ML18242A356
| ML18242A356 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/31/2018 |
| From: | NRC Region 1 |
| To: | |
| Shared Package | |
| ML17101A476 | List: |
| References | |
| U01938 | |
| Download: ML18242A356 (35) | |
Text
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
James A. Fitzpatrick Date of Examination:
May 2018 Examination Level: RO Operating Test Number:
17-1 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations M, R Core Thermal Heat Balance Verification Using Turbine Steam Pressure K/A 2.1.19 (3.9), OP-65, RAP-7.3.03 Conduct of Operations D, S Perform Daily Checks Per ST-40D K/A 2.1.18 (3.6), ST-40D Equipment Control P, D, R 14-2 NRC Perform ST-23C, Jet Pump Operability - Two Loop K/A 2.2.12 (3.7), ST-23C Radiation Control M, R Determine Worker Exposure for Emergent Work K/A 2.3.4 (3.2), EN-RP-201 Emergency Plan NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1, randomly selected)
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
James A. Fitzpatrick Date of Examination:
May 2018 Examination Level: SRO Operating Test Number:
17-1 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations M, R Core Thermal Heat Balance Verification Using Turbine Steam Pressure K/A 2.1.19 (3.8), OP-65, RAP-7.3.03 Conduct of Operations D, R Determine Reportability Requirements - Scram with HPCI and RCIC Start K/A 2.1.18 (3.8), NUREG 1022, LS-AA-1400 Equipment Control P, D, R 14-2 NRC Review ST-23C, Jet Pump Operability - Two Loop K/A 2.2.12 (4.1), ST-23C Radiation Control M, R Determine Worker Exposure for Emergent Work and Required Actions K/A 2.3.4 (3.7), EN-RP-201 Emergency Plan M, R Determine Emergency Classification and Initiate Event Notification K/A 2.4.40 (4.5), IAP-1, IAP-2 NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1, randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 water Facility:
James A. Fitzpatrick Date of Examination:
May 2018 Exam Level: RO / SRO-I / SRO-U Operating Test Number:
17-1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. Perform Actions for Fire in Plant - RWR System K/A 202001 A4.01 (3.7/3.7), AOP-28 P, D, A, S 16-1 NRC 4
- c. Roll Main Turbine, Low Bearing Oil Pressure K/A 241000 A4.11 (3.1/3.1), OP-9, ARP 09-5-2-07 M, A, L, S 3
- d. Perform Control Rod Operability Test, CRD Pump Trips K/A 201002 A4.01 (3.5/3.4), ST-20C, AOP-69 M, A, S 1
- g. Switching Relay Room Supply and Exhaust Fans K/A 288000 A4.01 (3.1/2.9), OP-56 D, EN, S 9
- h. Lower Torus Water Level (RO only)
K/A 223001 A2.11 (3.6/3.8), OP-13B D, EN, S 5
In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i. Vent Torus to Lower Primary Containment Pressure K/A 295010 AA1.05 (3.1/3.4), EP-6 P, D, A, E 14-2 NRC 5
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 8/ 4 1/ 1/ 1 1/ 1/ 1 (control room system) 1/ 1/ 1 2/ 2/ 1 3/ 3/ 2 (randomly selected) 1/ 1/ 1
Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-1 Op-Test No.: 17-1 Examiners: ___________________________ Operators:
Initial Conditions: The plant is operating at approximately 100% power. SRV A is inoperable.
Circulating Water pump A operation is degraded.
Turnover: Lower Reactor power to approximately 65% using Recirc and control rods per the provided RMI. Then, secure Circulating Water pump A per OP-4 section F. The procedure is in progress up to step F.2.1.
Event No.
Malf.
No.
Event Type*
Event Description 1
N/A R -
- ATC, SRO Lower Reactor Power with Recirculation Flow and Control Rods OP-26, OP-27 2
NM14:A I - ATC, SRO APRM Fails As-Is ARP 09-5-2-2(44), OP-16 3
N/A N -
PC05:J I - ATC, SRO Drywell Pressure Transmitter Fails High, then Low ARP 09-5-1-3(21), Technical Specifications 5
HP05 I - BOP, SRO HPCI Inadvertently Initiates, Trip Pushbutton Fails to Work AOP-77, AOP-32, Technical Specifications 6
MC01 C - All Loss of Main Condenser Vacuum AOP-31, AOP-1 7
RP01AB RP01BB RP09 M - All Failure of RPS and ARI to Actuate EOP-2, EOP-3 8
SL01 RR13 C -
EG01 TC04 C -
- BOP, SRO Main Generator Trip, Two Turbine Bypass Valves Fail Closed EOP-3 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-1 Op-Test No.: 17-1
- 1. Malfunctions after EOP entry (1-2)
Events 8, 9 2
- 2. Abnormal events (2-4)
Events 5, 6 2
- 3. Major transients (1-2)
Event 7 1
- 4. EOPs entered/requiring substantive actions (1-2)
EOP-2 1
- 5. Entry into a contingency EOP with substantive actions (1 per scenario set)
EOP-3 1
- 6. Pre-identified critical tasks (2) 2 CRITICAL TASK DESCRIPTIONS:
CT-1: Given a failure to scram with Reactor power above 2.5%, the crew will lower Reactor power by one or more of the following methods, in accordance with EOP-3:
Terminating and preventing all RPV injection except SLC, RCIC, and CRD Tripping Recirculation pumps Injecting boron CT-2: Given a failure to scram, the crew will initiate Control Rod insertion, in accordance with EOP-3.
