ML23247A001
| ML23247A001 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/21/2022 |
| From: | Joseph Demarshall NRC/RGN-I/DORS/OB |
| To: | Constellation Energy Generation |
| References | |
| EPID L-2022-OLL-0008 | |
| Download: ML23247A001 (1) | |
Text
Form 3.2-1 Administrative Topics Outline (Rev_071422)
Facility:
Ginna Date of Examination:
10/2022 Examination Level:
RO Operating Test Number:
N22-01 Administrative Topic (Step 1)
Activity and Associated KA (Step 2)
Type Code (Step 3)
Conduct of Operations 2.1.7 (4.4)
Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation JPM:
Perform 1/M Plot D, R Conduct of Operations 2.1.25 (3.9)
Ability to interpret reference materials, such as graphs, curves, and tables JPM:
Perform a Post-Trip Xenon Reactivity Determination M, R Equipment Control 2.2.12 (3.7)
Knowledge of Surveillance Procedures.
JPM:
Manually Calculate QPTR M, R Radiation Control 2.3.12 (3.2)
Knowledge of radiological safety principles and procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, or alignment of filters JPM:
Respond to a High Containment Area Radiation Monitor Alarm N, R
Form 3.2-1 Administrative Topics Outline (Rev_071422)
Instructions for completing Form 3.2-1, Administrative Topics Outline
- 1. For each license level, determine the number of administrative job performance measures (JPMs) and topic areas as follows:
Topic Number of JPMs
- Reactor operator (RO) applicants do not need to be evaluated on every topic (i.e.,
Equipment Control, Radiation Control, or Emergency Plan can be omitted by doubling up on Conduct of Operations),
unless the applicant is taking only the administrative topics part of the operating test with a waiver or excusal of the other portions).
RO*
SRO and RO Retakes Conduct of Operations 1 (or 2) 2 Equipment Control 1 (or 0) 1 Radiation Control 1 (or 0) 1 Emergency Plan 1 (or 0) 1 Total 4
5
- 2. Enter the associated knowledge and abilities (K/A) statement and summarize the administrative activities for each JPM.
- 3. For each JPM, specify the codes for location and source as follows:
Location:
(C)ontrol Room, (S)imulator, or Class(R)oom Source and Source Criteria:
(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams)
(D)irect from bank (no more than three for ROs, nor more than four for senior reactor operators (SROs) and RO retakes)
(N)ew or Significantly (M)odified from bank (no fewer than one)
RO Admin JPM Summary A1a This is a Bank JPM. The operator will be given the Estimated Critical Rod Position and told that the crew is performing a Reactor Startup per O-1.2, PLANT STARTUP FROM HOT SHUTDOWN TO FULL LOAD, that the Shutdown bank is already withdrawn, that the Control Bank B has been withdrawn to 220 Steps and that the raw data for each Control Bank Rod withdrawal has been entered on Attachment 1 of O-1.2.1, 1/M CURVES. The operator will be provided with this Attachment 1 and directed to complete the 1/M Plot on Attachment 1 of O-1.2.1 and (1) determine if the Reactor will go critical at a rod height greater than the 0% RIL and (2) determine if the Reactor will go critical within the +/-500 pcm bank positions. The operator will be expected to complete the 1/M Plot per
Form 3.2-1 Administrative Topics Outline (Rev_071422)
O-1.2.1 (per the provided KEY) and predict that the reactor will go critical above the 0%
RIL but outside the predicted +/-500 pcm bank positions (per the attached KEY).
A1b This is a modified Bank JPM. The applicant will be told that the plant has tripped from 75% power, current plant conditions and power history, that the XENON PREDICT program on the PPCS is NOT available, time of shutdown and current cycle burnup. The applicant will be provided with O-3, HOT SHUTDOWN WITH XENON PRESENT, and directed to perform a Xenon Reactivity determination in accordance with Steps 6.1.1 through 6.1.2 of O-3. The applicant will be expected to determine that it will take 18.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for Xenon to decay to the value at the time of shutdown, and that this will occur at 0200 on 10/18/22 in accordance with the KEY provided.
A2 This is a modified Bank JPM. The applicant will be told that Power Range Channel N-43 is INOPERABLE, that the QPTR Monitor Alarm has been declared NONFUNCTIONAL, that the plant was at 98.5% power when Control Rod G-9 dropped, that the crew entered AP-RCC.3, DROPPED ROD RECOVERY and that the plant is now at 91% power and stable. The applicant will be provided with the current Power Range Instrument readings (via photos) and the current Volts/µamp values and directed to calculate QPTR per Section 5 of O-6.4, QUADRANT POWER TILT CALCULATION. The applicant will be expected to determine QPTR in accordance with the KEY provided.
A3 This is a New JPM. The applicant will be told that the plant was shutdown per AP-RCS.1, REACTOR COOLANT LEAK, due to an RCS leak inside the Containment, that the plant has been in MODE 3 for 30 minutes, that R-29, CONTAINMENT HIGH RANGE AREA MONITOR, and R-2, CONTAINMENT AREA MONITOR have just alarmed and provided with current Containment Area Radiation Monitor readings. The applicant will be directed to use S-14.3, OPERATION OF CONTAINMENT HIGH RANGE AREA MONITORS R-29, R-30, to determine: (1) The Type and percent release to the Containment Atmosphere and (2) The Attachment that should be used to identify the isotopic concentrations in the Containment Atmosphere. The applicant will be expected to determine the type and percent of the release to the Containment Atmosphere to be a 1% Fuel release and that the isotopic concentrations in the Containment Atmosphere can be estimated using of S-14.3 per Section 6.1 of S-14.3 (per the attached KEY).
Form 3.2-1 Administrative Topics Outline (Rev_061522)
Facility:
Ginna Date of Examination:
10/2022 Examination Level:
SRO Operating Test Number:
N22-01 Administrative Topic (Step 1)
Activity and Associated KA (Step 2)
Type Code (Step 3)
Conduct of Operations 2.1.32 (4.0)
Ability to explain and apply system precautions, limitations, notes, or cautions JPM:
Determine Operating Limits for Station 13A Transmission D, P, R Conduct of Operations 2.1.20 (4.6)
Ability to interpret and execute procedure steps.
JPM:
Perform a Safety Function Determination D, R Equipment Control 2.2.41 (3.9)
Ability to obtain and interpret station electrical and mechanical drawings JPM:
Review a Clearance and Tagging Boundary for Work D, R Radiation Control 2.3.11 (4.3)
Ability to control radiation releases JPM:
Determine if a Radioactive Release is in Progress D, R Emergency Plan 2.4.41 (4.6)
Knowledge of the emergency action level thresholds and classifications (SRO Only)
JPM:
Classify an Emergency Event and Complete the Notification Form M, R
Form 3.2-1 Administrative Topics Outline (Rev_061522)
Instructions for completing Form 3.2-1, Administrative Topics Outline
- 1. For each license level, determine the number of administrative job performance measures (JPMs) and topic areas as follows:
Topic Number of JPMs
- Reactor operator (RO) applicants do not need to be evaluated on every topic (i.e.,
Equipment Control, Radiation Control, or Emergency Plan can be omitted by doubling up on Conduct of Operations),
unless the applicant is taking only the administrative topics part of the operating test with a waiver or excusal of the other portions).
RO*
SRO and RO Retakes Conduct of Operations 1 (or 2) 2 Equipment Control 1 (or 0) 1 Radiation Control 1 (or 0) 1 Emergency Plan 1 (or 0) 1 Total 4
5
- 2. Enter the associated knowledge and abilities (K/A) statement and summarize the administrative activities for each JPM.
- 3. For each JPM, specify the codes for location and source as follows:
Location:
(C)ontrol Room, (S)imulator, or Class(R)oom Source and Source Criteria:
(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams)
(D)irect from bank (no more than three for ROs, nor more than four for senior reactor operators (SROs) and RO retakes)
(N)ew or Significantly (M)odified from bank (no fewer than one)
SRO Admin JPM Summary A1a This is a Bank JPM. The applicant will be given a set of initial plant conditions including the current electrical grid conditions for Station 13A Bus Voltage and GSU Net MVARs.
The applicant will be provided with a blank copy of O-6.9, GINNA STATION OPERATING LIMITS FOR STATION 13A TRANSMISSION and directed to (1) Determine the OPERABILITY status of the Offsite Power Circuits, (2) Identify the O-6.9 Attachment used to make the OPERABILITY determination, and (3) Based on the current conditions, list required action(s) from O-6.9, if any, that the operating crew must take. The applicant will be expected to determine that for the given plant conditions the Offsite Power Circuits are INOPERABLE, in accordance with Attachment 2 of O-6.9 and identifies five required crew
Form 3.2-1 Administrative Topics Outline (Rev_061522) actions per the attached KEY. This JPM was previously used on the 2019 NRC Exam, randomly selected for use on the 2022 Exam.
A1b This is a Bank JPM. The applicant will be told that the plant is operating at 100% power, the status of two Important To Safety components that are out of service, that Containment Pressure Channel PI-946 has just failed that the crew has entered ER-INST.1, REACTOR PROTECTION BISTABLE DEFEAT AFTER INSTRUMENTATION LOOP FAILURE, and is preparing a Briefing for performance of Attachment 19, White Channel - CNMT CH 1B Pressure PI-946, and that they are completing A-52.4-F-01, CONTROL OF LIMITING CONDITIONS FOR OPERATING EQUIPMENT to assess the failure. The applicant will be directed to (1) identify the Tech Spec LCOs that have been entered because of the PI-946 failure and identify any required Tech Spec ACTION, (2) perform a Loss of Safety Function Determination PER A-52.4 Attachment 1 and (3) annotate whether a Loss of Safety Function has occurred and justify this decision. The applicant will be expected to determine that TS LCO 3.3.2 Function 2c and 4c are not met because of the failure of PI-946 requiring ACTION F.1 and J.1; perform a Safety Function Determination in accordance with Attachment 1 of A-52.4, and determine that a loss of Safety Function does NOT exist because at least one train of the affected functions is available to mitigate the accidents listed in the Applicable Safety Analysis bases section of the LCO(s) per the attached/provided KEYs (2).
A2 This is a Bank JPM. The applicant will be told that they are an Extra SRO in the Work Control Center and that piping failure has occurred on the Condensate Transfer Pump side of 4049C, CONDENSATE TRANSFER PUMP DISCHARGE CROSSTIE ISOLATION VALVE. The applicant will be provided with a previously prepared Tagging Request, OP-AA-109-101, PERSONNEL AND EQUIPMENT TAGOUT PROCESS and the necessary drawings and directed to review the Request for 4049C piping repair, recording comments for the Shift Manager. The applicant will be expected to determine that the tagging boundaries for 4049C are inadequate and identify all deficiencies in accordance with an attached KEY.
A3 This is a Bank JPM. The applicant will be told that the plant is operating at 100% power when a Tornado struck the Site resulting in the declaration of an Unusual Event. The initial report from the field indicates that a Tanker Truck carrying resin fines from the Retention Tank Cleanout has been overturned resulting in a radioactive spill that reaches the storm drains. The operator will be told that Initial responders including Radiation Protection personnel (RP) have responded to the scene and provided with preliminary radiological data. The applicant will be directed to use EP-AA-114-01, PWR RELEASE IN PROGRESS DETERIMNATION GUIDANCE, and determine whether a radioactive release is in progress, and IF SO, whether the release is liquid, airborne or both; and monitored or unmonitored. The applicant will be expected to identify that an unmonitored liquid release is in progress per the attached KEY.
