ML23356A144
| ML23356A144 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 02/28/2023 |
| From: | NextEra Energy Seabrook |
| To: | Joseph Demarshall NRC/RGN-I/DORS/OB |
| References | |
| EPID L-2023-OLL-0003 | |
| Download: ML23356A144 (1) | |
Text
Form 3.2-1 Administrative Topics Outline Facility: Seabrook Station Date of Examination: 3/6/23 Examination Level:
RO Operating Test Number: 2023 Administrative Topic (Step 1)
Activity and Associated K/A (Step 2)
Type Code (Step 3)
Conduct of Operations A1 RO Determine License Status 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.
RO IR: 3.3 (R), (P)*
- randomly selected via random number generator Conduct of Operations A2 RO Calculate an Estimated Critical Position 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.
RO IR 4.3 (R), (D)
Equipment Control A3 RO Steady State Leak Rate Calculation 2.2.12 Knowledge of surveillance procedures.
RO IR 3.7 (R), (D)
Radiation Control A4 RO Determine COP Rad Monitor Setpoints 2.3.11 Ability to control radiation releases.
RO IR 3.8 (R), (M)
Emergency Plan NA NA
Form 3.2-1 Administrative Topics Outline Facility: Seabrook Station Date of Examination: 3/6/23 Examination Level:
SRO Operating Test Number: 2023 Administrative Topic (Step 1)
Activity and Associated K/A (Step 2)
Type Code (Step 3)
Conduct of Operations A1 SRO Determine Tech Spec AOT 2.1.20, Ability to interpret and execute procedure steps.
SRO IR: 4.6 (R), (M)
Conduct of Operations A2 SRO Approve an Estimated Critical Position Calculation 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.
SRO IR 4.6 (R), (D)
Equipment Control A3 SRO Approve Steady State Leak Rate Calculation 2.2.12 Knowledge of surveillance procedures.
SRO IR 4.1 (R), (M)
Radiation Control A4 SRO Approve COP Rad Monitor Setpoints 2.3.11 Ability to control radiation releases.
SRO IR 4.3 (R), (M)
Emergency Plan A5 SRO GE PAR Determination 2.4.44 Knowledge of emergency plan implementing procedures protective action recommendations.
SRO IR 4.4 (R), (D)
Form 3.2-2 Control Room/In-Plant Systems Outline Facility: Seabrook Station Date of Examination: 3/6/23 Exam Level: RO Operating Test Number: 2023 System/JPM Title Type Code Safety Function Control Room Systems
- a. Rapid Boration for Stuck Rods (L0059J)
(CRDS 001 A2.03 3.8/4.2)
A, D, S 1
- b. SI Termination (L0078J)
(ESFAS 013 A4.03 4.4/4.4)
A, D, EN, P, S
2
- c. Pressurizer Pressure Inst Failure (L0190J)
(Pressurizer Pressure Control System 010 A2.01 3.6/3.4)
D, S 3
- d. Establish SG Feed Flow from SUFP (L0179J)
(Steam Generator System 035 A2.07 4.2/3.9)
A, D, S 4
Primary
(Service Water 076 A2.08 3.5/3.1)
A, D, P, S 4
Secondary
- f. E-0 Attachment A (CBS Fails to Actuate)
(CBS 026 A2.03 3.9/4.1)
A, N, S 5
N, L, S 8
- h. Recover From A CRFRM Actuation (L0121J)
(CRV 050 A2.01 3.3/3.8)
D, S 9
In-Plant Systems
- i. Emergency Spent Fuel Pool Makeup (L0116J)
(SFP Cooling 033 A2.03 3.9/3.7)
D, E, R 8
- j. Reset Steam Driven EFW Pump Trip Valve (L0016J)
(EFW 061 A2.04 4.1/4.0)
D, E 4
Secondary
- k. Locally Start an Emergency Diesel Generator (L0124J)
(EDG 064 A2.23 3.2/3.7)
A, D, E 6
Form 3.2-2 Control Room/In-Plant Systems Outline Facility: Seabrook Station Date of Examination:3/6/23 Exam Level: SRO Operating Test Number: 2023 System/JPM Title Type Code Safety Function Control Room Systems
- a. Rapid Boration for Stuck Rods (L0059J)
(CRDS 001 A2.03 3.8/4.2)
A, D, S 1
- b. SI Termination (L0078J)
(ESFAS 013 A4.03 4.4/4.4)
A, D, EN, P, S
2
- c. Pressurizer Pressure Inst Failure (L0190J)
(Pressurizer Pressure Control System 010 A2.01 3.6/3.4)
D, S 3
- d. Establish SG Feed Flow from SUFP (L0179J)
(Steam Generator System 035 A2.07 4.2/3.9)
A, D, S 4
Primary
(Service Water 076 A2.08 3.5/3.1)
A, D, P, S 4
Secondary
A, N, S 5
N, L, S 8
- h.
In-Plant Systems
- i. Emergency Spent Fuel Pool Makeup (L0116J)
(SFP Cooling 033 A2.03 3.9/3.7)
D, E, R 8
- j. Reset Steam Driven EFW Pump Trip Valve (L0016J)
(EFW 061 A2.04 4.1/4.0)
D, E 4
Secondary
- k. Locally Start an Emergency Diesel Generator (L0124J)
(EDG 064 A2.23 3.2/3.7)
A, D, E 6
Form 3.3-1 Scenario Outline Facility:
Seabrook Station Scenario #:
1 Scenario Source:
Mod. 2018-1 Op. Test #:
2023 Examiners:
See crew rotations Applicants/
See crew rotations Operators:
Initial Conditions: 75% power.
Core burnup is 10,000 MWD/MTU. Boron concentration is 1142 ppm and CBD is at 190 steps.
Turnover: SW-P-41D removed from service for maintenance. Increase power to 90% at 10%/hr.
Pressurizer sprays are being forced with all four backup heaters energized.
Critical Tasks:
- 1. Restore offsite power to Bus 5 before performing step 18 of ECA-0.0.
- 2. Isolate RCP seals by closing CS-V-154, 158, 162 and 166 from the MCB before starting one charging pump. (PWROG CT-27)
- 3. Establish >500 gpm of EFW flow from the SUFP before meeting the criteria to establish bleed and feed within FR-H.1. (PWROG CT-23)
Event No Malf.
No.
Event Type*
Event Description 1
See SEG for sim malf.
codes PSO R BOP N US N Crew begins a power increase to 90% at 10%/hr.
2 BOP C, MC US C Running EHC pump trips, standby fails to start and must be manually started.
3 PSO C US C, TS A PORV leak. Can be isolated by closing the block valve.
4 BOP I, MC US I, TS Steam Generator A Pressure Instrument, FW-PT-514 fails low.
5 BOP C US C High turbine vibrations leading to requirement for reactor trip.
BOP required to lower turbine load.
6 PSO M BOP M US M On the reactor trip a loss of offsite power occurs. A EDG fails to start, B EDG has a large lube oil leak, TDEFW Pump Failure, Loss of all AC power. (CT-1) 7 BOP C, MC US C Offsite power is restored to busses 5 and 6. In ECA-0.1, FW-P-37B OC trip, SUFP must be placed in service. (CT-2) (CT-3)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
NUREG 1021, ES-2.3, Form 2.3-2, Target Quantitative Attributes per Scenario Section, specifies a Target Range of 1-2 for Table item #4, EOPs entered/requiring substantive actions. A detailed review of Scenario #1 confirms that the scenario is built to directly transition from E-0 to ECA-0.0, Loss of All AC Power, and that no Westinghouse Primary EOP (E-1, E-2, or E-3) will be entered/used. Consequently, a value of 0 will be assigned for Table Attribute Item 4 on Form 2.3-2, which is outside of the specified Target Range.