Appendix D Scenario Outline Form ES-D-1 The scenario will begin with the plant operating at approximately 100% power. SRV A is inoperable and Circulating Water pump A operation is degraded. The crew will begin by lowering power with recirculation flow and then control rods to approximately 65%. During the power reduction, it will become evident that APRM A has failed as-is. The crew will bypass APRM A per OP-16. Once the power reduction is complete, the crew will secure Circulating Water pump A per OP-4.
Drywell pressure transmitter 05PT-12A will fail momentarily high, then low. This will cause a half scram on RPS A. The crew will reset the half scram. The SRO will determine the Technical Specification impact.
HPCI will inadvertently initiate. The crew will take action to trip HPCI per AOP-77. The first method (pushbutton) will fail, so the crew will take alternate actions to trip HPCI. The SRO will determine the Technical Specification impact of the resulting HPCI inoperability.
Elevated Main Condenser air in-leakage will occur. Main Condenser vacuum will degrade. The crew will enter AOP-31 and eventually insert a manual Reactor scram.
RPS B will fail to process the scram and ARI will also fail to insert control rods. The crew will enter EOP-2 and EOP-3. The ATWS system will fail to automatically trip the Recirculation pumps when required. The crew will lower Recirculation flow to minimum and then trip the Recirculation pumps. The crew will terminate and prevent injection except CRD, SLC, and RCIC. The crew will inject boron using SLC. The first pump started will trip after a time delay.
The second pump started will operate properly. The Main Turbine will be available until power lowers to approximately 40%, when a spurious turbine trip occurs. As power lowers, two Turbine Bypass Valves will fail closed, challenging Reactor pressure control and Primary Containment control. The crew will be able to manually insert control rods. Either pulling RPS fuses or venting the scram air header will result in all rods inserting.
The scenario will be terminated when control rods are being inserted or are all inserted and Reactor water level is controlled above 0.
Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-2 Op-Test No.: 17-1 Examiners: ___________________________ Operators:
Initial Conditions: The plant is operating at approximately 58% power. SRV A is inoperable.
Condensate pump A and Condensate Booster pump A are ready to start following breaker maintenance.
Turnover: Start Condensate pump A and Condensate Booster pump A per OP-3 section D.8. Then, secure Condensate Booster pump B and Condensate pump B per OP-3 section F.1. Then, perform a control rod pattern adjustment per the provided RMI.
Event No.
Malf.
No.
Event Type*
Event Description 1
N/A N -
N/A R -
- ATC, SRO Perform Control Rod Pattern Adjustment OP-26 3
RR22:B I - SRO RPS Level Transmitter Fails High ARP 09-5-2-60, Technical Specifications 4
SW01:A C -
FW25 C -
- ATC, SRO Condensate Booster Pump A Trip, Condensate Booster Pump B Fails to Start; Delayed Trip of Condensate Booster Pump C AOP-1, EOP-2 6
RP01A RP01B RP09 I - ATC, SRO RPS Fails to Scram, ARI Fails to Automatically Initiate AOP-1, EOP-2 7
HP01 RC02 I - BOP, SRO HPCI and RCIC Fail to Start Automatically OP-15, OP-19, EOP-2 8
CU07 CU10 CU12 Remotes M - All RWCU Steam Leak into Reactor Building; RWCU Fails to Isolate Automatically and Manually EOP-5, EOP-2 9
Override I - ATC, SRO Bypass Opening Jack Motor Fails EOP-2 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-2 Op-Test No.: 17-1
- 1. Malfunctions after EOP entry (1-2)
Events 7 & 9 2
- 2. Abnormal events (2-4)
Events 4, 5, 6 3
- 3. Major transients (1-2)
Event 8 1
- 4. EOPs entered/requiring substantive actions (1-2)
- 5. Entry into a contingency EOP with substantive actions (1 per scenario set)
EOP-2 Emergency Depressurization Leg 1
- 6. Pre-identified critical tasks (2) 2 CRITICAL TASK DESCRIPTIONS:
CT-1: Given the plant operating at power with a loss of Feedwater injection and failure of RPS to scram the Reactor, the crew will manually initiate ARI, in accordance with AOP-1 and/or EOP-3.
CT-2: Given an un-isolable primary system discharging into Secondary Containment and two areas exceeding Maximum Safe Temperatures, the crew will perform an emergency RPV depressurization, in accordance with EOP-5.
Appendix D Scenario Outline Form ES-D-1 The scenario will begin with the plant operating at approximately 58% power. SRV A is inoperable. Condensate pump A and Condensate Booster pump A are ready to start following breaker maintenance. The crew will begin by starting Condensate pump A and Condensate Booster pump A per OP-3 section D.8. Next, the crew will secure Condensate Booster pump B and Condensate pump B per OP-3 section F.1. Then, the crew will perform a control rod pattern adjustment per the provided RMI, which will move four control rods from position 10 to 12 and two control rods from position 08 to 04.
Reactor water level transmitter, 02-3LT-101B, will fail upscale. This transmitter is one of the inputs to the RPS scram function. With the transmitter failed high, the SRO will determine that it cannot perform its scram function, declare that function inoperable, and determine the Technical Specification impact.