A4 This is a Modified Bank JPM. The applicant will be told that the plant was at 30% power and shutting down due to a 5 gpm RCS leak when an earthquake was felt by all personnel in the control room and then provided a listing of key events occurring over the next several minutes. The applicant will be given a current plant data sheet and directed to (1) make the Initial classification of the event in accordance with EP-AA-111, EMERGENCY CLASSIFICATION AND PROTECTIVE ACTION RECOMMENDATIONS, to provide the classification to the examiner; and then prepare a GNP NY STATE RADIOLOGICAL
Form 3.2-1 Administrative Topics Outline (Rev_061522)
EMERGENCY DATA FORM (PART 1) for the event, and present it to the Emergency Director (i.e., Examiner) for approval. The applicant will be expected to declare a Site Area Emergency based on EAL FS1 within 15 minutes and complete the GNP NY STATE RADIOLOGICAL EMERGENCY DATA FORM (PART 1), within 15 minutes of the time of event classification. The JPM will have two Time Critical time periods, the first requiring the applicant to classify the event within 15 minutes and the second requiring the applicant to complete the form within 15 minutes of event classification.
Form 3.2-2 Control Room/In-Plant Systems Outline (Rev_062722)
Facility:
Ginna Date of Examination:
10/2022 Exam Level:
RO / SRO-I / SRO-U Operating Test No.:
N22-01 System / JPM Title Type Code Safety Function Control Room Systems A.
056 Condensate System [056 A4.01 (3.3)]
Swapping Condensate Booster Pumps S, N, A 4S B.
064 Emergency Diesel Generators [064 A4.06 (3.9)]
Start the Emergency Diesel Generator during a Station Blackout S, M, A, L, EN 6
C.
010 Pressurizer Pressure Control System [010 A4.03 (3.8)]
Placing LTOP In Service S, M, A, L 3
D.
004 Chemical and Volume Control System [004 A4.08 (4.0)]
CVCS Leak Isolation S, D, A 2
E.
015 Nuclear Instrumentation System [APE 032 AA1.01 (2.9)]
Manually Energize the Source Range Instruments S, N, A, L 7
F.
005 Residual Heat Removal System [APE 025 AA1.11 (3.1)]
Lineup RCDT Pump for Core Cooling S, D, L 4P G.
026 Containment Spray System [026 A4.01 (3.9)]
Evaluate CNMT Spray Flow Requirements and Reduce Flow in ECA-1.1 S, D, EN, L 5
H.
001 Control Rod Drive System [APE 003 AA1.08 (3.7)] (RO only)
Recover a Dropped Control Rod S, D 1
In-Plant Systems I.
086 Fire Protection System [086 A2.05 (4.0/3.8)]
Take Local Manual Control of Charging Pump D, R, E 8
J.
001 Control Rod Drive System [EPE 029 EA1.20 (4.0)]
Locally Open the Control Rod Drive MG Set Breakers D, E 1
K.
039 Main and Reheat Steam System [EPE 038 EA1.16 (3.6)]
Perform ATT-16.0, Ruptured S/G M, E 4S
Form 3.2-2 Control Room/In-Plant Systems Outline (Rev_062722)
Form 3.2-2 Instructions for Control Room/In-Plant Systems Outline
- 1. Determine the number of control room system and in-plant systems job performance measures (JPMs) to develop using the following table:
License Level Control Room In-Plant Total Reactor Operator (RO) 8 3
11 Senior Reactor Operator
- Instant (SRO-I) 7 3
10 Senior Reactor Operator
- Upgrade (SRO-U) 2 or 3 3 or 2 5
- 2. Select safety functions and systems for each JPM as follows:
Refer to Section 1.9 of the applicable knowledge and abilities (K/A) catalog for the plant systems organized by safety function. For pressurized-water reactor operating tests, the primary and secondary systems listed under Safety Function 4, Heat Removal from Reactor Core, in Section 1.9 of the applicable K/A catalog, may be treated as separate safety functions (i.e., two systems, one primary and one secondary, may be selected from Safety Function 4). From the safety function groupings identified in the K/A catalog, select the appropriate number of plant systems by safety functions to be evaluated based on the applicants license level (see the table in step 1).
For RO/SRO-I applicants: Each of the control room system JPMs and, separately, each of the in-plant system JPMs must evaluate a different safety function, and the same system or evolution cannot be used to evaluate more than one safety function in each location. One of the control room system JPMs must be an engineered safety feature.
For SRO-U applicants: Evaluate SRO-U applicants on five different safety functions.
One of the control room system JPMs must be an engineered safety feature, and the same system or evolution cannot be used to evaluate more than one safety function.
- 3. Select a task for each JPM that supports, either directly or indirectly and in a meaningful way, the successful fulfillment of the associated safety function. Select the task from the applicable K/A catalog (K/As for plant systems or emergency and abnormal plant evolutions) or the facility licensees site-specific task list. If this task has an associated K/A, the K/A should have an importance rating of at least 2.5 in the RO column. K/As that have importance ratings of less than 2.5 may be used if justified based on plant priorities; inform the NRC chief examiner if selecting K/As with an importance rating less than 2.5.
The selected tasks must be different from the events and evolutions conducted during the simulator operating test and tasks tested on the written examination. A task that is similar to a simulator scenario event may be acceptable if the actions required to complete the task are significantly different from those required in response to the scenario event.
Apply the following specific task selection criteria:
At least one of the tasks shall be related to a shutdown or low-power condition.
Form 3.2-2 Control Room/In-Plant Systems Outline (Rev_062722)
Four to six of the tasks for RO and SRO-I applicants shall require execution of alternative paths within the facility licensees operating procedures. Two to three of the tasks for SRO-U applicants shall require the execution of alternative paths within the facility licensees operating procedures.
At least one alternate path JPM must be new or modified from the bank.
At least one of the tasks conducted in the plant shall evaluate the applicants ability to implement actions required during an emergency or abnormal condition.
At least one of the tasks conducted in the plant shall require the applicant to enter the radiologically controlled area. This provides an excellent opportunity for the applicant to discuss or demonstrate radiation control administrative subjects.
If it is not possible to develop or locate a suitable task for a selected system, return to step 2 and select a different system.
- 4. For each JPM, specify the codes for type, source, and location:
Code License Level Criteria RO SRO-I SRO-U (A)lternate path 4-6 (5) 4-6 (5) 2-3 (3)
(C)ontrol room (D)irect from bank 9 (6) 8 (5) 4 (2)
(E)mergency or abnormal in-plant 1 (3) 1 (3) 1 (2)
(EN)gineered Safety Feature (For control room system) 1 (2) 1 (2) 1 (1)
(L)ow-Power / Shutdown 1 (5) 1 (5) 1 (2)
(N)ew or (M)odified from bank (must apply to at least one alternate path JPM) 2 (5) 2 (5) 1 (3)
(P)revious 2 exams (randomly selected) 3 (0) 3 (0) 2 (0)
(R)adiologically Controlled Area 1 (1) 1 (1) 1 (1)
(S)imulator JPM Summary JPM A This is a New JPM. The applicant will be told that the plant is operating at power and that the operating crew is required to swap Condensate Booster Pumps. The applicant will be directed to start Condensate Booster Pump B and stop Condensate Booster Pump C in accordance with P-17.22, SWAPPING CONDENSATE BOOSTER PUMPS. The applicant will be expected to manually adjust the Trim Valve positions, start Condensate Booster Pump B, stop Condensate Booster Pump C and place the Trim Valves back in automatic control per P-17.22 and then respond to high bearing temperature alarms on Condensate Pump A by stopping the pump and verifying that Condensate Pump B automatically starts in accordance with AR-PPCS-T2108, AR-PPCS-T2119 and/or AR-PPCS-T2120 (Alternate Path).
JPM B This is a Modified Bank JPM. The applicant will be told that the plant was at 100% power when it experienced a loss of all AC power, that the crew is at Step 8 of ECA-0.0, LOSS OF ALL AC POWER, and that the Secondary Operator reports that B D/G has been RESET locally and is ready to start. The applicant will be directed to perform Step 7.b
Form 3.2-2 Control Room/In-Plant Systems Outline (Rev_062722)
RNO of ECA-0.0 to start the B D/G. The applicant will be expected to manually start the B D/G and then manually stop the B D/G when it is determined that cooling water cannot be established per Step 7 of ECA-0.0. (Alternate Path).
JPM C This is a Modified Bank JPM. The applicant will be told that a unit shutdown to Cold Shutdown is in progress, that the operating crew is performing O-2.2, PLANT SHUTDOWN FROM HOT STANDBY TO COLD CONDITIONS, and are ready to place LTOP in service, and that O-7, ALIGNMENT AND OPERATION OF THE REACTOR VESSEL OVERPRESSURE PROTECTION SYSTEM, has been started and is complete through Section 6.1, Initial Conditions. The applicant will be directed to place PCV-430 in the LTOP Mode in accordance with O-7, Section 6.2 and then place PCV-431C in the LTOP Mode in accordance with O-7, Section 6.3. The applicant will be expected to place PCV-430 in the LTOP mode of operation per Section 6.2 of O-7, place PCV-431C in the LTOP mode of operation per Section 6.3 of O-7, and respond to MCB Annunciator AA-23, RCS OVER-PRESS PROTECTION TRAIN B HI PRESS, when PT-451 fails HIGH (Alternate Path), ultimately defeating Wide Range RCS Pressure Channel PT-451 per AR-AA-23, RCS OVER-PRESS PROTECTION TRAIN B HI PRESS.
JPM D This is a Bank JPM. The applicant will be told that the plant is at 100% power, that the crew has entered AP-CVCS.1, CVCS LEAK, completing steps 1-5, that Pressurizer level is stable, and that the leak is believed to be in Containment. The applicant will be directed to take appropriate actions to isolate the leak starting with Step 6 of AP-CVCS.1. During the course of the action, the applicant will recognize that both operating Charging Pumps have tripped (Alternate Path) resulting in a complete loss of Charging. The applicant will be expected to isolate Normal Letdown and Charging, recognize that a complete loss of Charging flow has occurred, start Charging Pump A, and isolate the normal Charging Header from the RCS in accordance with the Annunciator Response Procedures, AP-CVCS.1 and/or AP-CVCS.3. There are four potential success paths scripted in the JPM:
(1) Use AR-G-25 to start Charging Pump A and return to AP-CVCS.1, (2) Use AR-G-25 to transition to AP-CVCS.3, use this procedure to start Charging Pump A and return to AP-CVCS.1, (3) Use AR-B-9 and/or 10 to transition to AP-CVCS.3, use this procedure to start Charging Pump A and return to AP-CVCS.1 and (4) enter AP-CVCS.3 directly, start Charging Pump A and return to AP-CVCS.1.
JPM E This is a New JPM. The applicant will be told that the plant was manually tripped 25 minutes ago from 100% power and the crew entered E-0, REACTOR TRIP OR SAFETY INJECTION, and that the crew has completed Step 12 of ES-0.1, REACTOR TRIP RESPONSE. The applicant will be directed to perform Step 13 of ES-0.1, Check if Source Range Detectors Should be Energized. The applicant will be expected to attempt to manually energize the Source Range instruments in accordance with ES-0.1, recognize that this was unsuccessful (Alternate Path) and manually energize the Source Range instruments by pulling the instrument fuses in accordance with ER-NIS.1, SR MALFUNCTION.
