ML22175A168
| ML22175A168 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/11/2022 |
| From: | Public Service Enterprise Group |
| To: | Joseph Demarshall Operations Branch I |
| Shared Package | |
| ML21020A061 | List: |
| References | |
| EPID L-2022-OLL-0002 | |
| Download: ML22175A168 (51) | |
Text
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Salem Date of Examination:
2/14/2022 Examination Level: RO SRO Operating Test Number:
20-01 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations R,D,P Determine the maximum reactor vessel vent time in response to head voiding IAW Attachment 1 of 2-EOP-FRCI-3
[2.1.25, RO-3.9; Ability to interpret reference materials, such as graphs, curves, tables, etc.]
Conduct of Operations R,D Determine the amount of time to borate for 3 stuck rods and the final BAST level IAW 2-EOP-TRIP-2.
[2.1.20, RO-4.6; Ability to interpret and execute procedure steps].
Equipment Control R,D Perform manual QPTR calculation surveillance IAW S2.OP-ST.NIS-0002
[2.2.12, RO-3.7; Knowledge of surveillance procedures]
Radiation Control R,N Determine radiological dose and stay times for a Containment Entry.
[2.3.13, RO-3.4; 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.]
Emergency Plan NOT USED 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 Summary:
(RO-A1) Determine the maximum reactor vessel vent time in response to head voiding IAW Attachment 1 of 2-EOP-FRCI-3:
The plant has experienced voiding in the reactor vessel and the crew is responding by implementing 2-EOP-FRCI-3, RESPONSE TO VOIDS IN REACTOR VESSEL. The candidate will be required to perform a calculation using from 2-EOP-FRCI-3 and Figure 1 to determine the maximum vent time from the reactor vessel.
(RO-A2) Determine the amount of time to borate for 3 stuck rods and the final BAST level IAW 2-EOP-TRIP-2:
The candidate is directed to perform a rapid boration for 3 stuck rods per step 4 of 2-EOP-TRIP-2, Reactor Trip Response. The candidate is being asked to determine the amount of time to borate for 3 stuck rods and the final BAST level change following the completion of the boration.
(RO-A3) Perform manual QPTR calculation surveillance IAW S2.OP-ST.NIS-0002:
The candidate will be required to manually perform a QPTR calculation IAW S2.OP-ST.NIS-0002. The candidate will determine that the QPTR surveillance is UNSAT.
(RO-A4) Determine radiological dose and stay times for a Containment Entry:
The candidate will be required to use a radiological survey map and determine the personal exposure for performing a task inside Containment. The candidate will also be required to determine the radiological stay times for Gamma and Neutron sources and select the most limiting stay time. This JPM is considered modified or new because determining neutron dose stay times is different than previously used JPMs.
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Salem Date of Examination:
2/14/2022 Examination Level: RO SRO Operating Test Number:
20-01 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations R,D,P Identify and Isolate Non-Essential Chilled Water Loads IAW S2.OP-SO.CH-0001, Attachment 2
[2.1.7, SRO-4.7: Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.]
Conduct of Operations R,M Determine the amount of the time to borate for 3 stuck control rods and evaluate final BAST levels to determine any applicable TS LCO(s) and action(s).
[2.1.20, SRO-4.6: Ability to interpret and execute procedure steps]
Equipment Control R,N Review and approve a completed Containment Ventilation Valve surveillance.
[2.2.12, SRO-4.1: Knowledge of surveillance procedures]
Radiation Control R,N Determine Personnel Exposure and Authorization for Containment entry at power
[2.3.4, SRO-3.7: Knowledge of radiation exposure limits under normal and emergency conditions]
Emergency Plan R,D Classify an Event and Determine PARs (Time Critical)
[2.4.41, SRO-4.6: Knowledge of the emergency action level thresholds and classifications]
NOTE:
All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 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
SUMMARY
(SRO-A1) Identify and Isolate Non-Essential Chilled Water Loads IAW S2.OP-SO.CH-0001, Attachment 2:
The SRO will be directed to determine the non-essential heat load from Table A and then use Table B to select the heat loads required to be isolated to comply with Tech Spec. The SRO will need to choose the heat loads in accordance with S2.OP-SO.CH-0001 Attachment 2.