All RBCLC flow to Recirculation pump A will isolate. This will cause multiple high temperature alarms. If left unmitigated, this will cause degradation of both Recirculation pump A seals and loss of coolant into the Drywell. The crew will secure Recirculation pump A. If damage has occurred to both pump seals, the crew will also isolate Recirculation pump A to stop the loss of coolant. The crew will execute AOP-8 due to the reduction in core flow. The SRO will determine the Technical Specification impact.
Condensate Booster pump A will trip and Condensate Booster pump B will fail to start.
Condensate Booster pump C will also trip after a 3 minute time delay. The crew will enter AOP-1 and insert a manual Reactor scram. On the scram attempt, the RPS pushbuttons and Mode Switch will fail to work. The crew will insert control rods by initiating ARI. The crew will enter EOP-2 and stabilize the plant. HPCI and RCIC will fail to automatically start. With all Condensate Booster and Feedwater pumps unavailable for injection, the crew will likely manually start HPCI and/or RCIC to restore Reactor water level.
RWCU will develop a steam leak. This will cause high area temperatures in the Reactor Building. RWCU will fail to automatically isolate. The crew will be able to close one isolation valve, however the breaker for the other isolation valve will trip open, preventing isolation of the steam leak. The crew will execute EOP-5. As Reactor Building area temperatures approach max safe levels, the crew will attempt to anticipate Emergency Depressurization by rapidly lowering Reactor pressure with Turbine Bypass Valves. The Bypass Opening Jack will fail to open Turbine Bypass Valves. The crew may open Turbine Bypass Valves by adjusting the pressure regulator setpoint, but this will limit how quickly Reactor pressure can be lowered.
Once two max safe temperatures are exceeded, the crew will perform an Emergency Depressurization.
The scenario will be terminated when all control rods are inserted, the Emergency RPV Depressurization is in progress, and Reactor water level is controlled above 0.
Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-3 Op-Test No.: 17-1 Examiners: ___________________________ Operators:
Initial Conditions: The plant is operating at approximately 85% power. SRV A is inoperable. RHR loop A is operating in the Torus Cooling lineup.
Turnover: Secure Torus Cooling per OP-13B. Then, raise Reactor power using control rods and Recirculation flow per the provided RMI.
Event No.
Malf.
No.
Event Type*
Event Description 1
N/A N -
Remote RH35 C -
SRO Torus Cooling and Spray Valve (10MOV-39A) Power Loss ARP 09-3-1-3, Technical Specifications 3
N/A R -
- ATC, SRO Raise Reactor Power with Control Rods and Recirculation Flow OP-65, OP-26 4
RD11 C -
- ATC, SRO Uncoupled Control Rod OP-26, AOP-25 5
ED18:A C - All Electrical Fault on 10500 Bus AOP-18, AOP-59 Technical Specifications 6
MS02:A C - All Steam Leak in Drywell AOP-39, AOP-1, EOP-2, EOP-4 7
ED44 C - All Loss of Offsite Power AOP-72, EOP-2 8
RR15:A M - All Loss of Coolant Accident EOP-2, EOP-4 9
HP02 C - All HPCI Trips EOP-2 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-3 Op-Test No.: 17-1
- 1. Malfunctions after EOP entry (1-2)
Events 7, 9 2
- 2. Abnormal events (2-4)
Events 4, 5, & 6 3
- 3. Major transients (1-2)
Event 8 1
- 4. EOPs entered/requiring substantive actions (1-2)
- 5. EOP contingencies requiring substantive actions (0-2)
EOP-2 Alternate Level Control Leg EOP-2 Emergency Depressurization Leg 2
- 6. Pre-identified critical tasks (2) 2 CRITICAL TASK DESCRIPTIONS:
CT-1: Given a coolant leak inside the Containment, the crew will spray the Drywell, in accordance with EOP-4.
CT-2: Given a coolant leak, a loss of high pressure injection systems, and the inability to restore and maintain Reactor water level above the Top of Active Fuel (TAF), the crew will initiate actions for an Emergency RPV Depressurization, in accordance with EOP-2.
Appendix D Scenario Outline Form ES-D-1 The scenario will begin at approximately 85% power with SRV A inoperable. RHR loop A is operating in the Torus Cooling lineup. The crew will begin by securing the Torus Cooling lineup per OP-13B. As the lineup is being secured, the breaker will trip for Torus Cooling and Spray Valve (10MOV-39A). The SRO will determine the Technical Specification impact of this failure.
The crew will raise Reactor power using a combination of Recirculation flow and control rods.
They will begin by moving control rods 26-43 and 26-11 from position 24 to 48. When control rod 26-11 is at position 48, it will become apparent that the control rod is uncoupled. The crew will execute AOP-25 to re-couple the control rod and continue with the power ascension.
An electrical fault on the 4160 VAC 10500 bus will occur. The crew will execute AOP-18 (Loss of 10500 Bus) and AOP-59 (Loss of RPS Bus A). This will significantly impact the availability of Core Spray and RHR for the remainder of the scenario. The SRO will address Technical Specifications.
After the plant is stabilized, a steam leak inside the Drywell develops. The crew will insert a manual Reactor scram due to rising Drywell pressure.
Approximately 3 minutes after the scram, all 115KV offsite power is lost. This further degrades the availability of equipment. The crew will execute AOP-72. The crew will control Reactor water level with HPCI and/or RCIC due to the loss of all Condensate and Feedwater.
Approximately 7 minutes after the scram, the steam leak will degrade further into a significant loss of coolant accident. Rising inventory losses will require additional injection to the Reactor.