JPM F This is a Bank JPM. The applicant will be told that the plant was performing a normal cooldown when both RHR Pumps tripped, that the S/Gs are unavailable, that AP-RHR.1, LOSS OF RHR, was entered and attempts to restart the pumps were unsuccessful, that all attempts to start an SI Pump have failed, that the crew is in progress of establishing Containment Integrity and all personnel are clear of Containment. The applicant will be directed to lineup and start the RCDT pumps to provide core cooling per Section 6.2 of
Form 3.2-2 Control Room/In-Plant Systems Outline (Rev_062722)
ER-RHR.1, RCDT PUMP OPERATION FOR CORE COOLING. The applicant will be expected to line up the RCDT pumps to provide core cooling per Section 6.2 of ER-RHR.1.
JPM G This is a Bank JPM. The applicant will be told that the plant has experienced a SBLOCA, that the crew entered E-0, REACTOR TRIP OR SAFETY INJECTION; and then transitioned to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, that the SBLOCA degraded to a LBLOCA and then the crew transitioned into ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION. The applicant will be directed to perform Step 4 of ECA-1.1. The applicant will be expected to reset the CNMT Spray Signal, stop both Containment Spray Pumps and close the Containment Spray Pump Discharge Valves per Step 4 of ECA-1.1.
JPM H This is a Bank JPM. The applicant will be told that 30 minutes ago the plant was at 50%
power when Control Rod E-7 dropped, that the cause of the dropped rod has been determined and corrected, that the crew is performing AP-RCC.3, DROPPED ROD RECOVERY and ER-RCC.1, RETRIEVAL OF A DROPPED RCC, and that the Reactor Engineer has prepared a Reactivity Plan to complete the Control Rod withdrawal. The applicant will be directed to recover the rod by performing ER-RCC.1, steps 6.2.4 through 6.2.6 and told that the CO will be available to operate the Boric Acid controls and IMD is standing by in the Relay Room to operate the P/A converter as directed. The applicant will be expected to coordinate with and direct the IMD Technicians and CO in order to withdraw Control Rod E-7 to the first MRPI transition (8 Steps) per ER-RCC.1.
JPM I This is a Bank JPM. The applicant will be told that a fire in the Cable Tunnel is on-going, forcing the crew to implement ER-FIRE.1, ALTERNATE SHUTDOWN FOR CONTROL ROOM ABANDONMENT, and that the Shift Manager has directed you to perform ER-FIRE.1, Attachment 4, which is complete through Step 14.0. The applicant will be directed to continue with Step 15.0 of Attachment 4 of ER-FIRE.1. The applicant will be expected to take local control of and start Charging Pump A in accordance with Attachment 4 of ER-FIRE.1.
JPM J This is a Bank JPM. The applicant will be told that the plant has experienced a reactor trip signal and the crew entered procedure E-0, REACTOR TRIP OR SAFETY INJECTION; and that the reactor trip could not be verified, and the crew entered FR-S.1, RESPONSE TO REACTOR RESTART/ATWS. The applicant will be directed to locally trip the reactor per FR-S.1. The applicant will be expected to locally trip the Control Rod Drive Motor Generator Set Breakers by depressing the TRIP pushbutton for 52-1/MG1A and 52-1/MG1B per Step 1 RNO of FR-S.1.
JPM K This is a Modified Bank JPM. The applicant will be told that the plant was operating at 100% power when a Steam Generator Tube Rupture occurred in B Steam Generator, that the crew is currently at Step 5 of E-3, STEAM GENERATOR TUBE RUPTURE, and that B All Volatile Treatment (AVT) mixed bed is in service. The applicant will be directed to complete ATT-16.0, ATTACHMENT RUPTURED S/G, for B S/G. The applicant will be expected to locally isolate B S/G by closing the B MSIV locally and the TDAFW Pump Steam Root Valve from the B S/G in accordance with ATT-16.0, ATTACHMENT RUPTURED S/G.
Form 3.3-1 Scenario Outline (Rev_083122)
Facility:
Ginna Scenario No.:
1 Op Test No.:
N22-1 Scenario Source: Modified from N19-1-3 Examiners:
Operators:
(SRO)
(RO)
(BOP)
Initial Conditions:
The plant is at 100% power (MOL). It is expected to maintain power stable at the current power level throughout the shift.
Turnover:
The following equipment is Out-Of-Service: Control Rod Shroud Fan A is OOS for breaker corrective maintenance, and the Condensate Booster Pump A is OOS for thrust bearing replacement. It is expected to realign the electric plant from 0/100 to 50/50 normal lineup. See additional Turnover information on last page of Form 3.3-
- 2.
Critical Tasks:
See Below Event No.
Malf.
No.
Event Type*
Event Description 1
NA N-BOP N-SRO Shift Electric Plant 2
MAL PRZ04 I(MC)-RO I(TS)-SRO Master Pressure Controller (431K) fails HIGH 3
MAL RCS02D C(MC)-RO C(TS)-SRO 25 GPM RCS Loop A Cold Leg Leak 4
NA R-RO C-BOP C-SRO Rapid Downpower 5
MAL TUR09D OVR TUR05O/P C(MC)-BOP C-SRO Failure of Turbine Control/EHC 6
MAL RCS02D M-ALL 1000 gpm Loop A Cold Leg SBLOCA (CT-2) 7 MAL RPS07A SIS03C C(MC)-RO C-SRO Failure of A SI Pump to Auto Start, Trip of C SI Pump (CT-1)
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec, (MC) Manual Control
Form 3.3-1 Scenario Outline (Rev_083122)
Ginna 2022 NRC Scenario #1 The plant is at 100% power (MOL). It is expected to maintain power stable at the current power level throughout the shift.
The following equipment is Out-Of-Service: Control Rod Shroud Fan A is OOS for breaker corrective maintenance, and the Condensate Booster Pump A is OOS for thrust bearing replacement. It is expected to realign the electric plant from 0/100 to 50/50 normal lineup.
Shortly after taking the watch, the Operator will shift the Electric Line-up from 0/100 to 50/50 normal in accordance with O-6.9.2, ESTABLISHING AND/OR TRANSFERRING OFFSITE POWER TO BUS 12A/12B.
Following this, the Master Pressure Controller (431K) will fail such that the output in AUTO goes to 100%, causing both Pressurizer Spray Valves to OPEN, and PRZR pressure to LOWER. The operator will respond by placing 431K in MANUAL to stabilize and restore PRZR pressure to the normal band using the guidance of A-503.1, EMERGENCY AND ABNORMAL OPERATING PROCEDURES USERS GUIDE, AR-F-10, PRESSURIZER LO PRESS 2205 PSI, and AP-PRZR.1, ABNORMAL PRESSURIZER PRESSURE. As an alternative the operator may place both the Pressurizer Spray Valve controllers (HCV-431A and 431B) in MANUAL to stabilize and restore PRZR pressure to the normal band. The operator will address Technical Specification LCO 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS, and Technical Requirements Manual TR-3.4.3, ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) MITIGATION. Using Time Compression, the Master Pressure Controller (431K) will be repaired, returned to AUTO and the scenario will continue.
Subsequently, an RCS leak in the Loop A Cold Leg will occur. The operator will respond by controlling charging pump speed and flow in MANUAL to stabilize and maintain PRZR level at program per AR-F-14, CHARGING PUMP SPEED and AP-RCS.1, REACTOR COOLANT LEAK.
Additionally, the operator will determine the RCS leak rate using HC-2.0, RCS LEAKRATE DETERMINATION. The operator will address Technical Specification LCO 3.4.13, RCS OPERATIONAL LEAKAGE.
At Step 17 of AP-RCS.1 the crew will initiate a downpower per AP-TURB.5, RAPID LOAD REDUCTION and S-3.1, BORON CONCENTRATION CONTROL. In preparation for bringing the unit off-line the crew will perform ATT-23.0, ATTACHMENT TRANSFER 4160V LOADS.
During the load reduction, the Main Turbine will fail in automatic control. The operator will diagnose the failure and use manual control of the turbine to conduct the downpower per AP-TURB.5, RAPID LOAD REDUCTION using the guidance of A-503.1, EMERGENCY AND ABNORMAL OPERATING PROCEDURES USERS GUIDE.
After this, the RCS leak in the Loop A Cold Leg will abruptly degrade to a 1000 gpm LOCA, the reactor/turbine will trip and Safety Injection will actuate. On SI actuation, the A SI Pump will fail to start automatically and the C SI Pump will trip. The operator will be required to manually start the A SI Pump. RCP trip criteria will be met in this event. The operator will enter E-0, REACTOR TRIP OR SAFETY INJECTION, and transition to E-1, LOSS OF REACTOR OR SECONDARY COOLANT.
The crew will transition to ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION and the scenario will terminate at Step 8 of ES-1.2, after the crew has initiated an RCS Cooldown.
Form 3.3-1 Scenario Outline (Rev_083122)
Critical Tasks:
CT-1: Establish flow from at least two SI pumps before transition to E-1, Loss of Reactor or Secondary Coolant Initiating Cue:
SI Annunciator is LIT (Step 4 of E-0), SI Pump A and C breaker status lights indicate that these pumps are NOT running (Step 1.a of ATT-27.0).
Performance Feedback:
SI Pump A breaker status lights and total SI flow raising indicate that SI Pump is running (Indication is available to the Control Operator (HCO or CO) assigned to perform ATT-27.0. (Step 1.b of ATT-27.0). If the action is not taken RVLIS will indicate lower level in the core and Core Exit Thermocouples will rise over the course of the scenario.
Success Path:
SI Pump A is started such that two SI Pumps (A & B) are running BEFORE the transition out of E-0 which is expected to occur at Step 17 of E-0.
Measurable Performance Standard:
Expected Actions: SI Pump A is started (SI Pump B started automatically).
Safety Significance: Failure to manually start at least two SI pumps under the postulated conditions constitutes "mis-operation or incorrect crew performance which leads to degraded ECCS capacity." In this case, at least two SI pumps can be manually started from the control room. Therefore, failure to manually start SI pumps also represents a "demonstrated inability by the crew to (1) Recognize a failure/incorrect auto actuation of an ESF system or component and (2) Effectively direct/manipulate ESF controls. The acceptable results obtained in the FSAR analysis of a small-break LOCA are predicated on the assumption of minimum ECCS pumped injection. The analysis assumes that a minimum pumped ECCS flow rate, which varies with RCS pressure, is injected into the core. The flow-rate values assumed for minimum pumped injection are based on operation of the following ECCS pumps: Two SI pumps and one RHR pump.
Operation of this minimum required complement of ECCS injection pumps is consistent with the FSAR assumption that only minimum safeguards are actuated. Because compliance with the assumption of the FSAR is part of the facility license condition, failure to perform the critical task (under the postulated plant conditions) constitutes a violation of the license condition.
CT-2: Trip all RCPs within 5 minutes of reaching the trip criteria Initiating Cue:
Indications of a Small Break LOCA (PRZR pressure lowers to less than 1750 psig, PRZR level lowers to 0%, CNMT pressure rises to greater than 4 psig); SI Annunciator is LIT (Step 4 of E-0); RCP Trip Criteria is met: BOTH (1) Two SI Pumps running (A & B) and (2) RCS pressure minus maximum S/G pressure - LESS THAN 210 psi [240 psi adverse CNMT] (E-0 Foldout Page item 1).
Performance Feedback:
RCP A and B breaker status lights indicate that both RCPs are OFF and RCS Loop flow lowers (Indication is available to HCO).
Form 3.3-1 Scenario Outline (Rev_083122)
Success Path:
The RCPs are tripped within five minutes of the E-0 Foldout Page Item #1 criteria being met.
Measurable Performance Standard:
Expected Actions: Trip RCPs A and B within 5 minutes of meeting the E-0/E-1 Foldout Page Item #1 criteria.