NUREG-1021, ES-3.3, Section B.2.g, EOP Operating Procedures Used, states Moreover, the primary scram response procedure that serves as the entry point for the EOPs is not counted. A value of 0 for Table Attribute Item 4 on Form 2.3-2 was determined to be acceptable by the Chief Examiner on the basis that: (a) Scenario #1 is a complex scenario that exercises two Contingency EOP Procedures; ECA-0.0 for the Loss of All AC Power, and ECA-0.1 for the Loss of All AC Power Recovery Without SI Required, (b) ECA-0.0 and ECA-0.1 both require the use of alternate decision paths and prioritization of actions within the EOPs to mitigate the Loss of All AC Power and Loss of Secondary Heat Sink conditions, and (c) ECA-0.0 and ECA-0.1 both have measurable actions that must be taken by the crew.
Scenario 1 Event Summary The crew will take the watch at 75% power. SW-P-41D, D Ocean Service Water Pump is tagged out for motor replacement. They will be directed via the turnover sheet to increase power to 90% at a rate of 10%/hr. The crew will develop a reactivity plan during their brief outside of the control room simulator before taking the watch.
During the power increase, the running EHC (Electro Hydraulic Control) pump will trip and the standby pump will fail to auto start. The crew should start the standby EHC pump via Skill of the Operator as defined in OP 9.2 or with the VPRO for, D7185, EHC Pump A Trouble.
The A PORV will develop a leak. The crew should enter OS1201.02, RCS Leak and close RC-V-122, the A PORV block valve to isolate the leakage. Tech Spec 3.4.4 action a applies.
Steam Generator A pressure transmitter, FW-PT-514 will fail low causing a decrease in controlling steam flow to the A Main Feed Regulating Valve (FRV). The crew should take manual control of the A FRV and the Main Feedwater Pumps to restore feedwater flow and use OS1235.04, SG Feed Flow - Steam Flow or Steam Pressure Instrument Failure AOP to swap controlling channels and restore SG level control to automatic. Tech Spec 3.3.2 action 18 and 3.3.3.6 Action a apply.
Turbine vibration levels will increase to the point that requires entry into ON1231.01, Turbine Generator High Vibration. Initially vibration levels will remain below the reactor trip requirement and the crew should reduce turbine load in attempt to lower vibration levels. When the crew places the load switch in lower vibrations will initially decrease briefly. Once the crew lowers turbine load, the vibration levels will increase to the reactor trip setpoint and the crew should trip the reactor, or an automatic trip will occur.
On the reactor trip a loss of offsite power will occur. The A EDG fails to automatically start and cannot be started. The B EDG has a large lube oil leak. Both emergency generators remain unavailable and will not repower the Emergency busses.
The crew will transition from E-0 to ECA-0.0. The SEPS engines will not start and are unavailable. The turbine driven emergency feedwater pump, FW-P-37A will fail to automatically start on the loss of power. It remains unavailable for the remainder of the scenario.
The crew will be able to restore power to both busses 5 and 6 via offsite power after competing step 6 of ECA-0.0. Once power is restored the crew will transition to ECA-0.1, Loss of all AC power Recovery without SI Required.
The crew must isolate the RCP seals before starting a charging pump. When attempting to start FW-P-37B in ECA-0.1 it will trip on overcurrent. The crew must place the SUFP in service per SUP-008 to restore secondary heat sink.
The scenario will complete once the crew takes actions in step 10 of ECA-0.1 so that FRPs again apply.
Scenario 1 Critical Tasks CT-1: Restore offsite power to Bus 5 before performing step 18, Depressurize Intact SGs to Reduce RCS Leakage of ECA-0.0.
Initiating Cue:
4160kV Emergency Bus 5 is deenergized. A EDG output breaker will not close and B EDG has a large lube oil leak.
Performance Feedback:
Bus 5 is reenergized.
Success Path:
Restore offsite power to Bus 5 through either the UAT or RATs via step 8 of ECA-0.0.
Measurable Performance Standard:
Crew reenergizes Bus 5 offsite power before performing step 18 of ECA-0.0. This is a preferred boundary condition as performance of step 18 in ECA-0.0 represents a change to the mitigative strategy in ECA-0.0.
Safety Significance:
Failure to energize an AC emergency bus constitutes mis-operation or incorrect crew performance in which the crew does not prevent degraded emergency power capacity.
Failure to perform the critical task also results in needless degradation of any barrier to fission product release, specifically of the RCS barrier at the point of the RCP seals.
Additionally, power to bus 5 is required to start the SUFP in this scenario. Ref: PWROG-14043.
CT-2: Isolate RCP seals by closing CS-V-154, 158, 162 and 166 from the MCB before starting one charging pump. (PWROG CT-27)
Initiating Cue:
Seal injection throttle valves have not been locally closed. Crew is performing step 1 of ECA-0.1.
Performance Feedback:
CS-V-154, 158, 162 and 166 are shown as closed on the MCB.
Success Path:
Close CS-V-154, 158, 162 and 166 from the MCB.
Measurable Performance Standard:
CS-V-154, 158, 162 and 166 are closed before starting a charging pump in step 3 of ECA-0.1. This is a preferred boundary condition as failure to isolate the seals before starting a charging pump will result in a severe challenge to the safety related function of the RCP seals which is preservation of RCS inventory.
Safety Significance:
Failure to isolate RCP seal injection before starting a charging pump, under the postulated plant conditions, can result in unnecessary and avoidable degradation of the RCS fission-product barrier, specifically at the point of the RCP seals. Ref: PWROG-14043.
CT-3: Establish >500 gpm of EFW flow from the SUFP before meeting the criteria to establish bleed and feed within FR-H.1. (PWROG CT-23)
Initiating Cue:
FW-P-37A, Turbine Driven EFW Pump fails to start and cannot be started. MS-V-129, the EFW turbine trip throttle valve will have tripped closed and cannot be reopened locally. When FW-P-37B is started in ECA-0.1 it will trip on overcurrent.
Performance Feedback:
Greater than 500 gpm of EFW flow from SUFP is established.
Success Path:
Start the SUFP per step 7 of ECA-0.1 and SUP-008.
Measurable Performance Standard:
EFW flow is established in ECA-0.1 from the SUFP before meeting the criteria to establish bleed and feed within FR-H.1. This is a preferred boundary condition.
Safety Significance:
Failure to establish the minimum required AFW flow rate, under the postulated plant conditions, results in adverse consequence or a significant degradation in the mitigative capability of the plant. In this case, the minimum required AFW flow rate can be established by performing the appropriate manual action. Failure to establish the required minimum EFW flow results in a challenge to core cooling.
"Per NUREG-1021, ES-3.3, if an applicants actions or inactions create a challenge to plant safety, those actions or inactions may form the basis for a Critical Task identified in the post scenario review.
Form 3.3-1 Scenario Outline Facility:
Seabrook Station Scenario #:
2 Scenario Source:
New Op. Test #:
2023 Examiners:
See crew rotations Applicants/
See crew rotations Operators:
Initial Conditions: 100% power.
Core burnup is 10,000 MWD/MTU. Boron concentration is 1032 ppm and CBD is at 230 steps.
Turnover: Maintain current power. SW-P-41D removed from service for maintenance. The plant is in a post-contingency back down to 1100 MWel net. Pre-brief ISO down power before taking the watch.
Critical Tasks:
- 1. Manually actuate main steamline isolation before transitioning to ECA-2.1. (PWROG CT-12)
Event No.
Malf.
No.
Event Type*
Event Description 1
See SEG for sim malf.
codes PSO C, MC US C A PCCW temperature controller fails to full cooling.