Degrading Containment parameters will required Torus and then Drywell sprays.
Approximately 10 minutes after the scram HPCI will trip. The crew will maximize other injection systems (RCIC, SLC), but will be unable to keep up with lowering Reactor water level. The crew will execute the Alternate Level Control leg of EOP-2 to attempt to maintain Reactor water level above the Top of Active Fuel (zero inches).
Due to lowering Reactor water level and insufficient high pressure injection sources, the crew will perform an Emergency Depressurization to allow low pressure injection sources to restore and\\or maintain Reactor water level >0 inches.
The scenario will be terminated when all control rods are inserted, Emergency Depressurization is in progress, and Reactor water level is being controlled above 0 inches.
Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-4 Op-Test No.: 17-1 Examiners: ___________________________ Operators:
Initial Conditions: The plant is operating at approximately 94% power. SRV A is inoperable.
Turnover: Start RBCLC pump A and then secure RBCLC pump B per OP-40 section G.1. Then, raise Reactor power with Recirculation flow.
Event No.
Malf.
No.
Event Type*
Event Description 1
N/A N -
N/A R -
SL04:A C - SRO SLC Squib Valve Continuity Loss ARP 09-3-3-30, Technical Specifications 4
AD06:C C -
FW05:A FW01:A C - All Feedwater Pump Vibration and Delayed Pump Trip ARP 09-6-4-11, AOP-41 6
RX01 C - All Fuel Failure AOP-3, EOP-5, AOP-1, EOP-2 7
MS05 MS08B Remote M - All Main Steam Leak in Turbine Building; One Main Steam Line Fails to Isolate; Turbine Building Ventilation Fails to Isolate AOP-40, EOP-2, EOP-6 8
RP12 I - BOP, SRO MSIVs Fail to Automatically Isolate AOP-40, EOP-2, EOP-6 9
Overrides I - ATC, SRO Master Feedwater Level Controller Fails Low AOP-1, EOP-2 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-4 Op-Test No.: 17-1
- 1. Malfunctions after EOP entry (1-2)
Events 8 & 9 2
- 2. Abnormal events (2-4)
Events 4, 5, 6 3
- 3. Major transients (1-2)
Event 7 1
- 4. EOPs entered/requiring substantive actions (1-2)
- 5. Entry into a contingency EOP with substantive actions (1 per scenario set)
EOP-2 Emergency Depressurization Leg 1
- 6. Pre-identified critical tasks (2) 2 CRITICAL TASK DESCRIPTIONS:
CT-1: Given a fuel failure, the crew will scram the Reactor, in accordance with AOP-3.
CT-2: Given an un-isolable primary system discharging outside of primary and secondary containments and off-site radioactivity release rates approaching the General Emergency level, the crew will perform an emergency RPV depressurization, in accordance with EOP-6.
Appendix D Scenario Outline Form ES-D-1 The scenario will begin at approximately 94% power with SRV A inoperable. The crew will begin the shift by swapping RBCLC pumps per OP-40.
After the RBCLC pumps are swapped, the crew will raise Reactor power using Recirculation flow.
Once Reactor power has been raised, Standby Liquid Control Squib valve A will lose continuity in its initiation circuit. The SRO will address Technical Specifications (TS).
After the TS LCO has been determined, Safety Relief valve C will inadvertently open. The crew will execute AOP-36 (Stuck Open Relief Valve). Actions taken will be successful in closing the SRV and once again the SRO will address TSs.
Feedwater Pump A will then experience high pump vibrations. Efforts to mitigate the pump vibrations will be unsuccessful and the Feedwater pump will trip.
The Feedwater transient will result in fuel clad damage and radiation levels in the Turbine Building will begin to rise. The crew will enter AOP-3 (High Activity in Reactor Coolant or Off-gas) and attempt to minimize the rise in radiation levels.
The SRO will determine a Reactor scram is warranted and direct a manual scram to be inserted. EOP-2 (RPV Control) and EOP-6 (Radioactive Release Control) will be entered.
Turbine Building ventilation radiation levels will approach the General Emergency level, therefore the crew will perform an Emergency RPV Depressurization.
The scenario will be terminated when all control rods are inserted, an Emergency RPV Depressurization is in progress, and RPV level is controlled above zero inches.
ES-401 Written Examination Outline Form ES-401-1 Facility:
JAF 17-1 NRC Date of Exam:
May 2018 Tier Group RO K/A Category Points SRO-Only Points K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G
Total A2 G*
Total
- 1.
Emergency Plant Evolutions 1
3 3
3 4
4 3
20 3
4 7
2 1
2 1
1 1
1 7
1 2
3 Tier Totals 4
5 4
5 5
4 27 4
6 10
- 2.
Plant Systems 1
2 2
2 3
2 3
3 2
2 3
2 26 2
3 5
2 1
0 2
1 2
1 1
1 1
1 1
12 0
2 1
3 Tier Totals 3
2 4
4 4
4 4
3 3
4 3
38 4
4 8
- 3. Generic Knowledge & Abilities Categories 1
2 3
4 10 1
2 3
4 7
2 3
2 3
1 2
2 2
Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply.) Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Q#
295028 High Drywell Temperature / 5 X
EA2.05 - Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE:
Torus/suppression chamber pressure:
Plant-Specific 3.8 76 295019 Partial or Complete Loss of Instrument Air / 8 X
AA2.01 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Instrument air system pressure 3.6 77 295005 Main Turbine Generator Trip / 3 X
AA2.05 - Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP:
Reactor power 3.9 78 295038 High Off-site Release Rate / 9 X
2.4.41 - Emergency Procedures / Plan:
Knowledge of the emergency action level thresholds and classifications.