Safety Significance: Failure to trip the RCPs under the postulated plant conditions leads to core uncovery and to fuel cladding temperatures in excess of 2200°F, which is the limit specified in the ECCS acceptance criteria. Thus, failure to perform the task represents "mis-operation or incorrect crew performance which leads to degradation of the fuel cladding barrier to fission product release" and to "violation of the facility license condition." The analysis presented in the FSAR for a small-break LOCA assumes that the RCPs trip because of a loss of offsite power that coincides with the reactor trip. However, during a small-break LOCA, offsite power might remain available and RCPs might continue to run for some period of time creating a window of time in which if the RCPs were to trip the core would uncover and fuel cladding temperatures in excess of 2200°F would be reached. According to Appendix B.16 of WCAP-17711-NP, Pressurized Water Reactor Owners Group Westinghouse Emergency Response Guideline Revision 2-Based Critical Tasks, if the RCPs are tripped within 5 minutes of the trip criteria being met, PCT (i.e.,
fuel cladding temperatures) remains below 2200°F.
"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 (Rev_083122)
Facility:
Ginna Scenario No.:
2 Op Test No.:
N22-1 Scenario Source: Modified from N17-1-1 Examiners:
Operators:
(SRO)
(RO)
(BOP)
Initial Conditions:
The plant is at 100% power (EOL). Per the daily work schedule, P-17.7, SWAPPING SERVICE WATER PUMPS, is to be performed this shift, swapping Service Water Pumps C & D so that maintenance can be performed on SWP C.
Turnover:
The following equipment is Out-Of-Service: SAFW Pump C (TS 3.7.5 Condition E Action E.1) is OOS for breaker maintenance, and the Condensate Booster Pump A is OOS for thrust bearing replacement.
Critical Tasks:
See Below Event No.
Malf.
No.
Event Type*
Event Description 1
NA N-BOP N-SRO Swap Service Water Pumps (SWP) 2 MAL CLG09C CLG01D C-BOP C(TS)-SRO SWP C Check Valve sticks OPEN/SWP D Trips 3
MAL PZR03B I(MC)-RO I-BOP I(TS)-SRO Pressurizer Level LT-427 fails LOW 4
MAL CVC07A C(MC)-RO C-SRO Letdown Pressure Control Valve fails CLOSED in AUTO 5
MAL TUR11B R-RO C-BOP C-SRO Turbine Control Valve CV-L4 Drifts Closed/Downpower 6
MAL STM03 STM05A STM05B FPS01-Z37 FPS01-Z38 M-ALL Steamline Break in Intermediate Building/Delayed closure of MSIVs 7
MAL RPS05A RPS05B C(MC)-RO C-SRO Automatic Rx Trip fails (Manual Pushbutton Available) (CT-1) 8 MAL FDW12 FDW11A FDW11B C-RO C-SRO MDAFW and TDAFW Pumps fail to START (CT-2)
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec, (MC) Manual Control
Form 3.3-1 Scenario Outline (Rev_083122)
Ginna 2022 NRC Scenario #2 The plant is at 100% power (EOL). Per the daily work schedule, P-17.7, SWAPPING SERVICE WATER PUMPS, is to be performed this shift, swapping Service Water Pumps C & D so that maintenance can be performed on SWP C.
The following equipment is Out-Of-Service: SAFW Pump C (TS 3.7.5 Condition E Action E.1) is OOS for breaker maintenance, and the Condensate Booster Pump A is OOS for thrust bearing replacement.
Shortly after taking the watch, the operator will swap the Service Water Pumps (SWP) C & D in accordance with P-17.7, SWAPPING SERVICE WATER PUMPS. The operator will start SWP D and then stop SWP C.
When the operator stops SWP C, its Discharge Check Valve will stick partially open, resulting in lower system pressure. The operator may restart SWP C based on these indications, or enter AP-SW.1, SERVICE WATER LEAK, and restart SWP C. When SWP C is re-started SWP D will trip.
The operator will respond in accordance with AR-J-9, SAFEGUARD BREAKER TRIP, and AP-SW.2, LOSS OF SERVICE WATER. The operator will address Technical Specification LCO 3.7.8, SERVICE WATER (SW) SYSTEM.
Following this, Pressurizer Level Channel 427 will fail LOW, resulting in letdown isolation and de-energizing the pressurizer heaters. The crew will respond per AR-F-11, PZR LOW LEVEL 13%, and ER-INST.1, REACTOR PROTECTION BISTABLE DEFEAT AFTER INSTRUMENTATION LOOP FAILURE. The crew will use HC-1.0, LOSS OF LETDOWN, to control RCS inventory when Normal Letdown isolates. The crew will defeat the failed channel, reset PZR heaters, reduce charging to a single charging pump with speed control in Manual, and re-establish letdown per S-3.2E, PLACING IN OR REMOVING FROM SERVICE NORMAL LETDOWN/EXCESS LETDOWN. Then, the crew will slowly restore PZR level to program (56%) and return charging pump with speed control to Auto.
The operator will address Technical Specification LCO 3.3.1, REACTOR TRIP SYSTEM (RTS)
INSTRUMENTATION and LCO 3.4.9, PRESSURIZER.
Subsequently, the Letdown Pressure Control Valve PCV-135 will fail CLOSED in automatic control causing a loss of letdown flow. The operator will take manual control of PCV-135 and restore Letdown flow/pressure per AR-A-11, LETDOWN LINE HI PRESS 400 PSI.
Then, Turbine Control Valve CV L-4 will drift closed. The crew will respond per AP-TURB.2, TURBINE LOAD REJECTION, and begin a load reduction to approximately 82% power using AP-TURB.5, RAPID LOAD REDUCTION and S-3.1, BORON CONCENTRATION CONTROL.
After this, a large steam break occurs downstream of the MSIVs in the Intermediate Building, and both MSIVs will fail to automatically or manually close. The reactor will fail to trip automatically. The operator will be able to manually trip the Reactor using the manual pushbutton(s). Additionally, all AFW pumps fail to start due to the high-energy break. The crew will implement E-0, REACTOR TRIP OR SAFETY INJECTION.
Both MSIVs will automatically close after 2.5 minutes. The crew will transition to FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, at Step 9 of E-0; and will be required to initiate RCS Bleed and Feed. Upon successful implementation of RCS Bleed and Feed, the crew will start SAFW Pump D restore a feed source to at least one S/G in accordance with ATT-5.1, ATTACHMENT SAFW FROM SW and ATT-22.0, ATTACHMENT RESTORING FEED FLOW.
Form 3.3-1 Scenario Outline (Rev_083122)
The scenario will terminate at Step 29.b of FR-H.1, after feed flow has been restored from SAFW Pump D.
Critical Tasks:
CT-1: Manually trip the reactor from the control room within 60 seconds of the first alarm indicating that an Automatic Reactor Trip is required Initiating Cue:
Indication and/or annunciation that plant parameter(s) exist that should result in automatic reactor trip but reactor does not automatically trip.
Performance Feedback:
After depressing one of two manual Rx Trip Pushbuttons the Breaker Status lights of the Rx Trip breakers indicate the breakers are OPEN (Green Status lights LIT, Red status lights OFF) and MRPI indicates that the Control Rods are on the bottom of the core.
Success Path:
Depressing one of two manual Rx Trip Pushbuttons.
Measurable Performance Standard:
Expected Actions: Depressing one of two manual Rx Trip Pushbuttons within 60 seconds of indication and/or annunciation that plant parameter(s) exist that should result in automatic reactor trip but reactor does not automatically trip.
Safety Significance: Failure to manually trip the reactor causes a challenge to the subcriticality CSF beyond that irreparably introduced by the postulated conditions. Additionally, it constitutes an incorrect performance that necessitates the crew taking action which complicates the event mitigation strategy that demonstrates the inability by the crew to recognize a failure of the automatic actuation of the RPS. According to GINNA/UFSAR, Section 15.8.3.2, Operator Actions Assumed, The following operator actions ensure successful mitigation of the broad range of ATWS events: Manual scram within the first 60 seconds of the event if the RTS fails to automatically scram. This action is a backup to the RTSs failure to generate an automatic trip signal. This requirement is identified on Attachment 1 of OP-GI-102-106, OPERATOR RESPONSE TIME PROGRAM AT GINNA STATION, Type Action # TCA-1101A which requires the operator to manual scram if auto scram fails within 60 seconds.
CT-2: Once FR-H.1 RCS Bleed and Feed criteria have been met, Stop the Reactor Coolant Pumps and Open the Pressurizer PORVs BEFORE the Pressurizer Code Safety Valves open (2485 psig).
Initiating Cue:
The following five conditions exist: (1) Extreme (RED path) challenge to the heat sink CSF, (2) Indication that RCS pressure is greater than the pressure in any non-faulted SG (3) Indication that RCS temperature is greater than the highest temperature at which the RHR system can be placed in service in the shutdown cooling mode (350F), (4) Indication and/or annunciation that no AFW is available, and (5) Both S/G Wide Range levels indicate less than 120 inches.
Performance Feedback:
RCP breaker status lights indicate the RCPs are OFF (Green lights LIT, Red lights OFF); After opening both PRZR PORVs the PORV
Form 3.3-1 Scenario Outline (Rev_083122) position indicating status lights indicate the valves are OPEN (Red lights LIT, Green lights OFF) and PRT level, pressure and temperature are rising.
Success Path:
Crew determines at Step 2 of FR-H.1 that RCS Bleed and Feed Criteria are met, stops BOTH RCPs per Step 2.b and opens both PRZR PORVs per Step 17.
Measurable Performance Standard:
Expected Actions: Trips both RCPs, Opens both PRZR PORVS using ATT-12.0, BEFORE the Pressurizer Code Safety Valves open.
Safety Significance: Failure to initiate RCS bleed and feed before the RCS saturates at a pressure above the shutoff head of the high-head ECCS pumps results in significant and sustained core uncovery. If RCS bleed is initiated so that the RCS is depressurized below the shutoff head of the high-head ECCS pumps, then core uncovery is prevented or minimized. If at Step 2 of FR-H.1 the crew does not recognize that RCS Bleed and Feed is required, they will not stop the RCPs, or open the Pressurizer PORVs, as required; and continue attempts to restore a Secondary Heat Sink. The crew will be unable to restore a Secondary Heat Sink at this time.
This will result in RCS temperature and pressure increase. The Pressurizer PORVs will not automatically open (No available Instrument Air) and eventually the Pressurizer Code Safety Valves will open resulting in a loss of RCS inventory without the ability to add inventory with the High-Head SI Pumps. Tests show that this condition will result 18 minutes after the time that the RCS Bleed and Feed criteria are met. If the crew fails to stop the Reactor Coolant Pumps and Open the Pressurizer PORVs BEFORE the Pressurizer Code Safety Valves open once RCS Bleed and Feed criteria are met; when opportunity exists to do so; it 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.
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 #2 confirms that the scenario is built to directly transition from E-0 to Functional Recovery Procedure FR-H.1, Response to Loss of Secondary Heat Sink, 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 #2 is a complex scenario that exercises Contingency EOP Procedure FR-H.1 for the Loss of Secondary Heat Sink, (b) FR-H.1 requires the use of alternate decision paths and prioritization of actions within the EOP to mitigate the CSFST Heat Sink Red Path condition, and (c) FR-H.1 has measurable actions that must be taken by the crew.