2 PSO R BOP N US N ISO down power to 1100 MW electric net in 30 minutes.
3 BOP C, MC US C, TS SW-P-41B trips on overcurrent. B Tower Actuation required.
During CT alignment, SW-V-5 fails to auto close and must be manually closed.
4 PSO I, MC US I, TS Pressurizer pressure channel RC-PT-455 fails high.
5 PSO M BOP M US M Loss of offsite power, fault of A SG main steam line outside containment upstream of MSIV. (CT-2) 6 BOP C, MC US C Main Steam Isolation fails to automatically actuate and must be manually actuated. (CT-1) 7 BOP C, MC US C Faulted A SG EFW flow control valves fail to auto isolate and must be manually closed. (CT-2) 8 PSO C, MC US C B cooling tower pump fails to automatically start and must be manually started.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Scenario 2 Event Summary The crew will take the watch at 100% power. SW-P-41D D Ocean Service Water Pump is tagged out for motor replacement.
The A Train PCCW temperature controller will fail to the full cooling position. The crew should take manual control and stabilize temperature and enter OS1212.01, PCCW Malfunction. The controller will remain in manual for the remainder of the scenario.
ISO NE will contact the crew and request that electrical output be lowered to 1100 MW within 30 minutes. This will require entry into OS1231.04, Rapid Down Power.
Once the down power is complete, SW-P-41B will trip on overcurrent. B train cooling tower actuation (TA) is required due to the unavailability of SW-P-41D. The crew should enter OS1216.01, Degraded Ultimate Heat Sink. During automatic realignment to the cooling tower, SW-V-5, SW Isolation to Secondary Loads will fail to close and must be manually closed. The loss of SW-P-41B (with 41D removed from service for maintenance) requires entry into TS action statement 3.7.4a.
Pressurizer pressure channel RC-PT-455 will fail high causing the pressurizer sprays to open.
The crew should close the spray valves and enter OS1206.01, Pressurizer Pressure/Component Failure. The controlling channel for pressurizer pressure control will be swapped and automatic control restored. Failure of RC-PT-455 requires entry into TS 3.3.1 action 6A and 3.3.2 actions 18 and 19.
A loss of offsite power will occur due to degraded grid conditions. On the trip, a fault will occur on the A SG main steam line upstream of the MSIV (Main Steam Isolation Valve), outside of containment. The MSI (Main Steam Isolation) fails to automatically actuate and must manually actuated by the BOP.
The faulted A SGs EFW flow control valves fail to auto isolate and must be manually closed during faulted steam generator isolation in E-2.
The B cooling tower pump will fail to restart once bus 6 is repowered from the B Emergency Diesel Generator and must be manually started.
The crew should process through E-2 to isolate the faulted A SG and transition to ES-1.1, SI Termination. The scenario will complete once the crew terminates SI in ES-1.1.
Scenario 2 Critical Tasks CT-1: Manually actuate main steamline isolation before transitioning to ECA-2.1. (PWROG CT-
- 12)
Initiating Cue:
SG pressures less than Auto MSI setpoint of 585 psig or step 1 of E-2.
Performance Feedback:
MSIVs are closed.
Success Path:
Manually actuate MSI or close individual MSIVs.
Measurable Performance Standard:
MSIVs must be closed before the crew transitions to ECA-2.1 via step 2a. RNO of E-2.
This is a preferred boundary condition.
Safety Significance:
Failure to close the MSIVs under the postulated plant conditions causes challenges to CSFs beyond those irreparably introduced by the postulated conditions due to an uncontrolled cooldown.
CT-2: Isolate the faulted SG before transitioning to FR-P.1. (PWROG CT-17)
Initiating Cue:
Step 4 of E-2 to isolate the faulted SG.
Performance Feedback:
EFW flow to the A SG is isolated (first element), MS-V-393 (second element) and MSD-V-44 (third element) are closed.
Success Path:
Isolate EFW flow to the A SG by closing FW-FV-4214A or B, closing MS-V-393 and MSD-V-44.
Measurable Performance Standard:
The faulted, A SG is isolated before the crew transitions to FR-P.1. This is a preferred boundary condition.
Safety Significance:
Failure to isolate a faulted SG that can be isolated causes challenges to CSFs beyond those irreparably introduced by the postulated conditions. Failure to isolate a faulted SG can result in challenges to the Integrity and Subcriticality CSFs.
"Per NUREG-1021, ES-3.3, if an applicants actions or inactions create a challenge to plant safety, those actions or inactions may form the basis for a Critical Task identified in the post scenario review.
Form 3.3-1 Scenario Outline Facility:
Seabrook Station Scenario #:
3 Scenario Source:
New Op. Test #:
2023 Examiners:
See crew rotations Applicants/
See crew rotations Operators:
Initial Conditions: 57% power.
Core burnup is 518 MWD/MTU. Boron concentration is 1546 ppm and CBD is at 186 steps.
Turnover: SW-P-41D removed from service for maintenance. Increase power to 70% at 10%/hr.
Pressurizer sprays are being forced with all four backup heaters energized.
Critical Tasks:
- 3. Transfer to Cold Leg Recirculation within 3 minutes. (PWROG CT-36)
Event No.
Malf.
No.
Event Type*
Event Description 1
See SEG for sim malf.
codes PSO R BOP N US N Crew begins a power increase to 70% at 10%/hr.
2 BOP I, MC US I, TS B steam generator controlling level channel fails low.
3 PSO C US C Inadvertent A Train T signal.
4 BOP I, MC US I MS-PT-507 fails high.
5 PSO C US C, TS 40 gpm RCS leak.
6 PSO M BOP M US M RCS leak increases to large LOCA requiring manual reactor trip and safety injection, the automatic SI is blocked. (CT-1) (CT-3) 7 BOP C, MC US C Loss of offsite power on reactor trip. A EDG fails to automatically start and must be manually started.
8 PSO C, MC US C Both RHR pumps fail to automatically start and must be manually started. (CT-2)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Scenario 3 Event Summary The crew will take the watch at 57% power. SW-P-41D D Ocean Service Water Pump is tagged out for motor replacement. They will be directed via the turnover sheet to increase power to 70% at a rate of 10%/hr.
During the power increase, the controlling level channel for the B steam generator will fail low.
This will cause the B FRV (Feed Regulating Valve) to fail open requiring the BOP to take manual control and restore B SG level. The crew should enter OS1235.03, Steam Generator Level Instrument Failure, swap to a backup channel and restore automatic control. B SG level channel failure requires entry into TS 3.3.1, table 3.3-1, item 13, table action 6A and 3.3.2, action c, table 3.3-3, item 5b, 6a, 7c and 10c, table action 18.
An inadvertent T signal will occur. Letdown will isolate and the PSO should reduce charging flow to minimum. The crew should enter OS1205.01, Inadvertent Phase A Containment Isolation and restore letdown.
MS-PT-507, Main Steam Header Pressure will fail high causing the Main Feedwater pumps to increase speed, raising SG levels. The crew should take manual control of the main feed water pumps and enter ON1230.01, MS-PT-507 or 508 Pressure Instrument Failure.
A 40 gpm RCS leak will occur requiring the PSO to lower letdown flow to stabilize pressurizer level. The leak requires entry into TS action statement 3.4.6.2.b action b. Once the crew has entered OS1201.02, RCS Leak and quantified the leak rate, the leak will propagate to a LOCA. The crew should manually trip the reactor, actuate SI and transition to E-0. The automatic safety injection is blocked.
On the reactor trip a loss of offsite power will occur. The A Emergency Diesel Generator will fail to automatically start and must be manually started by the BOP.
Both RHR (Residual Heat Removal) pumps will fail to auto start and must be manually started by the PSO.
The crew should process through E-1, transition to ES-1.3 and establish cold leg recirculation.
The exam will terminate once ES-1.3 is complete.