(SRO-A2) Determine the amount of the time to borate for 3 stuck control rods and evaluate final BAST levels to determine any applicable TS LCO(s) and action(s).:
The candidate is directed to perform a rapid boration for 3 stuck rods per step 4 of 2-EOP-TRIP-2, Reactor Trip Response. The candidate is being asked to determine the amount of time to borate for 3 stuck rods and the final BAST level change following the completion of the boration. The SRO candidate will evaluate the final BAST level and determine LCO 3.1.2.6(a) is NOT MET, and action is required.
(SRO-A3) Review and approve a completed Containment Ventilation Valve surveillance:
The candidate is required to determine if a completed Containment Ventilation Valve surveillance (S2.OP-ST.CBV-0001) was completed correctly and can be approved. In this case, the SRO will determine that the surveillance is UNSAT and identify that 2VC5 stroke time is in the Required Action Range and must be declared inoperable.
(SRO-A4) Determine Personnel Exposure and Authorization for Containment entry at power:
The candidate will determine the dose exposure for two individuals performing a task inside Containment using a radiological survey map. The SRO will determine if the dose exceeds the Administrative Annual limit (2000 mR).
The SRO candidate will also identify the required authorization for entry inside Containment while at power with a 10% per hour downpower in progress. In this case, the SRO will determine that RP Supervisor authorization is required.
(SRO-A5) Make a PAR recommendation during a GE (Time Critical):
The candidate will be given plant conditions (Loss of ALL AC Power) to assess and classify the emergency as a General Emergency (GE) and make an initial PAR recommendation per EP-SA-325-F4, Attachment 4. This is a Time Critical JPM task.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Salem Date of Examination:
2/14/2022 Exam Level: RO SRO-I SRO-U Operating Test Number:
20-01 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title (K/A, Importance RO/SRO)
Type Code*
Safety Function
- a. Control Rod exercise surveillance (001, A2.11, 4.4/4.7) Alternate Path -
Control rods continue to insert after placing in Auto & Manual requiring Rx Trip.
A,D,P,S 1
- b. Perform manual make up to VCT (004, A4.04, 3.2/3.6)
D,S 2
- c. Isolate ECCS Accumulators in EOP-LOCA-1 (006, A3.01 4.0/3.9) Alternate Path - SJ54 fails to close requiring operator to vent accumulator.
A,D,E,L,S 3
- d. Respond to RCP standpipe low level alarm (003, A1.10, 2.5/2.7)
D,E,S 4-Pri
- e. Perform immediate actions for a loss of a SGFP IAW S2.OP-AB.CN-0001 (059, A4.01 3.1/3.1) Alternate Path - automatic turbine runback fails to actuate and control rods fail to insert in Auto. Operator is required to manually initiate turbine runback and take control rods to manual and insert at 48 spm.
A,E,N,S 4-Sec
- f. Manually actuate Containment Spray (026, A2.03, 4.1/4.4) Alternate Path -
Phase B valves fail to close automatically on a CS signal requiring operators to manually close valves.
A,D,E,EN,P,S 5
- g. Respond to loss of 2A 4KV Vital Bus IAW S2.OP-AB.4KV-0001 (062, A2.04 3.1/3.4) Alternate Path - 21 Charging Pump trips on start requiring operator to start 22 Charging Pump.
A,E,N,S 6
- h. Respond to Fire Alarm IAW S2.OP-AB.FIRE-0001 (086, A4.02, 3.5/3.5)
E,N,S 8
In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i. Transfer PZR Backup Heater bus to Emergency Power Supply IAW S2.OP-SO.PZR-0010 (010, A4.02, 3.6/3.4)
D,E 3
- j. Locally open Reactor Trip, 13 CV pump and 1CV175 breakers IAW S1.OP-AB.CR-0001 (012, A4.06, 4.3/4.3) [Salem Unit 1 ONLY]
D,E,L 7
- k. Perform a radioactive liquid release IAW S2.OP-SO.WL-0001 (068, A4.03 3.9/3.8)
D,R 9
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
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 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
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 JPM Summary:
Sim-a - Control Rod exercise surveillance The operator will be directed to continue performing control rod surveillance testing. The operator will complete rod exercise test for Control Bank D by inserting 15 steps. The operator will verify proper rod insertion and then withdraw rods to original position. When the operator places control rods in Auto, rods will continuously insert. The operator will attempt to place rods in manual to stop rod motion, but rods will continue to insert. The operator will take action to manually trip the reactor.
Sim-b - Perform manual make up to VCT The operator will be directed to perform a manual make-up to raise level in the VCT as a result of a failed level channel.