Form 3.3-1 Scenario Outline (Rev_083122)
Facility:
Ginna Scenario No.:
3 Op Test No.:
N22-1 Scenario Source: Modified from N17-1-2 Examiners:
Operators:
(SRO)
(RO)
(BOP)
Initial Conditions:
The plant is at 77.5% power (MOL). The crew is expected to remain at this power level while maintenance is completed during the shift.
Turnover:
The following equipment is Out-Of-Service: SAFW Pump C (TS 3.7.5 Condition E Action E.1) is OOS for breaker maintenance, and the Condensate Booster Pump A is OOS for thrust bearing replacement.
Critical Tasks:
See Below Event No.
Malf.
No.
Event Type*
Event Description 1
MAL CLG05 CLG02B C-RO C(TS)-SRO Leak on the CCW System/CCW Pump B Trips 2
MAL FDW22A C(MC)-BOP C-SRO FRV A Oscillations 3
MAL CND01C R-RO C-BOP C-SRO Condensate Booster Pump C trips/Downpower 4
MAL ROD02-F2 C-RO C-BOP C(TS)-SRO Dropped Rod 5
MAL ROD02-B8 C-RO C-SRO Subsequent 2nd Dropped Rod 6
MAL SGN04A M-ALL Steam Generator Tube Rupture in S/G A (CT1, 2 and 3) 7 MAL PZR01D PZR05A PZR05B C-RO C-SRO RCS Depressurization Valve Fails OPEN after depressurization (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec, (MC) Manual Control
Form 3.3-1 Scenario Outline (Rev_083122)
Ginna 2022 NRC Scenario #3 The plant is at 77.5% power (MOL). The crew is expected to remain at this power level while maintenance is completed during the shift.
The following equipment is Out-Of-Service: SAFW Pump C (TS 3.7.5 Condition E Action E.1) is OOS for breaker maintenance, and the Condensate Booster Pump A is OOS for thrust bearing replacement.
Shortly after taking the watch, a 30 gpm CCW System leak will develop on the Boric Acid Evaporator Air Ejector Condenser. Approximately two minutes afterwards CCW Pump B will trip, and CCW Pump A will automatically start. The operator will start a Reactor Makeup Water Pump and refill the CCW Surge Tank in accordance with AR-A-17, MOTOR OFF RCP CCWP, and AP-CCW.2, LOSS OF CCW DURING POWER OPERATION. Ultimately, the leak will be isolated by coordination between the control room and field operators. The operator will address Technical Specification LCO 3.7.7, COMPONENT COOLING WATER SYSTEM.
After this, Feed Regulating Valve (FRV) A will start to oscillate to the point that G-3, S/G A LEVEL DEVIATION +/-7% will alarm. The operator will take manual control of the valve and control S/G A level using the guidance of the Annunciator Response Procedure, A-503.1, EMERGENCY AND ABNORMAL OPERATING PROCEDURES USERS GUIDE, or AP-FW.1, ABNORMAL MFW PUMP FLOW OR NPSH. The FRV A controller will remain in MANUAL until after the power reduction is performed in the next event.
Subsequently, Condensate Booster Pump C will trip. The operator will respond in accordance with AR-AA-10, CONDENSATE BOOSTER PUMP TRIP, and implement AP-FW.1, ABNORMAL MFW PUMP FLOW OR NPSH. Ultimately, the crew will enter AP-TURB.5, RAPID LOAD REDUCTION, and S-3.1, BORON CONCENTRATION CONTROL and lower power to less than 70%. During the load reduction FRV A will be controlled in MANUAL. When the plant is stabilized the FRV A controller will be repaired, returned to AUTO and the scenario will continue.
Shortly afterwards, Control Rod F-2 in Control Bank A will drop into the core. The operator will respond by taking the Rod Control Switch to MANUAL and by lowering Turbine Load in MANUAL using AR-C-14, ROD BOTTOM, AR-E-28, POWER RANGE ROD DROP -5%/5SEC, and AP-RCC.3, DROPPED ROD RECOVERY. The operator will address Technical Specification 3.1.4, ROD GROUP ALIGNMENT LIMITS, and 3.2.4, QUADRANT POWER TILT RATIO (QPTR).
While the dropped rod is being investigated, a second Control Rod will drop, and the operator will manually trip the reactor in accordance with the previous guidance in Step 1 of AP-RCC.3 using the guidance of A-503.1, EMERGENCY AND ABNORMAL OPERATING PROCEDURES USERS GUIDE.
On the reactor trip, a 200 gpm Steam Generator Tube Rupture will occur in S/G A. The crew will implement E-0, REACTOR TRIP OR SAFETY INJECTION, and then transition to ES-0.1, REACTOR TRIP RESPONSE when it is determined that Safety Injection has not actuated nor is needed. While in ES-0.1, the crew will manually actuate Safety Injection and transition back to E-0 and then ultimately to E-3, STEAM GENERATOR TUBE RUPTURE, to isolate S/G A, cooldown the RCS and depressurize the RCS.
Depending on the pace at which the crew proceeds through E-3, either both PRZR Spray Valves or one PRZR PORV will be used to complete the RCS depressurization. If the PRZR Spray Valves are used, one valve will fail open and require that the operator stop one or both Reactor Coolant
Form 3.3-1 Scenario Outline (Rev_083122)
Pumps to stop the depressurization. If one PRZR PORV is used, PRZR PORV PCV-431C will fail to open if the operator attempts to open it. When the operator attempts to open PRZR PORV PCV-430 it will open and fail open requiring the operator to close the block valve to isolate the leak.
The scenario will terminate at Step 22.b of E-3, after the operator has stopped the SI and RHR Pumps.
Critical Tasks:
CT-1: Isolate feedwater flow into and steam flow from the ruptured SG (A) so that minimum P between the B SG and A SG is not less than 250 psid once target temperature is reached BEFORE a transition to ECA-3.1 occurs Initiating Cue:
Rising narrow Range Level in S/G A, PRZR pressure and level lowering, Charging flow rising, Radiation alarms on R-47 and R-31 Performance Feedback:
A ARV indicates closed, S/G A TDAFW pump steam supply valve (MOV-3505A) control switch is in PULL STOP, MDAFW pump discharge valve (MOV-4007) indicates CLOSED, MDAFW Pump A Control Switch is in PULL STOP, TDAFW pump flow control valve (AOV-4297) indicates closed, EO Acknowledges direction to perform ATT-16.0 Success Path:
Isolate S/G A from the Control Room and direct EO to locally isolate S/G A using ATT-16.0 per E-3 (E-3 Steps 4-6) such that minimum P between S/G B and S/G A is not less than 250 psid once target temperature is reached at Step 16 of E-3 (Entry into ECA-3.1 at Step 16 RNO).
Measurable Performance Standard:
Expected Actions: (1) Adjust S/G A ARV controller to 1050 psig in AUTO, (2) Place S/G A TDAFW pump steam supply valve (MOV-3505A) control switch in PULL STOP, (3) Close MDAFW A pump discharge valve (MOV-4007), (4) Close TDAFW pump flow control valve (AOV-4297) to S/G A (5) direct EO to locally isolate S/G A using ATT-16.0.
Safety Significance: Failure to isolate the ruptured SG causes a loss of P between the ruptured SG and the intact SG. Upon a loss of P, the crew must transition to a contingency procedure that constitutes an incorrect performance that necessitates the crew taking compensating action which complicates the event mitigation strategy. If the crew fails to isolate steam from the SG, or feed flow into the SG, the ruptured SG pressure will tend to decrease to the same pressures as the intact SG, requiring a transition to a contingency procedure, and delaying the stopping of RCS leakage into the SG.
CT-2: While in EOP-E-3, establish/maintain an RCS temperature so that transition from E-3 does not occur because the RCS temperature is either (1) Too high to maintain 20°F of RCS Subcooling OR (2) below 311°F (RCS Integrity Orange Path Limit)
Initiating Cue:
Step 10 of E-3 directs the crew to initiate RCS Cooldown
Form 3.3-1 Scenario Outline (Rev_083122)
Performance Feedback:
Table from Step 10.a of E-3, CETs on PPCS, MSIV position indicating lights, B ARV position indicating lights, Steam flow noise heard in the Control Room, S/G B Narrow Range level, AFW flow indicator to S/G B Success Path:
The crew determines the target temperature based on the Table in Step 10 of E-3 and initiates a cooldown to the target temperature using ARV B, then stops the cooldown and stabilizes RCS temperature at the target temperature (Step 10-15 of E-3) BEFORE transition to ECA-3.1 (At Step 17 of E-3), FR-P.1 (Orange or Red path on Integrity CSF) or FR-S.1 (Orange Path on Subcriticality CSF).
Measurable Performance Standard:
Expected Actions: (1) Determine the target temperature based on the Table in Step 10 of E-3, (2) Manually open S/G B ARV at maximum rate, (3) Control feed flow to maintain narrow range level
> 7% in S/G B or > 200 gpm AFW flow, (4) Manually close S/G B ARV when CET at target temperature (5) stabilize CET temperatures.
Safety Significance: Failure to establish and maintain the correct RCS temperature during a SGTR leads to a transition from E-3 to a contingency procedure. This failure constitutes an incorrect performance that necessitates the operator taking compensating action that would unnecessarily complicate the event mitigation strategy and delay the stopping of RCS leakage into the SG. Terminating the RCS cooldown before reaching the target temperature prevents achieving the minimum RCS subcooling. Failure to achieve the required RCS subcooling results in a condition that forces the crew to transition to contingency ECA-3.1, thereby delaying the RCS depressurization and SI termination. In addition to achieving the minimum target temperature, the crew must maintain that temperature to avoid a similar delay. Terminating the cooldown too late challenges either the subcriticality CSF or the integrity CSF. Because the crew is directed to cool down at the maximum rate, late termination of cooldown could force the RCS temperature low enough to challenge the integrity CSF and require entry into FR-P.1. The transition also delays RCS depressurization and SI termination.
CT-3: Depressurize the RCS to meet SI termination criteria before Steam Generator Overfill is reached based on Water in the Steam Lines (Insight THLECELL (118) = 100)
Initiating Cue:
Steps 18-19 of E-3 directs the crew to depressurize the RCS to minimize break flow and refill the PRZR Performance Feedback:
PRZR Spray Valve position indication or PRZR PORV position indication, PRZR pressure indication, ruptured S/G level, CET temperature indication on PPCS, PRZR level Success Path:
The crew will use the PRZR Spray valves if S/G A Narrow Range level is less than 90% (Step 18 of E-3), or one PRZR PORV if S/G A Narrow Range level is greater than 90% (Step 19 of E-3). In either case, RCS pressure will be lowered to meet specific depressurization stop criteria and SI termination criteria of Step 21 of E-3.
Form 3.3-1 Scenario Outline (Rev_083122)
Measurable Performance Standard:
Expected Actions: (1) Open both PRZR Spray Valves fully, OR (2)
Open one PRZR PORV Safety Significance: Failure to stop reactor coolant leakage into a ruptured SG by depressurizing the RCS (when it is possible to do so) needlessly complicates mitigation of the event. It also constitutes a significant reduction of safety margin beyond that irreparably introduced by the scenario. A SGTR allows radioactive RCS inventory to leak into the SG. As a result, the SG inventory, radioactivity, and pressure increase. If the primary-to-secondary leakage is not stopped, the SG pressure increases until either the SG ARV or the safety valve(s) opens, releasing radioactivity to the environment. If the leakage continues, the SG inventory increase leads to water release through the ARV or safety valve(s) or to SG overfill, which could cause an unisolable fault in the ruptured SG, greatly complicating mitigation.
