8 016 (SF7 NNI) Nonnuclear X A3 Ability to monitor automatic operation of the 2.9 Instrumentation NNIS, including:

[Question 58] A3.02 Relationship between meter readings and actual parameter value (CFR: 41.7 / 45.5) 029 (SF8 CPS) Containment X A4 Ability to manually operate and/or monitor in 2.5 Purge the control room:

[Question 59] A4.01 Containment purge flow rate (CFR: 41.7 / 45.5 to 45.8) 034 (SF8 FHS) Fuel-Handling X K4 Knowledge of design feature(s) and/or 2.5 Equipment interlock(s) which provide for the following:

[Question 60] K4.02 Fuel movement (CFR: 41.7) 041 (SF4S SDS) Steam X K5 Knowledge of the operational implications of 2.6 Dump/Turbine Bypass Control the following concepts as the apply to the SDS:

[Question 61] K5.05 Basis for RCS design pressure limits (CFR: 41.5 / 45.7) 071 (SF9 WGS) Waste Gas X K3 Knowledge of the effect that a loss or 3.2 Disposal malfunction of the Waste Gas Disposal System will have on the following:

[Question 62]

K3.05 ARM and PRM systems (CFR: 41.7 / 45.6) 072 (SF7 ARM) Area Radiation X A1 Ability to predict and/or monitor changes in 3.4 Monitoring parameters (to prevent exceeding design limits) associated with operating the ARM system

[Question 63] controls including:

A1.01 Radiation levels (CFR: 41.5 / 45.5)

NUREG-1021, Revision 11 RO Page 11 of 13 BVPS Rev. AG

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2(RO) Continued System # / Name K K K K K K A A A A G K/A Topic(s)

IR #

1 2 3 4 5 6 1 2 3 4 075 (SF8 CW) Circulating Water X 2.4.31 Knowledge of annunciator alarms, 4.2 indications, or response procedures.

[Question 64]

(CFR: 41.10 / 45.3) 086 Fire Protection X K6 Knowledge of the effect of a loss or 2.6 malfunction on the Fire Protection System

[Question 65] following will have on the:

K6.04 Fire, smoke, and heat detectors (CFR: 41.7 / 45.7)

K/A Category Point Totals: 1 0 1 1 1 1 1 1 1 1 1 Group Point Total: 10 NUREG-1021, Revision 11 RO Page 12 of 13 BVPS Rev. AG

ES 401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Beaver Valley Unit 1 1LOT21 Date of Exam: 3/1 through 3/12/2021 RO SRO Only Category K/A# Topic IR # IR #

1. G2.1.15 2.1.15 Knowledge of administrative requirements for temporary management 2.7 Conduct directives, such as standing orders, night orders, Operations memos, etc.

of Operations (CFR: 41.10 / 45.12)

[Question 66]

G2.1.17 2.1.17 Ability to make accurate, clear, and concise verbal reports. 3.9 (CFR: 41.10 / 45.12 / 45.13)

[Question 67]

G2.1.4 2.1.4 Knowledge of individual licensed operator responsibilities related to shift 3.3 staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

(CFR: 41.10 / 43.2)

[Question 68]

Subtotal 3

2. G2.2.2 2.2.2 Ability to manipulate the console controls as required to operate the 4.6 Equipment facility between shutdown and designated power levels.

Control (CFR: 41.6 / 41.7 / 45.2)

[Question 69]

G2.2.42 2.2.42 Ability to recognize system parameters that are entry-level conditions for 3.9 Technical Specifications.

(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3)

[Question 70]

G2.2.44 2.2.44 Ability to interpret control room indications to verify the status and 4.2 operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45.12)

[Question 71]

Subtotal 3

3. G2.3.11 2.3.11 Ability to control radiation releases. 3.8 Radiation (CFR: 41.11 / 43.4 / 45.10)

Control [Question 72]

G2.3.12 2.3.12 Knowledge of radiological safety principles pertaining to licensed 3.2 operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12 / 45.9 / 45.10)

[Question 73]

Subtotal 2

4. G2.4.37 2.4.37 Knowledge of the lines of authority during implementation of the 3.0 Emergency emergency plan.

Procedures/ (CFR: 41.10 / 45.13)

Plan [Question 74]

G2.4.9 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of 3.8 coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13)

[Question 75]

Subtotal 2 Tier 3 Point Total 10 10 NUREG-1021, Revision 11 RO Page 13 of 13 BVPS Rev. AG

ES-401 PWR Examination Outline Form ES-401-2 Facility: Beaver Valley Unit 1 1LOT21 Date of Exam: 3/1 through 3/12/2021 SRO K/A Category Points SRO ONLY Points Tier Group K K K K K K A A A A G TOTAL A2 G* TOTAL 1 2 3 4 5 6 1 2 3 4

1 3 3 6 1.

Emergency 2

& 2 2 4 Abnormal Plant Tier Totals 5 5 10 Evolutions 1

2.

3 2 5 Plant Systems 2 2 1 3 Tier Totals 5 3 8 1 2 3 4

3. Generic Knowledge and 7 Abilities Category 1 2 2 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.)