Sim-c - Isolate ECCS Accumulators The operator will be directed to perform actions in EOP-LOCA-1, Reactor Trip or Safety Injection, to isolate the Si Accumulators. When performing steps to close the SJ54s, one of the SJ54s will not close. The operator will then perform alternate steps to vent the accumulator.
Sim-d - Respond to RCP standpipe low level alarm The operator will respond to an unexpected console alarm for a RCP standpipe level low. The operator will take actions in accordance with the console Alarm Response Procedure to open two valves and start a Primary Water Pump to make up to the standpipe until the low level clears and the standpipe level high comes in. Once the standpipe high level alarm comes in, the operator will take the actions to stop the Primary Water Pump and close the valves.
Sim-e - Perform immediate actions for a loss of a SGFP The operator will take the watch and directed to respond to all alarms and indications. After taking the watch, 21 SGFP will trip and the main turbine runback circuit will fail to actuate and rods will fail to move while in Auto. The operator will perform immediate actions per S2.OP-AB.CN-0001, Main Feedwater/Condensate System Abnormality, and recognize that auto turbine runback has not actuate and manually initiate turbine runback. The operator will next verify control rods in Auto, but soon recognize that control rods are not inserting with a raising Tavg.
The operator will place rods in Manual and insert rods to control Tavg.
Sim-f - Manually actuate Containment Spray The operator will be directed to continue on in EOP-TRIP-1 with a large break LOCA in progress. The operator will recognize that containment pressure has not remained less than 15 psig and that Containment Spray/Phase B failed to auto actuate. The operator will manually actuate CS/Phase B. In addition, the operator will recognize that some Phase B valves failed to reposition and will manually reposition these valves.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Sim-g - Respond to loss of 2A 4KV Vital Bus The operator will take the watch and directed to respond to all alarms and indications. After taking the watch, the 2A 4KV Vital Bus will de-energize on bus differential. The operator will respond by entering S2.OP-AB.4KV-0001, Loss of 2A 4KV Vital Bus, and taking action to place 21 charging pump in service. Once 21 charging pump is started, it will trip. The operator will take the alternate path to start 22 charging pump.
Sim-h - Respond to Fire Alarm IAW S2.OP-AB.FIRE-0001 The operator will respond to a fire alarm from the control in accordance with S2.OP-AB.FIRE-0001. The operator will determine the location of the fire then take the actions of the procedure to make a plant page announcement of the fire and location, initiate Control Ventilation to FIRE INSIDE Mode, and then ensure the PZR PORVs are closed and closing the PZR PORV Block valves.
IP-I - Transfer PZR Backup Heater bus to Emergency Power Supply IAW S2.OP-SO.PZR-0010 The operator will locally transfer the PZR Backup Heater bus to the emergency bus in accordance with S2.OP-SO.PZR-0010.
IP-j - Locally open Reactor Trip, 13 CV pump and 1CV175 breakers IAW S1.OP-AB.CR-0001 The operator will be giving directions to locally open the reactor trip breakers, 13 Charging Pump breaker, and the 1CV175 (Rapid Boration Valve) using the S1.OP-AB.CR-0001 attachment.
IP-k - Perform a radioactive liquid release IAW S2.OP-SO.WL-0001 The operator will perform a radioactive liquid release in accordance with S2.OP-SO.WL-0001.
During the release the 2R18 radiation monitor will go into alarm and the operator will take the action to notify the control room to terminate the release by closing the 2WL51 from the control room.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Salem Date of Examination:
2/14/2022 Exam Level: RO SRO-I SRO-U Operating Test Number:
20-01 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title (K/A, Importance RO/SRO)
Type Code*
Safety Function
- a. Control Rod exercise surveillance (001, A2.11, 4.4/4.7) Alternate Path -
Control rods continue to insert after placing in Auto & Manual requiring Rx Trip.
A,D,P,S 1
- b. Perform manual make up to VCT (004, A4.04, 3.2/3.6)
D,S 2
- c. Isolate ECCS Accumulators in EOP-LOCA-1 (006, A3.01 4.0/3.9) Alternate Path - SJ54 fails to close requiring operator to vent accumulator.
A,D,E,L,S 3
- d. Respond to RCP standpipe low level alarm (003, A1.10, 2.5/2.7)
D,E,S 4-Pri
- e. Perform immediate actions for a loss of a SGFP IAW S2.OP-AB.CN-0001 (059, A4.01 3.1/3.1) Alternate Path - automatic turbine runback fails to actuate and control rods fail to insert in Auto. Operator is required to manually initiate turbine runback and take control rods to manual and insert at 48 spm.