Scenario 3 Critical Tasks CT-1: Energize Bus 5 and manually actuate SI before transitioning to FR-C.1. (PWROG CT-2).
Initiating Cue:
Pressurizer level continues to decrease with letdown isolated.
Performance Feedback:
Bus 5 is energized and both trains of SI are initiated.
Success Path:
Start the A Emergency Diesel Generator and manually initiate SI.
Measurable Performance Standard:
Bus 5 is energized and SI is actuated before the crew transitions to FR-C.1. This is a preferred boundary condition.
Safety Significance:
Failure to manually actuate SI under the postulated conditions constitutes misoperation or incorrect crew performance in which the crew does not prevent degraded emergency core cooling system capacity. Bus 5 must be energized so that both SI-P-6A and B will be started on the manual safety injection. Both SI pumps are required to maintain core cooling and avoid a requirement to transition to FR-C.1.
CT-2: Manually start at least one RHR pump before transitioning to ECA-1.1. (PWROG CT-5)
Initiating Cue:
Following the reactor trip and SI, neither RHR pump is running.
Performance Feedback:
At least one RHR pump is running.
Success Path:
Manually start at least one RHR pump.
Measurable Performance Standard:
At least one RHR pump is running before the crew transitions to ECA-1.1. This is a preferred boundary condition.
Safety Significance:
Failure to manually start at least one low-head ECCS pump under the postulated conditions constitutes misoperation or incorrect crew performance in which the crew does not prevent degraded emergency core cooling system capacity. Additionally a failure to manually start RHR pumps will prevent the ability to perform cold leg recirculation.
CT-3: Transfer to Cold Leg Recirculation within 3 minutes. (PWROG CT-36)
Initiating Cue:
RWST level decreases to recirculation swap over setpoint. Alarm for RWST Level Lo Lo actuates.
Performance Feedback:
SI is reset, CBS-V-8 and 14 are open with CBS-V-2 and 5 closed.
Success Path:
Perform steps 1 through 3 of ES-1.3.
Measurable Performance Standard:
Steps 1 through 3 of ES-1.3 are performed within 3 minutes of receiving alarm RWST Level Lo Lo. This is a preferred boundary condition.
Safety Significance:
Failure to manually align for cold leg recirculation will result in a loss of long term core cooling.
"Per NUREG-1021, ES-3.3, if an applicants actions or inactions create a challenge to plant safety, those actions or inactions may form the basis for a Critical Task identified in the post scenario review.
Form 3.3-1 Scenario Outline Facility:
Seabrook Station Scenario #:
Spare (Not Used)
Scenario Source:
New Op. Test #:
2023 Examiners:
See crew rotations Applicants/
See crew rotations Operators:
Initial Conditions: 100% power.
Core burnup is 10,000 MWD/MTU. Boron concentration is 1032 ppm and CBD is at 230 steps.
Turnover: Maintain current power. SW-P-41D removed from service for maintenance.
Critical Tasks:
- 1. Isolate feedwater flow into and steam flow from the ruptured SG before a transition to ECA-3.1 occurs. (PWROG CT-18)
- 2. Establish/maintain an RCS temperature so that transition from E-3 does not occur because the RCS temperature is in either of the following conditions (PWROG CT-19):
Too high to maintain 60°F of subcooling requiring transition to ECA-3.1.
OR Too low causing Red or Orange path on the sub criticality and/or the integrity CSF.
Event No.
Malf.
No.
Event Type*
Event Description 1
See SEG for sim malf.
codes BOP C, MC US C SCCW pump trips, standby fails to start and must be manually started.
2 PSO I, MC US I, TS Controlling pressurizer level channel, RC-LT-459 fails low.
3 PSO C US C, TS 20 gpm steam generator tube leak on A SG.
4 PSO R BOP C US C Plant down power to <50% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
5 PSO M BOP M US M A SG tube leak propagates into 500 gpm tube rupture, requiring manual reactor trip and safety injection. (CT-1) (CT-2) 6 BOP C, MC US C Main generator breaker fails to open on the turbine trip and must be manually opened.
7 PSO C, MC US C SI-P-6A fails to automatically start and must be manually started.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Scenario Spare Event Summary The crew will take the watch at 100% power. SW-P-41D D Ocean Service Water Pump is tagged out for motor replacement. The crew will be directed to maintain power.
The A SCCW pump will trip. The standby will fail to auto start and should be started by the BOP. The Crew should enter ON1237.01, Loss of Secondary Component Cooling.
RC-LT-459, the controlling channel for pressurizer level will fail low causing letdown to isolate.
The PSO should reduce charging to minimum. The crew should enter OS1201.07, Pressurizer Level Instrument Failure to swap controlling channels and restore letdown. RC-LT-459 failure requires entry into TS 3.3.1 action 6A.
A 20 gpm steam generator tube leak (SGTL) will occur on the A SG. The PSO should lower letdown flow to stabilize pressurizer level. The 20 gpm SGTL requires entry into TS action statement 3.4.6.2.a. The crew should enter OS1227.01, Steam Generator Tube Leak and once the leak is quantified, they should reduce plant power to less than 50% within one hour by implementing OS1231.04, Rapid Down Power. During the down power the SGTL will propagate into a 500 gpm tube rupture, requiring the crew to start the second charging pump and eventually trip the reactor and actuate SI.
On the turbine trip the main generator breaker will fail to automatically open and must be manually opened by the BOP.
Following the reactor trip and SI, SI-P-6A will fail to automatically start and must be manually started by the PSO.
The crew will process through E-0 to E-3, performing the required RCS cooldown. The exam will terminate once the cooldown is complete.
Scenario Spare Critical Tasks CT-1: Isolate feedwater flow into and steam flow from the ruptured SG before a transition to ECA-3.1 occurs. (PWROG CT-18)
Initiating Cue:
Step 3 of E-3.
Performance Feedback:
A MSIV, A ASDV, SB-V-9, MSD-V-44 and MS-V-393 are all closed.
Success Path:
Student closes valves per step 3 of E-3.
Measurable Performance Standard:
A SG is isolated with A MSIV, A ASDV, SB-V-9, MSD-V-44 and MS-V-393 closed before a transition to ECA-3.1 occurs. This is a preferred boundary condition.
Safety Significance:
Failure to isolate the ruptured SG causes a loss of differential pressure between the ruptured SG and the intact SGs. The fact that the crew allows the differential pressure to dissipate and, as a result, are then forced to transition to a contingency procedure constitutes an incorrect performance that necessitates the crew taking compensating action that would complicate the event mitigation strategy.
CT-2: Establish/maintain an RCS temperature so that transition from E-3 does not occur because the RCS temperature is in either of the following conditions: too high to maintain 60°F of subcooling requiring transition to ECA-3.1, or too low causing Red or Orange path on the sub criticality and/or the integrity CSF. (PWROG CT-19)
Initiating Cue:
Step 7 of E-3.
Performance Feedback:
Core exit temperatures are at required value as determined by step 7 of E-3.
Success Path:
Use condenser Steam Dumps to reach target temperature.
Measurable Performance Standard:
Establish/maintain an RCS temperature so that transition from E-3 does not occur because the RCS temperature is in either of the following conditions: too high to maintain 60°F of subcooling requiring transition to ECA-3.1, or too low causing Red or
Orange path on the sub criticality and/or the integrity CSF. This is a preferred boundary condition.
Safety Significance:
Failure to establish and maintain the correct RCS temperature during a SGTR leads to a transition from E-3 to an ECA procedure. The failure constitutes an incorrect performance that necessitates the crew taking compensating action that would complicate the event mitigation strategy.
"Per NUREG-1021, ES-3.3, if an applicants actions or inactions create a challenge to plant safety, those actions or inactions may form the basis for a Critical Task identified in the post scenario review.