"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 (Spare Scenario - Not Used)
(Rev_090222)
Facility:
Ginna Scenario No.:
4 Op Test No.:
N22-1 Scenario Source: New Examiners:
Operators:
(SRO)
(RO)
(BOP)
Initial Conditions:
The plant is at 31.7% power (MOL) and has been for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It is expected to raise power to 100% on this shift.
Turnover:
The following equipment is Out-Of-Service: The TD AFW Pump (TS 3.7.5 Condition C Action C.1) is OOS for oil cooler replacement, and the Condensate Booster Pump A is OOS for thrust bearing replacement.
Critical Tasks:
See Below Event No.
Malf.
No.
Event Type*
Event Description 1
MAL CVC10A I(MC)-RO I-SRO VCT Level LT-112 fails LOW 2
NA R-RO N-BOP N-SRO Load Ascension 3
MAL SGN03F I-BOP I(TS)-SRO Steam Generator Pressure PT-483A fails LOW 4
MAL EDS14A C-RO C-BOP C(TS)-SRO Open Phase 7T Line 5
MAL TUR03 TUR17A C-RO C-BOP C-SRO Turbine Trip w/one Stop Valve failing to auto CLOSE 6
MAL FDW09B M-ALL B Feed Line Rupture in Containment (CT-1) 7 MAL RPS07K C(MC)-BOP C-SRO MDAFW Pump A fails to START in AUTO (CT-2)
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec, (MC) Manual Control
Form 3.3-1 Scenario Outline (Spare Scenario - Not Used)
(Rev_090222)
Ginna 2022 NRC Scenario #4 The plant is at 31.7% power (MOL) and has been for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It is expected to raise power to 100% on this shift.
The following equipment is Out-Of-Service: The TD AFW Pump (TS 3.7.5 Condition C Action C.1) is OOS for oil cooler replacement, and the Condensate Booster Pump A is OOS for thrust bearing replacement.
Shortly after taking the watch and before the load ascension is initiated, VCT Level transmitter LT-112 will fail LOW. The operator will respond in accordance with the guidance in A-503.1, EMERGENCY AND ABNORMAL OPERATING PROCEDURES USERS GUIDE, and manually control VCT level. The operator may address AR-A-2, VCT LEVEL 14 % 86, in order to stabilize VCT level. Once the VCT level is stabilized, LT-112 will be returned to service.
Afterwards, the operator will raise power in accordance with O-5.2, LOAD ASCENSION. The operator will address S-3.1, BORON CONCENTRATION CONTROL, to start the load ascension using Alternate Dilute.
Subsequently, Steam Generator Pressure Channel PT-483 will fail LOW. The operator will respond in accordance with AR-G-13, LO STEAM PRESSURE LOOP B 600 PSI, and then enter ER-INST.1, REACTOR PROTECTION BISTABLE DEFEAT AFTER INSTRUMENTATION LOOP FAILURE to defeat the channel. The operator will address Technical Specification LCO 3.3.2, ENGINEERED SAFETY FEATURE ACTUATION SYSTEM (ESFAS) INSTRUMENTATION, and Technical Specification LCO 3.3.3, POST ACCIDENT MONITORING (PAM) INSTRUMENTATION.
Shortly afterwards, an Open Phase will be detected on the 7T line. The crew will respond in accordance with AR-PPCS-7TOPD, 7T OPEN PHASE DETECTION, trip the 52/7T1352 Breaker and enter AP-ELEC.1, LOSS OF 12A AND/OR 12B BUSSES. This will result in de-energizing Bus 14/18, Diesel Generator A automatically starting and repowering these buses and a loss of normal letdown. The crew will use HC-1.0, LOSS OF LETDOWN, to control RCS inventory when Normal Letdown isolates and restore Normal Letdown using ATT-9.0, ATTACHMENT LETDOWN. The operator will address Technical Specification LCO 3.8.1, AC SOURCES - MODES 1, 2, 3, AND 4, and Technical Requirement TR 3.8.1, OFFSITE POWER SOURCES.
After this, a loss of lube oil will occur on the Main Turbine causing the turbine to trip on low lube oil pressure, however, one Turbine Stop Valve will remain OPEN. The crew will respond per AR-K-9, TURBINE BEARING OIL LO PRESS TRIP 6PSI, and enter AP-TURB.1, TURBINE TRIP WITHOUT RX TRIP REQUIRED. The operator will need to manually trip the Turbine to close the open Turbine Stop Valve and manually trip the reactor when the Turbine Stop Valve fails to close. Once the reactor is tripped and the immediate actions have been taken, the MSIVs will be closed.
On the reactor trip, a large Feed Line Rupture will occur inside Containment, causing Safety Injection to actuate. The crew will enter E-0, REACTOR TRIP AND SAFETY INJECTION. MDAFW Pump A will fail to start automatically and will need to be started manually.
The crew will transition from E-0 to E-2, FAULTED STEAM GENERATOR ISOLATION, and isolate the B Steam Generator. Upon completion of E-2, the crew will transition to E-1, LOSS OF REACTOR OR SECONDARY COOLANT. While in E-1, the SI Termination Criteria will be met, and the operator will transition to ES-1.1, SI TERMINATION either on the Foldout Page Criteria #4 or at Step 12.e.
Form 3.3-1 Scenario Outline (Spare Scenario - Not Used)
(Rev_090222)
The scenario will terminate at Step 8 of ES-1.1, after the operator has stopped the SI/RHR Pumps and been briefed on the SI Reinitiation Criteria.
Critical Tasks:
CT-1: Isolate the Faulted Steam Generator before transitioning out of E-2 Initiating Cue:
Steam pressure in S/G B lowers uncontrollably as Containment Pressure rises, AFW flow continues to be delivered to S/G B Performance Feedback:
Indication that AFW flow to S/G B has stopped, RCS uncontrolled cooldown stops Success Path:
Isolate AFW flow to S/G B per step 4 and isolate steam flow per step 5 of E-2.
Measurable Performance Standard:
Expected Actions: (1) Close S/G B MDAFW pump discharge valve (MOV-4008), (2) Place MDAFW Pump B control switch in PULL STOP (3) Direct the EO to perform ATT-10.0, ATTACHMENT FAULTED S/G.
Safety Significance: Failure to isolate a Faulted SG that can be isolated causes challenges to the Critical Safety Functions that would not otherwise occur. Failure to isolate flow could result in an unwarranted Orange or Red Path condition on RCS Integrity, Subcriticality (if cooldown is allowed to continue uncontrollably) and/or Containment (if the break is inside Containment). In this case, continuation of AFW flow to the faulted SG results in the release of additional mass and energy to the containment. This additional release may result in the containment CSF not being satisfied when it otherwise would have been satisfied.
CT-2: Establish 200 gpm AFW flow to the A Steam Generator before transitioning out of E-0 Initiating Cue:
Annunciation that SI has occurred, Total AFW flow indicates less than minimum required, Breaker status lights for MDAFW Pump A indicate the pump is OFF Performance Feedback:
Breaker status lights for MDAFW Pump A indicate the pump is ON, AFW flow to S/G A is 200 gpm Success Path:
The operator starts MDAFW Pump A and verifies 200 gpm of AFW flow to S/G A BEFORE transition to FR-H.1 is made at Step 9 RNO of E-0.
Measurable Performance Standard:
Expected Actions: Manually start MDAFW A after it automatically failed to start.
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. Therefore, failure to manually establish the minimum required AFW flow rate also represents a failure by the crew to demonstrate the
Form 3.3-1 Scenario Outline (Spare Scenario - Not Used)
(Rev_090222) following abilities: (1) Effectively direct or manipulate engineered safety feature (ESF) controls that would prevent (degraded emergency core cooling system (ECCS) capacity), (2)
Recognize a failure or an incorrect automatic actuation of an ESF system or component, and (3)
Take one or more actions that would prevent a challenge to plant safety. Additionally, under the postulated plant conditions, failure to manually establish the minimum required AFW flow rate (when it is possible to do so) results in a significant reduction of safety margin beyond that irreparably introduced by the scenario. Finally, failure to manually actuate AFW under the postulated conditions is a violation of the facility license condition.
"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:
Ginna K/A Catalog Rev. 3 Rev.
2 Date of Exam:
10/17/2022 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
3 3
3 3
3 3
18 3
3 6
2 1
1 2
2 1
1 8
2 2
4 Tier Totals 4
4 5
5 4
4 26 5
5 10
- 2.
Plant Systems 1
2 3
1 3
2 2
3 3
3 4
2 28 3
2 5
2 1
1 1
1 1
1 1
1 0
0 1
9 0
2 1
3 Tier Totals 3
4 2
4 3
3 4
4 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 (Ginna)
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
(EPE 7) EA1.07 - Ability to operate and/or monitor the following as they apply to a Reactor Trip: MT/G.
(CFR: 41.7 / 45.5 / 45.6) 3.3 1
2 000008 (APE 8)
Pressurizer Vapor Space Accident X
AK1.05 - Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to a Pressurizer Vapor Space Accident: Probable PZR steam space leakage paths other than PORV or code safety.
(CFR: 41.8 / 41.10 / 45.3) 3.6 2
3 000009 (EPE 9)
Small Break LOCA X
EA2.29 - Ability to determine and/or interpret the following as they apply to a Small-Break LOCA: CVCS pump indicating lights for determining pump status.
(CFR: 43.5 / 45.13) 3.2 3
4 000011 (EPE 11)
Large Break LOCA X
G2.4.3 - Ability to identify post-accident instrumentation.
(CFR: 41.6 / 45.4) 3.7 4
5 000015 (APE 15)
Reactor Coolant Pump Malfunctions X
AK3.07 - Knowledge of the reasons for the following responses and/or actions as they apply to Reactor Coolant Pump Malfunctions: Ensuring that S/G levels are controlled properly for natural circulation enhancement.
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.7 5
6 000022 (APE 22)
Loss of Reactor Coolant Makeup X
AA1.01 - Ability to operate and/or monitor the following as they apply to Loss of Reactor Coolant Makeup: CVCS.
(CFR: 41.7 / 45.5 / 45.6) 3.8 6
7 000025 (APE 25)
Loss of Residual Heat Removal System X
AA2.17 - Ability to determine and/or interpret the following as they apply to the Loss of the Residual Heat Removal System:
Applicable TSs.
(CFR: 43.5 / 45.13) 3.2 7
8 000026 (APE 26)
Loss of Component Cooling Water X
AK2.03 - Knowledge of the relationship between Loss of Component Cooling Water and the following systems or components: RHR.
(CFR: 41.8 / 41.10 / 45.3) 3.8 8
9 000027 (APE 27)
Pressurizer Pressure Control System Malfunction X
G2.1.19 - Ability to use available indications to evaluate system or component status.
(CFR: 41.10 / 45.12) 3.9 9
10 000029 (EPE 29)
Anticipated Transient Without Scram X
EK2.14 - Knowledge of the relationship between Anticipated Transient Without Scram and the following systems or components: AMSAC.
(CFR: 41.7 / 45.7) 4.2 10 11 000054 (APE 54; CE E06) Loss of Main Feedwater X
(APE 54) AA2.04 - Ability to determine and/or interpret the following as they apply to Loss of Main Feedwater: Proper operation of AFW pumps and regulating valves.
(CFR: 43.5 / 45.13) 4.0 11 12 000055 (EPE 55)
Station Blackout X
G2.1.30 - Ability to locate and operate components, including local controls.
(CFR: 41.7 / 45.7) 4.4 12 13 000056 (APE 56)
Loss of Offsite Power X
AK3.02 - Knowledge of the reasons for the following responses and/or actions as they apply to Loss of Offsite Power: Actions contained in AOPs.