A,E,N,S 4-Sec
- f. Manually actuate Containment Spray (026, A2.03, 4.1/4.4) Alternate Path -
Phase B valves fail to close automatically on a CS signal requiring operators to manually close valves.
A,D,E,EN,P,S 5
- g. Not Used
- h. Respond to Fire Alarm IAW S2.OP-AB.FIRE-0001 (086, A4.02, 3.5/3.5)
E,N,S 8
In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i. Transfer PZR Backup Heater bus to Emergency Power Supply IAW S2.OP-SO.PZR-0010 (010, A4.02, 3.6/3.4)
D,E 3
- j. Locally open Reactor Trip, 13 CV pump and 1CV175 breakers IAW S1.OP-AB.CR-0001 (012, A4.06, 4.3/4.3) [Salem Unit 1 ONLY]
D,E,L 7
- k. Perform a radioactive liquid release IAW S2.OP-SO.WL-0001 (068, A4.03 3.9/3.8)
D,R 9
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
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 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
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Salem Date of Examination:
2/14/2022 Exam Level: RO SRO-I SRO-U Operating Test Number:
20-01 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title (K/A, Importance RO/SRO)
Type Code*
Safety Function
- d. Respond to RCP standpipe low level alarm (003, A1.10, 2.5/2.7)
D,E,S 4-Pri
- e. Perform immediate actions for a loss of a SGFP IAW S2.OP-AB.CN-0001 (059, A4.01 3.1/3.1) Alternate Path - automatic turbine runback fails to actuate and control rods fail to insert in Auto. Operator is required to manually initiate turbine runback and take control rods to manual and insert at 48 spm.
A,E,N,S 4-Sec
- f. Manually actuate Containment Spray (026, A2.03, 4.1/4.4) Alternate Path -
Phase B valves fail to close automatically on a CS signal requiring operators to manually close valves.
A,D,E,EN,P,S 5
In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- j. Locally open Reactor Trip, 13 CV pump and 1CV175 breakers IAW S1.OP-AB.CR-0001 (012, A4.06, 4.3/4.3) [Salem Unit 1 ONLY]
D,E,L 7
- k. Perform a radioactive liquid release IAW S2.OP-SO.WL-0001 (068, A4.03 3.9/3.8)
D,R 9
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
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 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: _____Salem______
Scenario No.:
____ESG-1_____
Op-Test No.: __20-01 NRC__
Examiners:
Operators:
Initial Conditions: IC-240: Unit 2 is at 100% power, EOL; 21 Charging Pump I/S. The following equipment is out of service: 23 Charging Pump and 21 RHR Pump are C/T for maintenance.
Turnover: The crew is directed to reduce power to 89% power at 10% per hour IAW S2.OP-IO.ZZ-0004 using boration, control rods and turbine load control in preparation for Main Turbine Valve testing.
Critical Tasks:
- 1. Manually actuate at least one train of SI before transition out of TRIP-1
- 2. Manually start one low head ECCS pump before transition out of TRIP-1
- 3. Make up to the RWST, minimize RWST outflow, and if RWST Lo-Lo level alarm received stops ECCS pumps prior to cavitation Event No.
Malf. No.
Event Type*
Event Description 1
N/A ATC (R)
BOP (N)
CRS (N)
Load reduction to 89% at 10% per hour IAW IOP-4 2
VC0087C BOP (C)
CRS (C) 24 vacuum pump trips 3
RC0015D ATC (I)
CRS (I, TS) 24 Loop Cold Leg RTD fails high 4
RC0002 ATC (C)
CRS (C,TS)
RCS leak inside containment (20 gpm) 5 RC0002 RC0001A ALL (M)
-RCS leak worsens to 350 gpm/ Large Break LOCA 6
RP0108 ATC (I)
CRS (I)
-Auto Safety Injection fails to actuate (CT-1) 7 RP318A2 ALL (I) 22 RHR Pump fails to start on SEC signal (CT-2) 8 RH0026B ALL (C) 22 RHR Pump trips (CT-3)
ABs IOP-4 AB.COND-1 AB.ROD-3 AB.RC-1 EOPs TRIP-1 LOCA-1 LOCA-5 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Scenario No.: 1 Target Quantitative Attributes per Scenario (See Section D.5.d)
Actual Attributes Event No.