Form 4.1-PWR Pressurized-Water Reactor Examination Outline Facility:
Seabrook K/A Catalog Rev. 3 Rev.
4 Date of Exam:
03/06/2023 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
2 3
3 3
4 3
18 3
3 6
2 2
1 1
2 1
1 8
2 2
4 Tier Totals 4
4 4
5 5
4 26 5
5 10
- 2.
Plant Systems 1
3 2
2 3
3 3
2 2
3 3
2 28 3
2 5
2 1
1 0
1 1
1 1
1 0
1 1
9 0
2 1
3 Tier Totals 4
3 2
4 4
4 3
3 3
4 3
37 5
3 8
- 3.
Generic Knowledge and Abilities Categories CO EC RC EM 6
CO EC RC EM 7
2 2
1 1
2 2
1 2
- 4. Theory Reactor Theory Thermodynamics 6
3 3
Notes: CO =
EM =
Conduct of Operations; EC = Equipment Control; RC = Radiation Control; Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan.
These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan.
ES-4.1-PWR PWR Examination Outline (Seabrook)
Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)
Item E/APE # / Name K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR Q#
1 (000007) (EPE 7; BW E02 & E10; CE E02) Reactor Trip, Stabilization, Recovery X
G2.1.45 - Ability to identify and interpret diverse indications to validate the response of another indication.
(CFR: 41.7 / 43.5 / 45.4) 4.3 1
2 (000008) (APE 8)
Pressurizer Vapor Space Accident X
AA2.12 - Ability to determine and/or interpret the following as they apply to a Pressurizer Vapor Space Accident: Pressurizer level.
(CFR: 43.5 / 45.13) 3.5 2
3 (000009) (EPE 9)
Small Break LOCA X
EK3.21 - Knowledge of the reasons for the following responses and/or actions as they apply to a Small-Break LOCA: Actions contained in an EOP for a small-break LOCA.
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 4.1 3
4 (000015) (APE 15)
Reactor Coolant Pump Malfunctions X
AK2.07 - Knowledge of the relationship between Reactor Coolant Pump Malfunctions and the following systems or components: RCP seals.
(CFR: 41.7 / 45.7) 3.8 4
5 (000025) (APE 25)
Loss of Residual Heat Removal System X
AK1.02 - Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to the Loss of the Residual Heat Removal System: Core cooling.
(CFR: 41.8 / 41.10 / 45.3) 4.2 5
6 (000026) (APE 26)
Loss of Component Cooling Water X
AA1.01 - Ability to operate and/or monitor the following as they apply to Loss of Component Cooling Water: CCW temperature indications.
(CFR: 41.5 / 41.7 / 45.5 to 45.8) 3.3 6
7 (000027) (APE 27)
Pressurizer Pressure Control System Malfunction X
AA1.01 - Ability to operate and/or monitor the following as they apply to a Pressurizer Pressure Control System Malfunction: PZR heaters, sprays, and PORVs.
(CFR: 41.7 / 45.5 / 45.6) 3.8 7
8 (000029) (EPE 29)
Anticipated Transient Without Scram X
EK1.01 - Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Anticipated Transient Without Scram: Reactor nucleonic and thermohydraulic behavior.
(CFR: 41.8 / 41.10 / 45.3) 3.8 8
9 (000038) (EPE 38)
Steam Generator Tube Rupture X
EK3.01 - Knowledge of the reasons for the following responses and/or actions as they apply to a Steam Generator Tube Rupture:
Controlling RCS pressure for equalizing pressure on primary and secondary sides of ruptured S/G..
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 4.1 9
10 (000040) (APE 40; BW E05; CE E05; W E12) Steam Line Rupture - Excessive Heat Transfer X
(APE 40) AK2.12 - Knowledge of the relationship between Steam Line Rupture and the following systems or components: MRSS.
(CFR: 41.7 / 45.7) 3.0 10 11 (000056) (APE 56)
AA2.33 - Ability to determine and/or interpret the following as they apply to Loss of Offsite Power: Status of bus voltage indication.
(CFR: 43.5 / 45.13) 3.5 11 12 (000057) (APE 57)
Loss of Vital AC Instrument Bus X
AA1.07 - Ability to operate and/or monitor the following as they apply to Loss of Vital AC Electrical Instrument Bus: Interlocks in effect on loss of AC vital electrical instrument bus that must be bypassed to restore normal equipment operation.
(CFR: 41.7 / 45.5 / 45.6) 3.7 12 13 (000062) (APE 62)
Loss of Service Water X
AA2.06 - Ability to determine and/or interpret the following as they apply to Loss of Service Water: The length of time after the loss of SWS flow to a component before that component may be damaged.
(CFR: 43.5 / 45.13) 3.3 13
14 (000065) (APE 65)
Loss of Instrument Air X
AK3.10 - Knowledge of the reasons for the following responses and/or actions as they apply to Loss of Instrument Air: Isolation of leaking components or headers.
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.3 14 15 (000077) (APE 77)
Generator Voltage and Electric Grid Disturbances X
G2.2.44 - Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.
(CFR: 41.5 / 43.5 / 45.12) 4.2 15 16 (W E04) LOCA Outside Containment X
EA2.03 - Ability to determine and/or interpret the following as they apply to LOCA Outside Containment: RCS pressure.
(CFR: 41.10 / 43.5 / 45.13) 3.9 16 17 (W E11) Loss of Emergency Coolant Recirculation X
EK2.15 - Knowledge of the relationship between Loss of Emergency Coolant Recirculation and the following systems or components: SGS.
(CFR: 41.7 / 41.8 / 45.2 / 45.4) 3.1 17 18 (BW E04; W E05)
Inadequate Heat Transfer - Loss of Secondary Heat Sink X
G2.4.21 - Knowledge of the parameters and logic used to assess the status of emergency operating procedures critical safety functions or shutdown critical safety functions.
(CFR: 41.7 / 43.5 / 45.12) 4.0 18 19 (000011) (EPE 11)
Large Break LOCA X
EA2.17 - Ability to determine and/or interpret the following as they apply to a Large-Break LOCA: Containment pressure, leakage, and/or temperature.
(CFR: 43.5 / 45.13) 4.1 76 20 (000022) (APE 22)
Loss of Reactor Coolant Makeup X
AA2.01 - Ability to determine and/or interpret the following as they apply to Loss of Reactor Coolant Makeup: Whether charging line leak exists.
(CFR: 43.5 / 45.13) 3.8 77 21 (000054) (APE 54; CE E06) Loss of Main Feedwater X
G2.4.20 - Knowledge of the operational implications of emergency and abnormal operating procedures warnings, cautions, and notes.
(CFR: 41.10 / 43.5 / 45.13) 4.3 78 22 (000055) (EPE 55)
Station Blackout X
G2.1.20 - Ability to interpret and execute procedure steps.
(CFR: 41.10 / 43.5 / 45.12) 4.6 79 23 (000058) (APE 58)
Loss of DC Power X
AA2.03 - Ability to determine and/or interpret the following as they apply to Loss of DC Power: Impact on ability to operate and monitor plant systems.
(CFR: 43.5 / 45.13) 4.0 80 24 (000062) (APE 62)
Loss of Service Water X
G2.2.45 - Ability to determine and/or interpret TS action statements of greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (SRO Only).
(CFR: 43.2 / 43.5 / 45.3) 4.7 81 K/A Category Totals:
2 3
3 3
7 6
Group Point Total:
24 ES-4.1-PWR PWR Examination Outline (Seabrook)
Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)
Item E/APE # / Name K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR Q#
25 (000024) (APE 24)
Emergency Boration X
AK2.09 - Knowledge of the relationship between Emergency Boration and the following systems or components: RCS.