4.6 79 295003 Partial or Complete Loss of AC Power / 6 X
2.4.21 - Emergency Procedures / Plan:
Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
4.6 80 295004 Partial or Complete Loss of DC Power / 6 X
2.2.36 - Equipment Control: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
4.2 81 295031 Reactor Low Water Level / 2 X
2.4.6 - Emergency Procedures / Plan:
Knowledge of EOP mitigation strategies.
4.7 82 700000 Generator Voltage and Electric Grid Disturbances X
AK1.03 - Knowledge of the operational implications of the following concepts as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Under-excitation 3.3 39 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X
EK1.06 - Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Cooldown effects on reactor power 4.0 40 295028 High Drywell Temperature / 5 X
EK1.02 - Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Equipment environmental qualification 2.9 41 295006 SCRAM / 1 X
AK2.03 - Knowledge of the interrelations between SCRAM and the following:
CRD hydraulic system 3.7 42 295024 High Drywell Pressure /
5 X
EK2.15 - Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: Containment spray logic: Plant-Specific 3.8 43 295038 High Off-site Release Rate / 9 X
EK2.06 - Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Process liquid radiation monitoring system 3.4 44
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Q#
295026 Suppression Pool High Water Temperature / 5 X
EK3.01 - Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Emergency/normal depressurization 3.8 45 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X
AK3.05 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:
Reduced loop operating requirements:
Plant-Specific 3.2 46 295030 Low Suppression Pool Water Level / 5 X
EK3.02 - Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCI operation: Plant-Specific 3.5 47 295019 Partial or Complete Loss of Instrument Air / 8 X
AA1.03 - Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Instrument air compressor power supplies 3.0 48 295018 Partial or Complete Loss of CCW / 8 X
AA1.03 - Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:
Affected systems so as to isolate damaged portions 3.3 49 600000 Plant Fire On-site / 8 X
AA1.05 - Ability to operate and / or monitor the following as they apply to PLANT FIRE ON SITE: Plant and control room ventilation systems 3.0 50 295005 Main Turbine Generator Trip / 3 X
AA2.01 - Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP:
Turbine speed 2.6 51 295004 Partial or Complete Loss of DC Power / 6 X
AA2.04 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: System lineups 3.2 52 295003 Partial or Complete Loss of AC Power / 6 X
AA2.02 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Reactor power, pressure, and level 4.2 53 295016 Control Room Abandonment / 7 X
2.1.23 - Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.
4.3 54 295021 Loss of Shutdown Cooling / 4 X
2.2.37 - Equipment Control: Ability to determine operability and / or availability of safety related equipment.
3.6 55 295023 Refueling Accidents / 8 X
2.2.22 - Equipment Control: Knowledge of limiting conditions for operations and safety limits.
4.0 56 295025 High Reactor Pressure /
3 X
EA1.02 - Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE:
Reactor/turbine pressure regulating system 3.8 57 295031 Reactor Low Water Level / 2 X
EA2.02 - Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL:
4.0 58
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Q#
Reactor power K/A Category Totals:
3 3
3 4
4/3 3/4 Group Point Total:
20/7
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Q#
295015 Incomplete SCRAM / 1 X
AA2.02 - Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM: Control rod position 4.2 83 295034 Secondary Containment Ventilation High Radiation / 9 X
2.2.25 - Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
4.2 84 295032 High Secondary Containment Area Temperature
/ 5 X
2.4.2 - Emergency Procedures / Plan:
Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.
4.6 85 295013 High Suppression Pool Temperature / 5 X
AK1.03 - Knowledge of the operational implications of the following concepts as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Localized heating 3.0 59 295015 Incomplete SCRAM / 1 X
AK2.05 - Knowledge of the interrelations between INCOMPLETE SCRAM and the following: Rod worth minimizer:
Plant-Specific 2.6 60 295007 High Reactor Pressure
/ 3 X
AK3.02 - Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE: HPCI operation: Plant-Specific 3.7 61 295022 Loss of CRD Pumps / 1 X
AA1.03 - Ability to operate and/or monitor the following as they apply to LOSS OF CRD PUMPS: Recirculation system: Plant-Specific 2.7 62 295020 Inadvertent Containment Isolation / 5 & 7 X
AA2.04 - Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION: Reactor pressure 3.9 63 295012 High Drywell Temperature / 5 X
2.4.1 - Emergency Procedures / Plan:
Knowledge of EOP entry conditions and immediate action steps.
4.6 64 295014 Inadvertent Reactivity Addition / 1 X
AK2.06 - Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following: Moderator temperature 3.4 65 K/A Category Totals:
1 2
1 1
1/1 1/2 Group Point Total:
7/3
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A2 A
3 A
4 G
Imp.
Q#
259002 Reactor Water Level Control X
A2.03 - Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of reactor water level input 3.7 86 218000 ADS X
A2.05 - Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of A.C. or D.C.
power to ADS valves 3.6 87 400000 Component Cooling Water X
2.1.25 - Conduct of Operations:
Ability to interpret reference materials, such as graphs, curves, tables, etc.
4.2 88 262001 AC Electrical Distribution X
2.2.37 - Equipment Control:
Ability to determine operability and/or availability of safety related equipment.
4.6 89 203000 RHR/LPCI: Injection Mode X
2.2.40 - Equipment Control:
Ability to apply technical specifications for a system.