- 1. Total malfunctions (5-8) 7 2-8
- 2. Malfunctions after EOP entry (1-2) 3 6,7,8
- 3. Abnormal events (2-4) 3 2,3,4
- 4. Major transients (1-2) 1 5
- 5. EOPs entered/requiring substantive actions (1-2) 1 5 (LOCA-1)
- 6. Entry into a contingency EOP with substantive actions ( 1 per scenario set) 1 8 (LOCA-5)
- 7. Preidentified critical tasks (2) 3 6,7,8
- 8. Tech Specs exercised ( 2) 2 3,4
Appendix D Scenario Outline Form ES-D-1 Event Summary (Scenario #1):
Event #1: The crew will perform a load reduction to 89% at 10% per hour in preparation for turbine valve testing. After the crew commences the load reduction, Event #2 will be entered.
Event #2: 24 vacuum pump will trip. The crew will enter S2.OP-AB.COND-0001, Loss of Condenser Vacuum. The crew will start all available vacuum pumps. When 22 vacuum pump starts, it will trip. The crew will then start 25 vacuum pump and 25AR25 will not automatically open. The PO will manually open 25AR25 and verify backpressure is recovering. After condenser backpressure recovers, Event #3 will be entered.
Event #3: 24 RC cold leg RTD will fail high. Control rods will continuously insert. The RO will verify no runback in progress and take rods to manual. The CRS will enter S2.OP-AB.ROD-0003, Continuous Rod Motion, and take actions to defeat the channel, place Master Flow Control in manual to restore charging flow, and then restore rods to previous position. The CRS will evaluate Tech Specs and enter TS 3.3.1.1 action 6 and TS 3.3.2.1 action 19. After the crew restores rods to previous position and evaluates Tech Specs, Event #4 will be entered.
Event #4: A 20 gpm RCS leak inside containment will occur. Alarms for 2R11 radiation monitor and lowering PZR level will be evident. The crew will enter S2.OP-AB.RC-0001, Reactor Coolant System Leak, and take actions to swap to a centrifugal charging pump and attempt to stabilize PZR level by raising charging flow and/or minimize letdown flow. The crew will determine RCS leak rate. The CRS will evaluate Tech Specs and enter TS 3.4.7.2.b action
- b.
Events #5 and #6: The RCS leak will worsen to 350 gpm. The RO will recognize rapidly lowering RCS pressure and PZR level. The CRS will take action per S2.OP-AB.RC-0001 to trip the reactor. Immediately following the reactor trip a Large Break LOCA will occur, Automatic SI will fail and the crew will manually actuate Safety Injection (CT#1).
Event #7: During EOP-TRIP-1, Reactor Trip or Safety Injection, implementation, the crew will recognize that 22 RHR pump failed to start. The crew will block and reset 2B SEC and manually start 22 RHR pump (CT#2). The CRS will transition to EOP-LOCA-1, Loss of Reactor Coolant.
Event #8: During EOP-LOCA-1 implementation, 22 RHR will trip resulting in no emergency coolant recirculation capability. The CRS will transition to EOP-LOCA-5, Loss of Emergency Recirculation, and perform actions to make-up to the RWST, stop Containment Spray pumps and reduce ECCS to single train (CT#3). During LOCA-5, the RWST Lo-Lo Level may be reached. If this occurs, then the crew will take the action to stop all ECCS Pumps taking suction from the RWST (CT#4). The scenario can be terminated when CT#3 or CT#4 evaluation is complete.
Appendix D Scenario Outline Form ES-D-1 Critical Tasks CT-1 (CT-2) - Manually actuate at last one train of Safety Injection before transition out of TRIP-1,, Reactor Trip or Safety Injection SAFETY SIGNIFICANCE -- Failure to manually actuate SI under the postulated conditions constitutes misoperation or incorrect crew performance in which the crew does not prevent degraded emergency core cooling system (ECCS)capacity.
In this case, SI can be manually actuated from the control room. Therefore, failure to manually actuate SI also represents a failure by the crew to demonstrate the following abilities:
Effectively direct or manipulate engineered safety feature (ESF) controls that would prevent (degraded emergency core cooling system (ECCS)capacity)
Recognize a failure or an incorrect automatic actuation of an ESF system or component Take one or more actions that would prevent a challenge to plant safety Additionally, under the postulated plant conditions, failure to manually actuate SI (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 SI under the postulated conditions is a violation of the facility license condition.