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 4.1 13 14 000057 (APE 57)
Loss of Vital AC Instrument Bus X
AK1.01 - Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Loss of Vital AC Electrical Instrument Bus: Effect of a loss of power to instruments powered by a vital instrument bus on plant operation.
(CFR: 41.8 / 41.10 / 45.3) 4.3 14
15 000058 (APE 58)
Loss of DC Power X
AA1.04 - Ability to operate and/or monitor the following as they apply to Loss of DC Power: AC distribution system breakers.
(CFR: 41.7 / 45.5 / 45.6) 3.6 15 16 000062 (APE 62)
Loss of Service Water X
AK1.01 - Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Loss of Service Water: Effect on loads cooled by service water.
(CFR: 41.8 / 41.10 / 45.3) 3.8 16 17 (W E04) LOCA Outside Containment X
EK2.05 - Knowledge of the relationship between LOCA Outside Containment and the following systems or components: RCS leakage paths to outside containment.
(CFR: 41.7 / 41.8 / 45.2 / 45.4) 4.0 17 18 (BW E04; W E05)
Inadequate Heat Transfer - Loss of Secondary Heat Sink X
(W E05) EK3.16 - Knowledge of the reasons for the following responses and/or actions as they apply to Loss of Secondary Heat Sink: Establishing maximum charging flow.
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.5 18 19 000008 (APE 8)
Pressurizer Vapor Space Accident 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 76 20 000038 (EPE 38)
Steam Generator Tube Rupture X
G2.4.18 - Knowledge of the specific bases for emergency and abnormal operating procedures.
(CFR: 41.10 / 43.1 / 45.13) 4.0 77 21 000040 (APE 40; BW E05; CE E05; W E12) Steam Line Rupture -
Excessive Heat Transfer X
(APE 40) AA2.07 - Ability to determine and/or interpret the following as they apply to a Steamline Rupture: Occurrence of an ATWS during a steamline rupture event.
(CFR: 41.10 / 43.5 / 45.13) 3.8 78 22 000065 (APE 65)
Loss of Instrument Air 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) 4.4 79 23 000077 (APE 77)
Generator Voltage and Electric Grid Disturbances X
AA2.09 - Ability to determine and/or interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: Status of EDGs.
(CFR: 41.5 / 43.5 / 45.5 / 45.7 / 45.8) 3.9 80 24 (W E11) Loss of Emergency Coolant Recirculation X
EA2.10 - Ability to determine and/or interpret the following as they apply to Loss of Emergency Coolant Recirculation: S/G level, pressure, and/or feedwater flow.
(CFR: 41.10 / 43.5 / 45.13) 3.3 81 K/A Category Totals:
3 3
3 3
6 6
Group Point Total:
24 ES-4.1-PWR PWR Examination Outline (Ginna)
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 000028 (APE 28)
Pressurizer (PZR)
Level Control Malfunction X
AK2.08 - Knowledge of the relationship between a Pressurizer Level Control Malfunction and the following systems or components: RCS.
(CFR: 41.7 / 45.7) 3.4 19
26 000036 (APE 36; BW/A08) Fuel-Handling Incidents X
(APE 36) AK3.04 - Knowledge of the reasons for the following responses and/or actions as they apply to Fuel Handling Incidents:
Establishing containment isolation or closure.
(CFR: 41.10 / 45.6 / 45.13) 3.9 20 27 000074 (EPE 74; W E06 & E07)
Inadequate Core Cooling X
(EPE 74) EK1.09 - Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Inadequate Core Cooling: Calculation of volume of water added to the RCS using tank level indicators.
(CFR: 41.8 / 41.10 / 45.3) 3.3 21 28 000076 (APE 76)
High Reactor Coolant Activity 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 22 29 (W E01 & E02)
Rediagnosis & SI Termination X
(W E02) EA1.15 - Ability to operate and/or monitor the following as they apply to SI Termination: ECCS.
(CFR: 41.5 to 41.8 / 45.5 to 45.8) 3.9 23 30 (W E15)
Containment Flooding X
EK3.05 - Knowledge of the reasons for the following responses and/or actions as they apply to Containment Flooding: Determining the source of the water in the sump.
(CFR: 41.5 /41.10 / 45.6 / 45.13) 3.4 24 31 (BW E08; W E03)
LOCA Cooldown -
Depressurization X
(W E03) EA2.09 - Ability to determine and/or interpret the following as they apply to LOCA Cooldown and Depressurization:
Containment pressure.
(CFR: 41.10 / 43.5 / 45.13) 3.7 25 32 (CE A11**; W E08)
RCS Overcooling -
Pressurized Thermal Shock X
(W E08) EA1.10 - Ability to operate and/or monitor the following as they apply to Pressurized Thermal Shock: RCS.
(CFR: 41.5 to 41.8 / 45.5 to 45.8) 3.6 26 33 000024 (APE 24)
Emergency Boration X
G2.1.37 - Knowledge of procedures, guidelines, or limitations associated with reactivity management.
(CFR: 41.1 / 41.5 / 41.10 / 43.6 / 45.6) 4.6 82 34 000037 (APE 37)
Steam Generator Tube Leak X
AA2.17 - Ability to determine and/or interpret the following as they apply to a Steam Generator Tube Leak: PZR level and/or pressure.
(CFR: 41.7 / 41.10 / 43.5 / 45.13) 3.8 83 35 000067 (APE 67)
Plant Fire On Site X
G2.4.6 - Knowledge of emergency and abnormal operating procedures major action categories.
(CFR: 41.10 / 43.5 / 45.13) 4.7 84 36 000068 (APE 68; BW A06) Control Room Evacuation X
(APE 68) AA2.03 - Ability to determine and/or interpret the following as they apply to Control Room Evacuation: T-hot, T-cold, and in-core temperatures.
(CFR: 41.10 / 43.5 / 45.13) 3.9 85 000001 (APE 1)
Loss Continuous Rod Withdrawal 000003 (APE 3)
Dropped Control Rod 000005 (APE 5)
Inoperable/Stuck Control Rod 000032 (APE 32)
Loss of Source Range Nuclear Instrumentation
000033 (APE 33)
Loss of Intermediate Range Nuclear Instrumentation 000051 (APE 51)
Loss of Condenser Vacuum 000059 (APE 59)
Accidental Liquid Radwaste Release 000060 (APE 60)
Accidental Gaseous Radwaste Release 000061 (APE 61)
Area Radiation Monitoring System Alarms 000069 (APE 69; W E14) Loss of Containment Integrity 000078 (APE 78*)
RCS Leak (W E13) Steam Generator Overpressure (W E16) High Containment Radiation (BW A01) Plant Runback (BW A02 & A03)
Loss of NNI-X/Y (BW A04) Turbine Trip (BW A05)
Emergency Diesel Actuation (BW A07) Flooding (BW E03)
Inadequate Subcooling Margin (BW E09; CE A13**;
W E09 & E10)
Natural Circulation (BW E13 & E14)
EOP Rules and Enclosures (CE A16) Excess RCS Leakage (CE E09) Functional Recovery
(CE E13*) Loss of Forced Circulation /
LOOP / Blackout K/A Category Totals:
1 1
2 2
3 3
Group Point Total:
12 ES-4.1-PWR PWR Examination Outline (Ginna)
Emergency and Abnormal Plant EvolutionsTier 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
K5.10 - Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Reactor Coolant Pump System:
Starting an RCP while RCS is cold.
(CFR: 41.5 / 45.7) 3.7 27 38 004 (SF1; SF2 CVCS) Chemical and Volume Control X
A4.19 - Ability to manually operate and/or monitor in the control room:
CVCS letdown orifice isolation valve and valve control switches.
(CFR: 41.5 to 41.7 / 45.5 to 45.8) 3.6 28 39 005 (SF4P RHR)
K2.01 - Knowledge of electrical power supplies to the following:
RHR pumps.
(CFR: 41.7) 4.1 29 40 006 (SF2; SF3 ECCS) Emergency Core Cooling X
COMPONENT (Pumps -
Centrifugal): 191004 K1.12 -
Runout of a centrifugal pump (definition, indications, causes, effects, and corrective measures).
(CFR: 41.3) 2.7 30 41 007 (SF5 PRTS)
Pressurizer Relief/Quench Tank X
K4.03 - Knowledge of Pressurizer Relief Tank / Quench Tank System design features and/or interlocks that provide for the following:
Nitrogen cover gas.
(CFR: 41.7) 2.7 31 42 008 (SF8 CCW)
Component Cooling Water X
A3.01 - Ability to monitor automatic features of the Component Cooling Water System, including: Setpoints for normal operations, warnings, and trips that are applicable to the CCWS.
(CFR: 41.7 / 45.5) 3.6 32
43 008 (SF8 CCW)
Component Cooling Water X
K1.01 - Knowledge of the physical connections and/or cause and effect relationships between the Component Cooling Water System and the following systems: SWS.
(CFR: 41.3 to 41.9 / 45.7 to 45.9) 4.0 33 44 010 (SF3 PZR PCS)
Pressurizer Pressure Control X
K2.05 - Knowledge of electrical power supplies to the following:
Pressure channels.
(CFR: 41.7) 3.3 34 45 010 (SF3 PZR PCS)
Pressurizer Pressure Control X
K4.03 - Knowledge of Pressurizer Pressure Control System design features and/or interlocks that provide for the following:
Overpressure control.
(CFR: 41.7) 3.9 35 46 012 (SF7 RPS)
Reactor Protection X
A1.01 - Ability to predict and/or monitor changes in parameters associated with operation of the Reactor Protection System, including: Trip setpoint adjustment.
(CFR: 41.5 / 45.5) 3.5 36 47 012 (SF7 RPS)
Reactor Protection X
A2.01 - Ability to (a) predict the impacts of the following on the Reactor Protection System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Faulty bistable operation.
(CFR: 41.5 / 43.5 / 45.3 / 45.5) 3.8 37 48 013 (SF2 ESFAS)
Engineered Safety Features Actuation X
K4.04 - Knowledge of Engineered Safety Features Actuation System design features and/or interlocks that provide for the following: AFW actuation/reset.
(CFR: 41.2 / 41.6 / 41.7) 4.1 38 49 013 (SF2 ESFAS)
Engineered Safety Features Actuation X
K6.12 - Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Engineered Safety Features Actuation System:
LOCA.
(CFR: 41.6 / 41.7 / 41.8 / 45.5 to 45.8) 4.1 39 50 022 (SF5 CCS)
Containment Cooling X
A1.05 - Ability to predict and/or monitor changes in parameters associated with operation of the Containment Cooling System, including: Lights and alarms.
(CFR: 41.5 / 45.5) 3.3 40
51 026 (SF5 CSS)
K3.01 - Knowledge of the effect that a loss or malfunction of the Containment Spray System will have on the following systems or system parameters: CCS.
(CFR: 41.7 / 45.6) 3.8 41 52 039 (SF4S MSS)
Main and Reheat Steam X
K5.08 - Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Main and Reheat Steam System: Effect of steam removal on reactivity.
(CFR: 41.5 / 45.7) 4.0 42 53 059 (SF4S MFW)
Main Feedwater X
K1.08 - Knowledge of the physical connections and/or cause and effect relationships between the Main Feedwater System and the following systems: Heater drains.
(CFR: 41.2 to 41.9 / 45.7 / 45.8) 3.0 43 54 061 (SF4S AFW)
Auxiliary /
Emergency Feedwater X
A4.01 - Ability to manually operate and/or monitor in the control room:
AFW pump.