(CFR: 41.7 / 45.7) 3.9 19
26 (000028) (APE 28)
Pressurizer Level Control Malfunction X
AK1.01 - Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to a Pressurizer Level Control Malfunction: PZR reference leg leak abnormalities.
(CFR: 41.7 / 41.8 / 41.10 / 45.3) 3.5 20 27 (000059) (APE 59)
Accidental Liquid Radwaste Release X
AK3.01 - Knowledge of the reasons for the following responses and/or actions as they apply to an Accidental Liquid Radwaste Release: Termination of a release of radioactive liquid.
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.7 21 28 (000061) (APE 61)
Area Radiation Monitoring System Alarms X
AA1.01 - Ability to operate and/or monitor the following as they apply to Area Radiation Monitoring System Alarms: Systems or components automatically actuated by ARM signals.
(CFR: 41.7 / 45.5 / 45.6) 3.5 22 29 (000067) (APE 67)
Plant Fire On Site X
AK1.02 - Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Plant Fire On Site: Fire-fighting methods for each type of fire.
(CFR: 41.8 / 41.10 / 45.3) 3.0 23 30 (000074) (EPE 74; W E06 & E07)
Inadequate Core Cooling X
(WE06) EA2.10 - Ability to determine and/or interpret the following as they apply to Degraded Core Cooling: Reactor vessel level.
(CFR: 41.10 / 43.5 / 45.13) 3.8 24 31 (W E15)
Containment Flooding X
G2.1.19 - Ability to use available indications to evaluate system or component status.
(CFR: 41.10 / 45.12) 3.9 25 32 (BW E08; W E03)
LOCA Cooldown -
Depressurization X
(WE03) EA1.20 - Ability to operate and/or monitor the following as they apply to LOCA Cooldown and Depressurization: AFW system.
(CFR: 41.5 to 41.8 / 45.5 to 45.8) 3.7 26 33 (000003) (APE 3)
Dropped Control Rod X
G2.1.7 - Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.
(CFR: 41.5 / 43.5 /45.12 / 45.13) 4.7 82 34 (000078) (APE 78*)
RCS Leak X
G2.2.25 - Knowledge of the bases in TS for limiting conditions for operation and safety limits (SRO Only).
(CFR: 43.2) 4.2 83 35 (W E13) Steam Generator Overpressure X
EA2.01 - Ability to determine and/or interpret the following as they apply to Steam Generator Overpressure: Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
(CFR: 41.10 / 43.5 / 45.13) 3.8 84 36 (CE A11**; W E08)
RCS Overcooling -
Pressurized Thermal Shock X
(WE08) EA2.05 - Ability to determine and/or interpret the following as they apply to Pressurized Thermal Shock: RCS pressure, temperature, and/or PZR level.
(CFR: 41.10 / 43.5 / 45.13) 3.9 85 000001 (APE 1)
Continuous Rod Withdrawal / 1 000005 (APE 5)
Inoperable/Stuck Control Rod / 1 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 000060 (APE 60)
Accidental Gaseous Radwaste Release /
9 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000076 (APE 76)
High Reactor Coolant Activity / 9 (W E01 & E02)
Rediagnosis & SI Termination / 3 (W E16) High Containment Radiation /9 (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 E09; CE A13**;
W E09 & E10)
Natural Circulation/4 (BW E13 & E14)
EOP Rules and Enclosures (CE A16) Excess RCS Leakage / 2
(CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation /
LOOP / Blackout / 4 K/A Category Totals:
2 1
1 2
3 3
Group Point Total:
12 ES-4.1-PWR PWR Examination Outline (Seabrook)
Plant SystemsTier 2/Group 1 (RO/SRO)
Item System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR Q#
37 (003) (SF4P RCP)
Reactor Coolant Pump X
G2.4.31 - Knowledge of annunciator alarms, indications, or response procedures.
(CFR: 41.10 / 45.3) 4.2 27 38 (003) (SF4P RCP)
Reactor Coolant Pump X
K4.07 - Knowledge of Reactor Coolant Pump System design features and/or interlocks that provide for the following: Minimizing RCS leakage (mechanical seals).
(CFR: 41.7).
3.8 28 39 (004) (SF1; SF2 CVCS) Chemical and Volume Control X
A4.05 - Ability to manually operate and/or monitor in the control room:
Letdown pressure and temperature control valves.
(CFR: 41.5 to 41.7 / 45.5 to 45.8) 3.7 29 40 (004) (SF1; SF2 CVCS) Chemical and Volume Control X
K6.26 - Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Chemical and Volume Control System: Methods of pressure control for solid plant (relief valves and water inventory).
(CFR: 41.5 to 41.7 / 45.7) 3.9 30 41 (005) (SF4P RHR)
A1.05 - Ability to predict and/or monitor changes in parameters associated with operation of the Residual Heat Removal System, including: Detection of RHR leak.
(CFR: 41.5 / 45.5) 3.7 31 42 (006) (SF2; SF3 ECCS) Emergency Core Cooling X
A3.03 - Ability to monitor automatic operation of the Emergency Core Cooling System, including: ECCS ESFAS/ESAS-operated valves.
(CFR: 41.7 / 45.5) 4.2 32
43 (006) (SF2; SF3 ECCS) Emergency Core Cooling X
K5.13 - Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Emergency Core Cooling System: Hot-leg injection.
(CFR: 41.5 / 45.7) 3.6 33 44 (007) (SF5 PRTS)
Pressurizer Relief/Quench Tank X
A2.08 - Ability to (a) predict the impacts of the following on the Pressurizer Relief Tank/Quench Tank System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Abnormal level in the PRT/quench tank.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.1 34 45 (008) (SF8 CCW)
Component Cooling Water X
K3.10 - Knowledge of the effect that a loss or malfunction of the Component Cooling Water System will have on the following systems or system parameters: ECCS.
(CFR: 41.4 to 41.7 / 45.7 to 45.9) 4.2 35 46 (010) (SF3 PZR PCS) Pressurizer Pressure Control X
K3.03 - Knowledge of the effect that a loss or malfunction of the Pressurizer Pressure Control System will have on the following systems or system parameters:
(CFR: 41.7 / 45.6) 4.2 36 47 (012) (SF7 RPS)
Reactor Protection X
K6.12 - Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions, on the Reactor Protection System: 120 V vital/instrument power system.
(CFR: 41.7 / 45.7) 3.9 37 48 (013) (SF2 ESFAS)
Engineered Safety Features Actuation X
K4.09 - Knowledge of Engineered Safety Features Actuation System design features and/or interlocks that provide for the following:
Spurious trip protection.
(CFR: 41.2 / 41.6 / 41.7) 3.4 38 49 (022) (SF5 CCS)
Containment Cooling X
A1.02 - Ability to predict and/or monitor changes in parameters associated with operation of the Containment Cooling System, including: Containment pressure.
(CFR: 41.5 / 45.5) 3.9 39
50 (022) (SF5 CCS)
Containment Cooling X
K1.01 - Knowledge of the physical connections and/or cause and effect relationships between the Containment Cooling System and the following systems: Cooling water system.
(CFR: 41.9 / 45.7 / 45.8) 3.6 40 51 (026) (SF5 CSS)
A3.01 - Ability to monitor automatic features of the Containment Spray System, including: Pump starts and correct valve positioning.
(CFR: 41.7 / 45.7) 4.1 41 52 (026) (SF5 CSS)
K2.02 - Knowledge of electrical power supplies to the following:
Motor-operated valves.
(CFR: 41.7) 3.6 42 53 (039) (SF4S MSS)
Main and Reheat Steam X
K4.05 - Knowledge of Main and Reheat Steam System design features and/or interlocks that provide for the following: Automatic isolation of steam line.
(CFR: 41.7).