In the scenario postulated by the plant conditions, failure to manually actuate SI results in the needless continuation of a situation in which there has been no systematic and thorough actuation of even one train of SIS-actuated safeguards. (Some safeguards components such as AFW and feedwater isolation components may be running because of other actuation signals. However, safeguards systems such as ECCS, phase A containment isolation, CCW/SW, and containment fan coolers will not be operating in their safeguards mode.)
Although the completely degraded status is not due to the crews action (was not initiated by operator error), continuation in the completely degraded status is a result of the crews failure to manually actuate SI.
The acceptable results obtained in the FSAR analyses are predicated on the assumption that, at the very least, one train of safeguards actuates. If SI is not actuated, the FSAR assumptions and results are invalid. Because compliance with the assumptions of the FSAR is part of the facility license condition, failure to manually actuate at least one train of SI (under the postulated conditions and when it is possible to do so) constitutes a violation of the license condition.
The following information is quoted from the ERG Executive Volume, Generic Issues section, document on Foldout Page Items:
The SI actuation criteria are only found on the FOLDOUT PAGE for the ES-0.1, ES-0.2, ES-0.3, and ES-0.4 guidelines for both the HP and LP plants. Although the criteria are identical to the ones found in the SI Reinitiation criteria, the actions are different. The operator is instructed to actuate safety injection rather than operate SI pumps as necessary. The criteria selected for SI actuation are either loss of RCS subcooling or the inability to maintain pressurizer level with charging. Each of these limits indicate that control of the plant is lost and that SI actuation is necessary.
Appendix D Scenario Outline Form ES-D-1 Note the last line of the preceding quote. Clearly, if control of the plant is lost, the preferred action is to manually actuate SI, rather than to manually operate individual safeguards components.
Cues:
Indication and/or annunciation that SI is required
- PRZR pressure or SG pressure less than SI actuation setpoint
- Containment pressure greater than SI actuation setpoint
- Subcooled margin less than the foldout page criterion for SI actuation in ES-0.1
- PRZR water level less than the foldout page criterion for SI actuation in ES-0.1 No indication or annunciation that SI is actuated Measurable Performance Standard:
Manually actuate at least on train of SI before entry into any of the following: transition to any LOCA, SGTR, or LOSC series procedures or FRGs Indication and/or annunciation that at least one train of SI is actuated Feedback:
Indication and/or annunciation that at least one train of SI is actuated CT-2 (CT-5) - Manually start at least one low head ECCS pump before transition out of TRIP-1.
SAFETY SIGNIFICANCE -- Failure to manually start at least one low-head ECCS pump under the postulated conditions constitutes misoperation or incorrect crew performance in which the crew does not prevent degraded emergency core cooling system (ECCS) capacity.
In this case, at least one low-head ECCS pump can be manually started from the control room.
Therefore, failure to manually start a low-head ECCS pump also represents a failure by the crew to demonstrate the following abilities:
Effectively direct or manipulate engineered safety feature (ESF) controls that would prevent a significant reduction of safety margin (beyond that irreparably introduced by the scenario)
Recognize a failure or an incorrect automatic actuation of an ESF system or component Additionally, under the postulated plant conditions, failure to manually start a low-head ECCS pump (when it is possible to do so) is a violation of the facility license condition.
The acceptable results obtained in the FSAR analysis of a large-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 one each of the following ECCS pumps: high-head pump, intermediate-head pump, and low-head pump.
Operation of this minimum required complement of ECCS injection pumps is consistent with the FSAR assumption that only minimum safeguards are actuated.
For both the minimum and maximum cases specified in Comment 1 of this critical task worksheet and for all cases in between, failure to perform the critical task means that the plant is needlessly left in an unanalyzed condition. Performance of the critical task would return the
Appendix D Scenario Outline Form ES-D-1 plant to a condition for which analysis shows acceptable results.
Because compliance with the assumptions 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.
Cues:
Indication and/or annunciation that low-head ECCS pumped injection is required
- SI actuation
- RCS pressure below the shutoff head of the low-head ECCS pumps AND Indication and/or annunciation that no low-head ECCS pump is injecting into the core
- Control switch indication that the circuit breakers or contactors for both low-head ECCS pumps are open
- All low-head ECCS pump discharge pressure indicators read zero
- All flow rate indicators for low-head pumped injection read zero Measurable Performance Standard:
Manually start one low head ECCS pump before transition out of TRIP-1.