4.7 90 259002 Reactor Water Level Control X
K1.05 - Knowledge of the physical connections and/or cause-effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: Reactor feedwater system 3.6 1
215003 IRM X
K1.02 - Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM)
SYSTEM and the following:
Reactor manual control 3.6 2
218000 ADS X
A4.01 - Ability to manually operate and/or monitor in the control room: ADS valves 4.4 3
263000 DC Electrical Distribution X
K2.01 - Knowledge of electrical power supplies to the following:
Major D.C. loads 3.1 4
264000 EDGs X
K3.02 - Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: A.C. electrical distribution 3.9 5
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A2 A
3 A
4 G
Imp.
Q#
215004 Source Range Monitor X
K3.01 - Knowledge of the effect that a loss or malfunction of the SOURCE RANGE MONITOR (SRM) SYSTEM will have on following: RPS 3.4 6
400000 Component Cooling Water X
K4.01 - Knowledge of CCWS design feature(s) and or interlocks which provide for the following: Automatic start of standby pump 3.4 7
K4.02 - Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following: Reactor SCRAM signals 4.1 8
211000 SLC X
K5.04 - Knowledge of the operational implications of the following concepts as they apply to STANDBY LIQUID CONTROL SYSTEM: Explosive valve operation 3.1 9
300000 Instrument Air X
K5.13 - Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: Filters 2.9 10 262002 UPS (AC/DC)
X K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.): A.C.
electrical power 2.7 11 217000 RCIC X
K6.03 - Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC):
Suppression pool water supply 3.5 12 206000 HPCI X
A1.05 - Ability to predict and/or monitor changes in parameters associated with operating the HIGH PRESSURE COOLANT INJECTION SYSTEM controls including: Suppression pool temperature: BWR-2,3,4 4.1 13 205000 Shutdown Cooling X
A1.02 - Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including: SDC/RHR pump flow 3.3 14
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A2 A
3 A
4 G
Imp.
Q#
262001 AC Electrical Distribution X
A2.01 - Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Turbine/generator trip 3.4 15 203000 RHR/LPCI: Injection Mode X
A2.04 - Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C. failures 3.5 16 212000 RPS X
A3.08 - Ability to monitor automatic operations of the REACTOR PROTECTION SYSTEM including:
Recirculation pump trip 3.7 17 209001 LPCS X
A3.01 - Ability to monitor automatic operations of the LOW PRESSURE CORE SPRAY SYSTEM including:
Valve operation 3.6 18 261000 SGTS X
A4.03 - Ability to manually operate and/or monitor in the control room: Fan 3.0 19 223002 PCIS/Nuclear Steam Supply Shutoff X
A4.05 - Ability to manually operate and/or monitor in the control room:
SPDS/ERIS/CRIDS/GDS:
Plant-Specific 2.5 20 239002 SRVs X
2.4.18 - Emergency Procedures
/ Plan: Knowledge of the specific bases for EOPs.
3.3 21 259002 Reactor Water Level Control X
2.1.7 - Conduct of Operations:
Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.
4.4 22 300000 Instrument Air X
K4.01 - Knowledge of INSTRUMENT AIR SYSTEM design feature(s) and or interlocks which provide for the following: Manual/automatic transfers of control 2.8 23 262001 AC Electrical Distribution X
K6.02 - Knowledge of the effect that a loss or malfunction of the following will have on the A.C.
ELECTRICAL DISTRIBUTION:
Off-site power 3.6 24
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A2 A
3 A
4 G
Imp.
Q#
263000 DC Electrical Distribution X
A1.01 - Ability to predict and/or monitor changes in parameters associated with operating the D.C. ELECTRICAL DISTRIBUTION controls including: Battery charging/discharging rate 2.5 25 215003 IRM X
K2.01 - Knowledge of electrical power supplies to the following:
IRM channels/detectors 2.5 26 K/A Category Totals:
2 2
2 3
2 3
3 2/2 2
3 2/3 Group Point Total:
26/5
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A2 A
3 A
4 G
Imp.
Q 201006 RWM X
A2.01 - Ability to (a) predict the impacts of the following on the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Power supply loss:
P-Spec(Not-BWR6) 2.8 91 272000 Radiation Monitoring System X
2.4.45 - Emergency Procedures
/ Plan: Ability to prioritize and interpret the significance of each annunciator or alarm.
4.3 92 286000 Fire Protection System X
A2.03 - Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C. distribution failure: Plant-Specific 3.0 93 201001 Control Rod Drive Hydraulic System X
K1.06 - Knowledge of the physical connections and/or cause-effect relationships between CONTROL ROD DRIVE HYDRAULIC SYSTEM and the following: Component cooling water systems: Plant-Specific 2.8 27 202002 Recirculation Flow Control X
K5.02 - Knowledge of the operational implications of the following concepts as they apply to RECIRCULATION FLOW CONTROL SYSTEM: Feedback signals 2.6 28 290003 Control Room HVAC X
K3.03 - Knowledge of the effect that a loss or malfunction of the CONTROL ROOM HVAC will have on following: Control room temperature 2.9 29 241000 Reactor/Turbine Pressure Regulator X
K4.03 - Knowledge of REACTOR/TURBINE PRESSURE REGULATING SYSTEM design feature(s) and/or interlocks which provide for the following: Turbine speed control 3.0 30 234000 Fuel Handling Equipment X
K5.01 - Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUIPMENT: Crane/hoist operation 2.9 31
ES-401 Form ES-401-1 JAF 17-1 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A2 A
3 A
4 G
Imp.