4.0 43 54 (039) (SF4S MSS)
Main and Reheat Steam X
K6.12 - Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions, on the Main and Reheat Steam System: MSIVs.
(CFR: 41.7 / 45.7) 3.8 44 55 (059) (SF4S MFW)
Main Feedwater X
A3.08 - Ability to monitor automatic features of the Main Feedwater System, including: S/G water LCS.
(CFR: 41.7 / 45.5) 3.8 45 56 (061) (SF4S AFW)
Auxiliary/Emergency Feedwater X
COMPONENT (Controllers and Positioners): 191003 K1.01: -
Function and operation of flow controller in manual and automatic modes.
(CFR: 41.7) 3.2 46 57 (062) (SF6 ED AC)
AC Electrical Distribution X
K2.01 - Knowledge of electrical power supplies to the following:
Major bus or motor control center power supplies.
(CFR: 41.7) 3.8 47 58 (063) (SF6 ED DC)
DC Electrical Distribution X
A4.02 - Ability to manually operate and/or monitor the control room:
Load shedding.
(CFR: 41.7 / 45.5 to 45.8) 3.6 48
59 (064) (SF6 EDG)
A4.05 - Ability to manually operate and/or monitor in the control room:
Transfer of EDG control between manual and automatic.
(CFR: 41.7 / 45.5 to 45.8) 3.6 49 60 (064) (SF6 EDG)
K1.07 - Knowledge of the physical connections and/or cause and effect relationships between the Emergency Diesel Generators and the following systems: EDG building ventilation system.
(CFR: 41.3 to 41.8 / 45.7 / 45.8) 3.3 50 61 (073) (SF7 PRM)
Process Radiation Monitoring X
K1.04 - Knowledge of the physical connections and/or cause and effect relationships between the Process Radiation Monitoring System and the following systems: S/GB.
(CFR: 41.7 to 41.9 / 41.11 / 45.8 /
45.9) 3.4 51 62 (076) (SF4S SW)
A2.11 - Ability to (a) predict the impacts of the following on the Service Water System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Strainer failure.
(CFR: 41.5 / 41.1 / 43.5 / 45.3 / 45.6
/ 45.13) 3.5 52 63 (078) (SF8 IAS)
Instrument Air X
K5.03 - Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Instrument Air System: Loss of instrument air.
(CFR: 41.5 / 45.7) 3.9 53 64 (103) (SF5 CNT)
Containment X
K5.01 - Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Containment System:
Containment isolation/containment integrity.
(CFR: 41.5 / 45.7) 4.1 54 65 (005) (SF4P RHR)
G2.2.37 - Ability to determine operability or availability of safety-related equipment (SRO Only).
(CFR: 43.2 / 43.5 / 45.12) 4.6 86
66 (013) (SF2 ESFAS)
Engineered Safety Features Actuation X
A2.05 - Ability to (a) predict the impacts of the following on the Engineered Safety Features Actuation System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Loss of DC control power.
(CFR: 41.5 / 41.7 / 41.10 / 43.5 /
45.3 / 45.13) 3.8 87 67 (062) (SF6 ED AC)
AC Electrical Distribution X
A2.21 - Ability to (a) predict the impacts of the following on the AC Electrical Distribution System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Loss of vital AC electrical instrument buses.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 4.3 88 68 (073) (SF7 PRM)
Process Radiation Monitoring X
A2.01 - Ability to (a) predict the impacts of the following on the Process Radiation Monitoring System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: PRM component failures.
(CFR: 41.5 / 43.5 / 45.3 / 45.5 / 45.8 to 45.9 / 45.13) 3.1 89 69 (103) (SF5 CNT)
Containment X
G2.4.18 - Knowledge of the specific bases for emergency and abnormal operating procedures.
(CFR: 41.10 / 43.1 / 45.13) 4.0 90 (025) (SF5 ICE) Ice Condenser (053) (SF1; SF4P ICS*) Integrated Control K/A Category Totals:
3 2
2 3
3 3
2 5
3 3
4 Group Point Total:
33 ES-4.1-PWR PWR Examination Outline (Seabrook)
Plant SystemsTier 2/Group 2 (RO/SRO)
Item System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR Q#
70 (001) (SF1 CRDS)
Control Rod Drive X
K1.03 - Knowledge of the physical connections and/or cause and effect relationships between the Control Rod Drive System and the following systems: CRDM.
(CFR: 41.2 / 41.3 / 41.5 to 45.7 /
45.5 to 45.8) 3.7 55 71 (011) (SF2 PZR LCS) Pressurizer Level Control X
K2.02 - Knowledge of electrical power supplies to the following: PZR heaters.
(CFR: 41.7) 3.3 56
72 (014) (SF1 RPI) Rod Position Indication X
A1.04 - Ability to predict and/or monitor changes in parameters associated with operation of the Rod Position Indication System, including: Axial and/or radial power distribution.
(CFR: 41.5 to 41.7 / 45.5) 3.8 57 73 (015) (SF7 NI)
Nuclear Instrumentation X
COMPONENT (Sensors and Detectors - Nuclear Instrumentation): 191002 K1.17: -
Effects of core voiding on neutron detection.
(CFR: 41.7) 3.5 58 74 (016) (SF7 NNI)
Nonnuclear Instrumentation X
K5.01 - Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Nonnuclear Instrumentation System: Separation of control and protection circuits.
(CFR: 41.5 / 45.7) 3.3 59 75 (029) (SF8 CPS)
Containment Purge X
K4.03 - Knowledge of Containment Purge System design features and/or interlocks that provide for the following: Automatic purge isolation.
(CFR: 41.7) 3.6 60 76 (035) (SF4P SG)
K6.11 - Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Steam Generator System: MRSS.
(CFR: 41.7 / 45.7) 3.3 61 77 (045) (SF4S MTG)
Main Turbine Generator X
A2.17 - Ability to (a) predict the impacts of the following on the Main Turbine Generator System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Malfunction of EHC.
(CFR: 41.5 / 43.5 / 45.3 / 45.5) 3.5 62 78 (075) (SF8 CW)
Circulating Water X
A4.02 - Ability to manually operate and/or monitor in the control room:
Circulating water pump.
(CFR:41.4 / 41.7 / 45.5 / 45.8) 3.1 63 79 (034 (SF8 FHS)
Fuel Handling Equipment X
G2.1.40 - Knowledge of refueling administrative requirements.
(CFR: 41.10 / 43.5 / 43.6 / 45.13) 3.9 91 80 (050) (SF9 CRV*)
Control Room Ventilation X
A2.03 - Ability to (a) predict the impacts of the following on the Control Room Ventilation System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Initiation/reconfiguration failure.
(CFR: 41.5 / 43.5 / 45.6) 3.8 92
81 (086) (SF8 FPS)
Fire Protection X
A2.05 - Ability to (a) predict the impacts of the following on the Fire Protection System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Fire in the plant.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.8 93 (002) (SF2; SF4P RCS) Reactor Coolant (017) (SF7 ITM) In Core Temperature Monitor (027) (SF5 CIRS)
Containment Iodine Removal (028) (SF5 HRPS)
Hydrogen Recombiner and Purge Control (033) (SF8 SFPCS)
Spent Fuel Pool Cooling (041) (SF4S SDS)
Steam Dump/Turbine Bypass Control (055) (SF4S CARS)
Condenser Air Removal (056) (SF4S CDS)
Condensate (068) (SF9 LRS)
Liquid Radwaste (071) (SF9 WGS)
Waste Gas Disposal (072) (SF7 ARM)
Area Radiation Monitoring (079) (SF8 SAS**)
Station Air K/A Category Totals:
1 1
0 1
1 1
1 3
0 1
2 Group Point Total:
12 Form 4.1-COMMON Common Examination Outline ES-4.1-COMMON COMMON Examination Outline (Seabrook)
Facility:
Seabrook Date of Exam:
03/06/2023 Generic Knowledge and Abilities Outline (Tier 3) (RO/SRO)
Category K/A #
Topic RO SRO-Only Item #
IR Q#
IR Q#
- 1.