Control switch indication that the circuit breaker or contactor for at least one low-head ECCS pump is closed Feedback:
Indication and/or annunciation that at least one low-head ECCS pump is injecting Flow rate indication of injection from at least one low-head ECCS pump CT-3 (CT-29) - Make up to the RWST, minimize RWST outflow, and if RWST Lo-Lo level alarm is received stops ECCS pumps prior to cavitation SAFETY SIGNIFICANCE -- Under the postulated plant conditions, failure to establish makeup flow to the RWST and/or to minimize RWST outflow leads to (or accelerates) depletion of RWST inventory to the point at which ECCS pumps taking suction on the RWST must be stopped. Loss of pumped injection (coincident with loss of emergency cooling recirculation) will lead to a severe or an extreme challenge to the core cooling CSF. Failure to perform the critical task causes these challenges to occur needlessly or, at best, prematurely (that is, before they would occur if the critical task is performed).
Thus, failure to perform the critical task under the postulated plant conditions leads to a significant reduction of safety margin beyond that irreparably introduced by the scenario. It also represents a demonstrated inability by the crew to take one or more actions that would prevent a challenge to plant safety.
Cues:
Indication and/or annunciation that SI is required
- RCS pressure
- Containment pressure AND Indication and/or annunciation that emergency cooling recirculation is not established
Appendix D Scenario Outline Form ES-D-1 despite continuing attempts to establish it
- Indication that both containment sump to RHR suction isolation MOVs remain closed despite attempts to open them remotely and locally
- [Indication of insufficient water level in the containment recirculation sump to allow recirculation]11 AND Indication and/or annunciation that RWST inventory is being depleted
- RWST level indication decreasing (See Comment 1.)
AND Procedural cue to make up to the RWST and/or to minimize outflow from the RWST Measurable Performance Standard:
Make up to the RWST, minimize RWST outflow, and if the RWST Lo-Lo alarm is received stop running ECCS pumps prior to cavitation:
Stopping containment Spray Pumps Initiating RWST makeup Reducing SI to one train If RWST Lo-Lo alarm is received then stops running ECCS pumps prior to cavitation Feedback:
Flow rate indication of makeup to the RWST Reduced depletion rate of RWST inventory
Appendix D Scenario Outline Form ES-D-1 Facility: _____Salem______
Scenario No.:
____ESG-3_____
Op-Test No.: __20-01 NRC__
Examiners:
Operators:
Initial Conditions: IC-242: Unit 2 is at 93% power, MOL; 2A EDG is running unloaded for maintenance run. The following equipment is out of service: 21 AFW Pump C/T for oil bubbler replacement and 21 Containment Spray pump for lube.
Turnover: The crew is directed to continue power ascension to 98% power at 10% per hour IAW S2.OP-IO.ZZ-0004 by use of dilution, control rods and turbine load control.
Critical Tasks:
- 1. Manually actuate main steamline isolation before a Red path to either subcriticality or the integrity CFST or transition to LOSC-2
Malf. No.
Event Type*
Event Description 1
N/A ATC (R)
BOP (N)
CRS (N)
Power ascension to 98% at 10% per hour IAW IOP-4 2
RD0316A BOP (C)
CRS (C) 21 CRDM Vent Fan damper fails closed.
3 PR0017A ATC (I)
CRS (I,TS)
PZR Level Controlling Channel fails high.
4 BF0109A ALL (C) 21CN22 Low Pressure FWH Inlet fails closed.
5 EL0161 CRS (TS) 2A EDG emergency trip.
6 RC0006C RC43CX RC43CY ATC (C)
CRS (C) 23 RCP Motor Oil Level low.
7 RP0279A RP0279B RP0073 RP0069 AF0181B AF0353C ALL (M)
-Main Turbine fails to trip by all means from the control room. (CT-1)
-Main Steam Line Isolation fails to Auto actuate.
(Expect Auto Safety Injection - No SGFPs available and No Steam Dumps are available)
-23 AFW Pump fails to Auto start 8
AF0183 ALL (C)
-22 AFW Pump trips during EOP-TRIP-1
-23 AFW Pump trips during EOP-TRIP-1 (CT-2)
(Loss of All AFW flow, CFST Heat Sink Red Path)
ABs AR.ZZ-11 AB.CVC-1 AB.CN-1 AR.ZZ-13 EOPs TRIP-1 FRHS-1 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Scenario No.: 3 Target Quantitative Attributes per Scenario (See Section D.5.d)
Actual Attributes Event No.