Q 290001 Secondary Containment X
K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the SECONDARY CONTAINMENT:
Reactor building ventilation:
Plant-Specific 3.5 32 259001 Reactor Feedwater X
A1.06 - Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including:
Feedwater heater level 2.7 33 268000 Radwaste X
A2.01 - Ability to (a) predict the impacts of the following on the RADWASTE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System rupture 2.9 34 215002 RBM X
A3.03 - Ability to monitor automatic operations of the ROD BLOCK MONITOR SYSTEM including: Alarm and indicating lights: BWR-3,4,5 3.1 35 256000 Reactor Condensate X
A4.01 - Ability to manually operate and/or monitor in the control room: Hotwell condensate / condensate booster pumps 3.3 36 239001 Main and Reheat Steam X
2.4.4 - Emergency Procedures /
Plan: Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
4.5 37 216000 Nuclear Boiler Instrumentation X
K3.24 - Knowledge of the effect that a loss or malfunction of the NUCLEAR BOILER INSTRUMENATION will have on following: Vessel level monitoring 3.9 38 K/A Category Totals:
1 0
2 1
2 1
1 1/2 1
1 1/1 Group Point Total:
12/3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility:
JAF 17-1 NRC Date:
May 2018 Category K/A #
Topic RO SRO-Only IR Q#
IR Q#
- 1.
Conduct of Operations 2.1.39 Knowledge of conservative decision making practices.
4.3 94 2.1.1 Knowledge of conduct of operations requirements.
3.8 66 2.1.30 Ability to locate and operate components, including local controls.
4.4 67 Subtotal 2
1
- 2.
Equipment Control 2.2.6 Knowledge of the process for making changes to procedures.
3.6 95 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
4.6 100 2.2.38 Knowledge of conditions and limitations in the facility license.
3.6 68 2.2.21 Knowledge of pre-and post-maintenance operability requirements.
2.9 69 2.2.13 Knowledge of tagging and clearance procedures.
4.1 75 Subtotal 3
2
- 3.
Radiation Control 2.3.14 Knowledge of radiation or containment hazards that may arise during normal, abnormal, or emergency conditions or activities.
3.8 96 2.3.11 Ability to control radiation releases.
4.3 98 2.3.13 Knowledge of Radiological Safety 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, aligning filters, etc.
3.4 70 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.
3.5 71
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Subtotal 2
2
- 4.
Emergency Procedures /
Plan 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, AOPs and SAMGs.
4.4 97 2.4.42 Knowledge of emergency response facilities.
3.8 99 2.4.26 Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage.
3.1 72 2.4.14 Knowledge of general guidelines for EOP usage.
3.8 73 2.4.3 Ability to identify post-accident instrumentation.
3.7 74 Subtotal 3
2 Tier 3 Point Total 10 7
ES-401 Record of Rejected K/As Form ES-401-4 Tier /
Group Randomly Selected K/A Reason for Rejection The following topics / K/As were excluded from the systematic and random sampling process:
1 / 1 295027 High Containment Temperature This topic applies to plants with Mark III containments only. The facility has a Mark I containment.
1 / 2 295011 High Containment Temperature This topic applies to plants with Mark III containments only. The facility has a Mark I containment.
2 / 1 207000 Isolation (Emergency)
Condenser This system is not installed at the facility.
2 / 1 209002 HPCS This system is not installed at the facility.
2 / 2 201004 RSCS This system is no longer installed at the facility.
2 / 2 201005 RCIS This system is not installed at the facility.
G 2.2.3 Knowledge of the design, procedural, and operational differences between units.
This K/A applies to multi-unit facilities only.
G 2.2.4 Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility.
This K/A applies to multi-unit facilities only.
ES-401 Record of Rejected K/As Form ES-401-4 The following K/As were rejected following the systematic and random sampling process:
2 / 1 Question 2 215003 IRM K1.05 - Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM)
SYSTEM and the following:
Display control system: Plant-Specific The facility does not have a display control system.
Randomly resampled K/A 215003 IRM K1.02 -
Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM)
SYSTEM and the following: Reactor manual control.
2 / 1 Question 3 218000 ADS K2.01 - Knowledge of electrical power supplies to the following:
ADS logic The randomly sampled K/A overlaps with the K/A for Question 87 (valve and logic power are the same).
Randomly resampled K/A 218000 ADS A4.01 -
Ability to manually operate and/or monitor in the control room: ADS valves.
2 / 2 Question 27 201006 RWM K1.06 - Knowledge of the physical connections and/or cause-effect relationships between ROD WORTH MINIMIZER SYSTEM (RWM)
(PLANT SPECIFIC) and the following: Rod sequence control system: P-Spec(Not-BWR6)
The facility no longer has RSCS installed.
Additionally, the randomly sampled system is also used in Question 91. Resampling system for better balance of coverage.
Randomly resampled K/A 201001 Control Rod Drive Hydraulic System K1.06 - Knowledge of the physical connections and/or cause-effect relationships between CONTROL ROD DRIVE HYDRAULIC SYSTEM and the following:
Component cooling water systems: Plant-Specific.
2 / 2 Question 28 202002 Recirculation Flow Control K2.01 - Knowledge of electrical power supplies to the following:
Hydraulic power unit: Plant-Specific The facility does not have hydraulic power units.