(CFR: 41.7 / 45.5 to 45.8) 4.2 44 55 062 (SF6 ED AC) AC Electrical Distribution X
A3.05 -Ability to monitor automatic operation of the AC Electrical Distribution System, including:
Safety-related actuations.
(CFR: 41.7 / 45.5) 4.1 45 56 062 (SF6 ED AC) AC Electrical Distribution X
K2.02 - Knowledge of electrical power supplies to the following:
Breaker control power.
(CFR: 41.7) 3.5 46 57 063 (SF6 ED DC)
DC Electrical Distribution X
A2.05 - Ability to (a) predict the impacts of the following on the DC Electrical Distribution System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Loss of all AC.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 4.2 47 58 064 (SF6 EDG)
G2.2.12 - Knowledge of surveillance procedures.
(CFR: 41.10 / 43.2 / 45.13) 3.7 48
59 073 (SF7 PRM)
Process Radiation Monitoring X
A1.01 - Ability to predict and/or monitor changes in parameters associated with operation of the Process Radiation Monitoring System, including: Radiation levels.
(CFR: 41.5 / 45.5) 3.5 49 60 073 (SF7 PRM)
Process Radiation Monitoring X
A4.04 - Ability to manually operate and/or monitor in the control room:
Alarm and/or interlock setpoint checks and adjustments.
(CFR: 41.7 / 45.8 / 45.9) 3.2 50 61 076 (SF4S SW)
A3.04 - Ability to monitor automatic features of the Service Water SYSTEM, including: Automatic start features associated with SWS pump controls.
(CFR: 41.7 / 45.5) 3.7 51 62 078 (SF8 IAS)
Instrument Air X
A2.03 - Ability to (a) predict the impacts of the following on the Instrument Air System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Cooling water malfunction.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 2.5 52 63 078 (SF8 IAS)
Instrument Air X
A4.02 - Ability to manually operate and/or monitor in the control room:
Instrument air compressors.
(CFR: 41.7 / 45.5 to 45.8) 3.2 53 64 103 (SF5 CNT)
Containment X
K6.09 - Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Containment System: CPS.
(CFR: 41.7 / 45.7) 3.4 54 65 003 (SF4P RCP)
Reactor Coolant Pump X
A2.02 - Ability to (a) predict the impacts of the following on the Reactor Coolant Pump System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Conditions that exist for an abnormal shutdown of an RCP compared to a normal shutdown of an RCP.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.8 86
66 005 (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 87 67 006 (SF2; SF3 ECCS) Emergency Core Cooling X
A2.11 - Ability to (a) predict the impacts of the following on the Emergency Core Cooling System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Rupture of ECCS header.
(CFR: 41.5 / 43.5 / 45.4 / 45.5) 3.7 88 68 063 (SF6 ED DC)
DC Electrical Distribution X
G2.2.45 - Ability to determine and/or interpret TS with 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 89 69 103 (SF5 CNT)
Containment X
A2.03 - Ability to (a) predict the impacts of the following on the Containment System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Containment isolation signal.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 4.0 90 025 (SF5 ICE) Ice Condenser 053 (SF1; SF4P ICS*) Integrated Control K/A Category Totals:
2 3
1 3
2 2
3 6
3 4
4 Group Point Total:
33 ES-4.1-PWR PWR Examination Outline (Ginna)
Emergency and Abnormal Plant EvolutionsTier 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
K4.07 - Knowledge of Control Rod Drive System design features and/or interlocks that provide for the following: Rod control stops and permissives.
(CFR: 41.6) 4.0 55 71 011 (SF2 PZR LCS)
Pressurizer Level Control X
K3.03 - Knowledge of the effect that a loss or malfunction of the Pressurizer Level Control System will have on the following systems or system parameters: PZR PCS.
(CFR: 41.7 / 45.6) 3.7 56 72 014 (SF1 RPI) Rod Position Indication X
A1.02 - Ability to predict and/or monitor changes in parameters associated with operation of the Rod Position Indication System, including: RPI.
(CFR: 41.5 to 41.7 / 45.5) 3.7 57 73 015 (SF7 NI) Nuclear Instrumentation X
A2.02 - Ability to (a) predict the impacts of the following on the Nuclear Instrumentation System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Faulty or erratic operation of detectors or compensating components.
(CFR: 41.5 / 43.5 / 45.3 / 45.5) 3.5 58 74 027 (SF5 CIRS)
Containment Iodine Removal X
K1.01 - Knowledge of the physical connections and/or cause and effect relationships between the Containment Iodine Removal System and the following systems:
CSS.
(CFR: 41.7 to 41.9 / 45.7 / 45.8) 3.2 59 75 034 (SF8 FHS) Fuel Handling Equipment X
K6.06 - Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Fuel Handling Equipment System: Mechanically bound fuel assembly.
(CFR: 41.6 / 41.7 / 43.5 / 45.7) 3.1 60 76 041 (SF4S SDS)
Steam Dump /
Turbine Bypass Control X
COMPONENT (Controllers and Positioners): 191003 K1.03-Operation of valve controllers in manual and automatic modes, including seal-in features.
(CFR: 41.3) 3.1 61
77 045 (SF4S MT/G)
Main Turbine Generator X
K5.28 - Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Main Turbine Generator System: Governor and load limits.
(CFR: 41.5 / 45.7) 2.9 62 78 075 (SF8 CW)
Circulating Water X
K2.01 - Knowledge of electrical power supplies to the following:
Circulating water pumps.
(CFR: 41.7) 2.8 63 79 016 (SF7 NNI)
Nonnuclear Instrumentation X
G2.1.32 - Ability to explain and apply system precautions, limitations, notes, or cautions.
(CFR: 41.10 / 43.2 / 45.12) 4.0 91 80 035 (SF4P SG)
A2.04 - Ability to (a) predict the impacts of the following on the Steam Generator System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Steam flow/feed flow mismatch.
(CFR: 41.5 / 43.5 / 45.3 / 45.5) 3.9 92 81 033 (SF8 SFPCS)
Spent Fuel Pool Cooling X
A2.02 - Ability to (a) predict the impacts of the following on the Spent Fuel Pool Cooling System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: SFPCS malfunction.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.2 93 002 (SF2; SF4P RCS) Reactor Coolant 016 (SF7 NNI)
Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 028 (SF5 HRPS)
Hydrogen Recombiner and Purge Control 029 (SF8 CPS)
Containment Purge
050 (SF9 CRV*)
Control Room Ventilation 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 086 (SF8 FPS) Fire Protection K/A Category Totals:
1 1
1 1
1 1
1 3
0 0
2 Group Point Total:
12 Form 4.1-COMMON Common Examination Outline ES-4.1-COMMON COMMON Examination Outline (Ginna)
Facility:
Ginna Date of Exam:
10/17/2022 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 G2.1.39 Knowledge of conservative decision-making practices.
(CFR: 41.10 / 43.5 / 45.12) 83 3.6 65 G2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, and maintenance of active license status, 10 CFR Part 55.
(CFR: 41.10 / 43.2) 84 3.8 94 G2.1.36 Knowledge of procedures and limitations involved in core alterations.
(CFR: 41.10 / 43.6 / 45.7) 85 4.1 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 G2.2.35 Ability to determine TS for mode of operation.
(CFR: 41.7 / 41.10 / 43.2 / 45.13) 87 3.6 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
G2.2.20 Knowledge of the process for managing troubleshooting activities.
(CFR: 41.10 / 43.5 / 45.13) 89 3.8 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 G2.3.6 Ability to approve liquid or gaseous release permits.
(CFR: 41.13 / 43.4 / 45.10) 91 3.8 98 Subtotal N/A 1
N/A 1
- 4.
Emergency Procedures /
Plan G2.4.26 Knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage.
(CFR: 41.10 / 43.5 / 45.12) 92 3.1 69 G2.4.23 Knowledge of the bases for prioritizing emergency operating procedures implementation.
(CFR: 41.10 / 43.5 / 45.13) 93 4.4 99 G2.4.29 Knowledge of emergency plan implementing procedures.
(CFR: 41.10 / 43.5 / 45.11) 94 4.4 100 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 (Ginna)
Facility:
Ginna Date of Exam:
10/17/2022 Theory (Tier 4) (RO)
Category K/A #
Topic RO Item #
IR Q#
Reactor Theory 192006 (K1.06)
K1.06 - FISSION PRODUCT POISONS: Describe the following processes and state their effect on reactor operations: -- transient xenon.
(CFR: 41.1) 95 3.4 70 192007 (K1.04)
K1.04 - FUEL DEPLETION AND BURNABLE POISONS: Describe how and why boron concentration changes over core life.
(CFR: 41.1) 96 3.4 71 192008 (K1.14)
K1.14 - REACTOR OPERATIONAL PHYSICS (Intermediate Range Operation): Describe reactor power and startup rate response prior to reaching the POAH.
(CFR: 41.1) 97 3.1 72 Subtotal N/A 3
Thermodynamics 193003 (K1.25)
K1.25 - STEAM: Explain and use saturated and superheated steam tables.
(CFR: 41.14) 98 3.4 73 193006 (K1.13)
K1.13 - FLUID STATICS AND DYNAMICS: Explain why flow measurements must be corrected for density changes.
(CFR: 41.14) 99 2.6 74 193008 (K1.22)
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 1/2 EPE W E15 EK2.03 (RO 24)
K/A W E15 EK2.07 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the selected K/A without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected K/A EK2.03 to maintain K/A category balance within the outline.
1/2 EPE W E15 EK3.05 (RO 24)
K/A W E15 EK2.03 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the re-selected K/A without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected K/A EK3.05 to maintain K/A category balance within the outline.
2/1 010 K4.03 (RO 35)
K/A 010 K4.05 rejected due to overlap concerns with the 2022 NRC Operating Test.
CHIEF EXAMINER randomly re-selected K/A K4.03 to maintain K/A category balance within the outline.
2/1 022 A1.05 (RO 40)
K/A 022 A1.04 rejected. Facility was unable to develop an operationally valid and discriminating RO level question to test the selected K/A without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected K/A A1.05 to maintain K/A category balance within the outline.
2/2 COMPONENT (Controllers and Positioners) 191003 K1.03 (RO 61)
K/A 191001 [COMPONENT (Valves)] K1.04 rejected. Facility was unable to develop an operationally valid and discriminating RO level GFE Component question to test the selected K/A without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected K/A 191003 [COMPONENT (Controllers and Positioners)] K1.03 to maintain K/A category balance within the outline.
1/1 EPE W E11 EA2.10 (SRO 81)
K/A W E11 EA2.06 rejected. Facility was unable to develop an operationally valid and discriminating SRO level question to test the selected K/A without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected K/A EA2.10 to maintain K/A category balance within the outline.
1/2 APE 037 AA2.17 (SRO 83)
K/A 037 AA2.10 rejected. Facility was unable to develop an operationally valid and discriminating SRO level question to test the selected K/A without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected K/A AA2.17 to maintain K/A category balance within the outline.
1/2 APE 067 G2.4.6 (SRO 84)
K/A G2.4.16 rejected. Facility was unable to develop an operationally valid and discriminating SRO level question to test the selected K/A without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected K/A G2.4.6 to maintain K/A category balance within the outline.
2/2 033 A2.02 (SRO 93)
K/A 056 A2.12 rejected. Facility was unable to develop an operationally valid and discriminating SRO level question to test the selected Plant System and K/A pair without expending an inordinate amount of resources and time.
CHIEF EXAMINER randomly re-selected Plant System 033 and K/A A2.02 to maintain K/A category balance within the outline.