Conduct of Operations G2.1.3 Knowledge of shift or short-term relief turnover practices.
(CFR: 41.10 / 45.13) 82 3.7 64 G2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, or switches.
(CFR: 41.10 / 45.1 / 45.12) 83 4.1 65 G2.1.1 Knowledge of conduct of operations requirements.
(CFR: 41.10 / 43.10 / 45.13) 84 4.2 94 G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.
(CFR: 41.1 / 41.5 / 41.10 / 43.6 / 45.6) 85 4.6 95 Subtotal N/A 2
N/A 2
- 2.
Equipment Control G2.2.6 Knowledge of the process for making changes to procedures.
(CFR: 41.10 / 43.3 / 45.13) 86 3.0 66 G2.2.13 Knowledge of tagging and clearance procedures.
(CFR: 41.10 / 43.1 / 45.13) 87 4.1 67 G2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.
(CFR: 41.10 / 43.5 / 45.13) 88 3.8 96 G2.2.21 Knowledge of pre-and post-maintenance operability requirements.
(CFR: 41.10 / 43.2) 89 4.1 97 Subtotal N/A 2
N/A 2
- 3.
Radiation Control G2.3.5 Ability to use RMSs, such as fixed radiation monitors and alarms or personnel monitoring equipment.
(CFR: 41.11 / 41.12 / 43.4 / 45.9) 90 2.9 68 G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities, such as analysis and interpretation of radiation and activity readings as they pertain to administrative, normal, abnormal, and emergency procedures or to analysis and interpretation of coolant activity, including comparison to emergency plan or regulatory limits (SRO Only).
(CFR: 43.4 / 45.10) 91 3.8 98 Subtotal N/A 1
N/A 1
- 4.
Emergency Procedures /
Plan G2.4.42 Knowledge of emergency response facilities.
(CFR: 41.10 / 45.11) 92 2.6 69 G2.4.40 Knowledge of SRO responsibilities in emergency plan implementing procedures (SRO Only).
(CFR: 43.5 / 45.11) 93 4.5 99 G2.4.43 Knowledge of emergency communications systems and techniques.
(CFR: 41.10 / 43.5 / 45.13) 94 3.8 100 Subtotal N/A 1
N/A 2
Tier 3 Point Total N/A 6
N/A 7
Form 4.1-COMMON Common Examination Outline ES-4.1-COMMON COMMON Examination Outline (Seabrook)
Facility:
Seabrook Date of Exam:
03/06/2023 Theory (Tier 4) (RO)
Category K/A #
Topic RO Item #
IR Q#
Reactor Theory 192003 (K1.01)
REACTOR KINETICS AND NEUTRON SOURCES - Explain the concept of subcritical multiplication.
(CFR: 41.1) 95 2.8 70 192004 (K1.03)
REACTIVITY COEFFICIENTS - Describe the effect on the magnitude of the temperature coefficient of reactivity from changes in the following: Core Age.
(CFR: 41.1) 96 3.1 71 192006 (K1.06)
FISSION PRODUCT POISONS - Describe the following processes and state their effect on reactor operations: Transient Xenon.
(CFR: 41.1) 97 3.4 72 Subtotal N/A 3
Thermodynamics 193004 (K1.15)
THERMODYNAMIC PROCESS: Throttling and the Throttling Process - Determine the exit conditions for a throttling process based on the use of steam and/or water.
(CFR: 41.14) 98 2.8 73 193009 (K1.05)
CORE THERMAL LIMITS - State why thermal limits are necessary (CFR: 41.14) 99 3.5 74 193008 (K1.22)
THERMAL HYDRAULICS: Natural Circulation - Describe means to determine whether natural circulation flow exists.
(CFR: 41.14) 100 4.2 75 Subtotal N/A 3
Tier 4 Point Total N/A 6
Form 4.1-1 Record of Rejected Knowledge and Abilities Refer to Examination Standard (ES)-4.2, Developing Written Examinations, Section B.3, for deviations from the approved written examination outline.
Tier/Group Randomly Selected K/A Reason for Rejection Tier 1 /
Group 1 APE 008 AA2.01 RO Q2 APE 008; K/A AA2.04 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected APE and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected APE 008; K/A AA2.01, for Q2 to maintain K/A category balance within the outline.
Tier 1 /
Group 1 APE 008 AA2.12 RO Q2 APE 008; K/A AA2.01 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the re-selected APE and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected APE 008; K/A AA2.12, for Q2 to maintain K/A category balance within the outline.
Tier 1 /
Group 1 APE 015 AK2.07 RO Q4 APE 015; K/A AK2.14 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected APE and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected APE 015; K/A AK2.07, for Q4 to maintain K/A category balance within the outline.
Tier 1 /
Group 1 EPE 029 EK1.01 RO Q8 EPE 029; K/A EK1.03 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected EPE and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected EPE 029; K/A EK1.01, for Q8 to maintain K/A category balance within the outline Tier 1 /
Group 1 APE 056 AA2.33 RO Q11 APE 056; K/A AA2.54 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected APE and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected APE 056; K/A AA2.33, for Q11 to maintain K/A category balance within the outline.
Tier 2 /
Group 1 003 K4.07 RO Q28 SYSTEM 003; K/A 4.10 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected SYSTEM and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected SYSTEM 003; K/A K4.07, for Q28 to maintain K/A category balance within the outline.
Tier 2 /
Group 1 039 K4.05 RO Q43 SYSTEM 039 K/A K2.01 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected SYSTEM and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected SYSTEM 039; K/A K4.05, for Q43 to maintain K/A category balance within the outline.
Tier 2 /
Group 1 064 K1.07 RO Q50 SYSTEM 064; K/A K1.06 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected SYSTEM and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected SYSTEM 064; K/A K1.07, for Q50 to maintain K/A category balance within the outline.
Tier 2 /
Group 1 073 K1.06 RO Q51 SYSTEM 073; K/A K1.07 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected SYSTEM and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected SYSTEM 073; K/A 1.06, for Q51 to maintain K/A category balance within the outline.
Tier 2 /
Group 1 073 K1.04 RO Q51 SYSTEM 073; K/A K1.06 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the re-selected SYSTEM and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected SYSTEM 073; K/A 1.04, for Q51 to maintain K/A category balance within the outline.
Tier 2 /
Group 1 103 K5.01 RO Q54 SYSTEM 103; K/A K5.02 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected SYSTEM and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected SYSTEM 103; K/A 5.01, for Q54 to maintain K/A category balance within the outline.
Tier 2 /
Group 2 035 K6.11 RO Q61 SYSTEM 035; K/A K6.03 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected SYSTEM and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected SYSTEM 035; K/A 6.11, for Q61 to maintain K/A category balance within the outline.
Tier 4 Thermodynamic Theory 193010; K1.04 RO Q74 Thermodynamic Theory Topic 193007; K/A K1.06 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the originally selected THERMODYNAMIC THEORY TOPIC and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected THERMODYNAMIC THEORY TOPIC 193010; K/A K1.04, for Q74, to maintain K/A category balance within the outline.
Tier 4 Thermodynamic Theory 193009; K1.05 RO Q74 Thermodynamic Theory Topic 193010; K/A K1.04 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the re-selected THERMODYNAMIC THEORY TOPIC and K/A pairing without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected THERMODYNAMIC THEORY TOPIC 193009; K/A K1.05, for Q74, to maintain K/A category balance within the outline.