Randomly resampled K/A 202002 Recirculation Flow Control K5.02 - Knowledge of the operational implications of the following concepts as they apply to RECIRCULATION FLOW CONTROL SYSTEM: Feedback signals.
ES-401 Record of Rejected K/As Form ES-401-4 2 / 2 Question 35 215002 RBM A3.01 - Ability to monitor automatic operations of the ROD BLOCK MONITOR SYSTEM including: Four rod display: BWR-3,4,5 The relationship between the RBM and the four rod display is too limited to develop an acceptable question.
Randomly resampled K/A 215002 RBM A3.03 -
Ability to monitor automatic operations of the ROD BLOCK MONITOR SYSTEM including:
Alarm and indicating lights: BWR-3,4,5.
2 / 2 Question 36 256000 Reactor Condensate A4.12 - Ability to manually operate and/or monitor in the control room: Feedwater heater level: Plant-Specific A question could not be developed for the randomly sampled K/A without overlapping Question 33.
Randomly resampled K/A 256000 Reactor Condensate A4.01 - Ability to manually operate and/or monitor in the control room: Hotwell condensate / condensate booster pumps.
1 / 1 Question 47 295030 Low Suppression Pool Water Level EK3.05 - Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: Suppression pool make-up operation: Mark-III The facility does not have a Mark-III containment.
Randomly resampled K/A 295030 Low Suppression Pool Water Level EK3.02 -
Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCI operation: Plant-Specific.
1 / 1 Question 54 295016 Control Room Abandonment 2.4.31 - Knowledge of annunciator alarms, indications, or response procedures.
An operationally relevant question could not be developed for the randomly sampled generic K/A due to lack of alarms related to Control Room Abandonment.
Randomly resampled K/A 295016 Control Room Abandonment 2.1.23 - Ability to perform specific system and integrated plant procedures during all modes of plant operation.
ES-401 Record of Rejected K/As Form ES-401-4 1 / 1 Question 57 295025 High Reactor Pressure EA1.08 - Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: RRCS: Plant-Specific The facility does not have RRCS.
Randomly resampled K/A 295025 High Reactor Pressure EA1.02 - Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: Reactor/turbine pressure regulating system.
1 / 2 Question 64 295012 High Drywell Temperature 2.4.11 - Knowledge of abnormal condition procedures.
The randomly sampled generic K/A no longer allowed in Tiers 1 and 2 with NUREG 1021 rev
- 11.
Randomly resampled K/A 295012 High Drywell Temperature 2.4.1 - Knowledge of EOP entry conditions and immediate action steps.
3 Question 66 2.1.21 - Ability to verify the controlled procedure copy.
The randomly sampled K/A is better tested on the operating exam.
Randomly resampled K/A 2.1.1 - Knowledge of conduct of operations requirements.
3 Question 75 2.2.37 - Ability to determine operability and / or availability of safety related equipment.
The randomly sampled K/A is also sampled in Question 55.
Randomly resampled K/A 2.2.13 - Knowledge of tagging and clearance procedures.
1 / 1 Question 79 295038 High Off-site Release Rate 2.4.3 - Ability to identify post-accident instrumentation.
The randomly sampled K/A is also sampled in Question 74.
Randomly resampled K/A 295038 High Off-site Release Rate 2.4.41 - Knowledge of the emergency action level thresholds and classifications.
1 / 1 Question 81 295004 Partial or Complete Loss of DC Power 2.4.1 - Knowledge of EOP entry conditions and immediate action steps.
An acceptable question could not be developed for the randomly sampled K/A at the SRO level.
Randomly resampled K/A 295004 Partial or Complete Loss of DC Power 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
ES-401 Record of Rejected K/As Form ES-401-4 1 / 1 Question 82 295031 Reactor Low Water Level 2.2.39 - Knowledge of less than one hour technical specification action statements for systems.
An acceptable question could not be developed for the randomly sampled K/A at the SRO level.
Randomly resampled K/A 295031 Reactor Low Water Level 2.4.6 - Knowledge of EOP mitigation strategies.
2 / 1 Question 89 262001 AC Electrical Distribution 2.1.7 - Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.
The randomly sampled K/A is also sampled in Question 22 and does not fit well with this system.
Randomly resampled K/A 262001 AC Electrical Distribution 2.2.37 - Ability to determine operability and/or availability of safety related equipment.
2 / 2 Question 92 Control Room HVAC Ability to prioritize and interpret the significance of each annunciator or alarm.
The randomly sampled system is also used in Question 29. Resampling system for better balance of coverage.
Randomly resampled K/A 272000 Radiation Monitoring System 2.4.45 - Ability to prioritize and interpret the significance of each annunciator or alarm.
2 / 2 Question 93 268000 Radwaste A2.02 - Ability to (a) predict the impacts of the following on the RADWASTE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High turbidity water An operationally relevant question could not be developed at the appropriate license level for the randomly sampled K/A due to lack of references for high turbidity water. Additionally, the randomly sampled system is also used in Question 34. Resampling system for better balance of coverage.
Randomly resampled K/A 286000 Fire Protection System A2.03 - Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C.
distribution failure: Plant-Specific.
ES-401 Record of Rejected K/As Form ES-401-4 3
Question 99 2.4.1 - Knowledge of EOP entry conditions and immediate action steps.
An acceptable question could not be developed for the randomly sampled K/A at the SRO level.
Randomly resampled K/A 2.4.42 - Knowledge of emergency response facilities.