- 1. Total malfunctions (5-8) 7 2-8
- 2. Malfunctions after EOP entry (1-2) 1 8
- 3. Abnormal events (2-4) 4 2,3,4,6
- 4. Major transients (1-2) 1 7
- 5. EOPs entered/requiring substantive actions (1-2) 0
(*)
- 6. Entry into a contingency EOP with substantive actions ( 1 per scenario set) 1 8 (FRHS-1)
- 7. Preidentified critical tasks (2) 2 7,8
- 8. Tech Specs exercised ( 2) 2 3,5
(*) NUREG-1021, Appendix D, Section C.2.f, EOPs Used, states Moreover, the primary scram response procedure that serves as the entry point for the EOPs is not counted. An Attribute Value of 0 for Table Item 5 was determined to be acceptable by the Chief Examiner on the basis that (a) Scenario
- 3 is a complex scenario that exercises Contingency EOP Procedure FRHS-1 for the Loss of Secondary Heat Sink, (b) FRHS-1 requires the use of alternate decision paths and prioritization of actions within the EOP to mitigate a CSFST Heat Sink Red Path condition, and (c) FRHS-1 has measurable actions that must be taken by the crew.
Appendix D Scenario Outline Form ES-D-1 Event Summary (Scenario #3):
Event #1: The crew will continue the power ascension to raise reactor power to 98% at 10%
per hour. The reactivity plan calls for performing a dilution first, then using control rods to adjust RCS temperature before raising turbine load. After the crew commences the power ascension, Event #2 will be entered.
Event #2: 21 CRDM vent fan discharge damper will fail closed. The crew will receive console alarm for sequence not complete. The crew will take actions as directed in the Alarm Response Procedure to stop 21 CRDM vent fan and start 23 CRDM vent fan. Once the standby fan is started, Event #3 can be entered.
Event #3: The controlling PZR level channel will fail high. The crew will respond per S2.OP-AB.CVC-0001, Loss of Charging, and take manual control of charging to raise charging flow.
The crew will select an operable control channel. The CRS will evaluate Tech Specs and enter 3.3.1.1 action 6. After Tech Specs has been evaluated, Event #4 will be entered.
Event #4: 21CN22 low pressure feedwater heater inlet valve will fail closed. The crew will respond per S2.OP-AB.CN-0001, Main Feedwater/Condensate System Abnormality, and take actions to bypass the condensate polisher and reduce turbine load to less than 1098 MWe per (about a 2-3% downpower is required). After the crew completed the load reduction, Event #5 will be entered.
Event #5: 2A EDG will emergency trip. The crew will receive console alarm for 2A EDG tripping. No impact to plant operations. The CRS will evaluate Tech Specs and enter TS 3.8.1.1 action b. The CRS will direct another NCO to perform the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> line surveillance. After the CRS evaluates Tech Specs, Event #6 will be entered.
Events #6: 23 RCP motor bearing oil level OHA will actuate and the crew will recognize elevated motor bearing temperatures in excess of S2.OP-AB.RCP-0001, RCP Abnormality, trip criteria of 175 °F, requiring the crew to manually trip the reactor and stop 23 RCP.
Event #7: During immediate actions of EOP-TRIP-1, Reactor Trip or Safety Injection, the RO will recognize that the main turbine failed to Auto trip following the Rx trip (MSLI will fail to automatically actuate). All attempts to trip the turbine from the control room will fail. The crew will take the action to actuate Fast Closure of the MSIVs (CT#1). Due to the turbine failing to trip, a Safety Injection signal is expected which will trip both SGFPs and setup conditions for condensate flow recovery in EOP-FRHS-1, Response to Loss of Secondary Heat Sink.
Following reactor trip, 23 AFW pump will fail to Auto start. The PO will manually start 23 AFW pump.
Events #8: During EOP-TRIP-1, 22 AFW Pump will trip on overcurrent, 23 AFW pump will trip on overspeed, and a CFST Heat Sink Red path will exist. The CRS will transition to EOP-FRHS-1. The crew will take actions to depressurize one SG using a SG Atmospheric Dump valve (MS10) and feed one SG using condensate feed flow. The scenario can be terminated when SG Wide Range level is rising or CETs are lowering (CT#2).
Appendix D Scenario Outline Form ES-D-1 Critical Tasks:
CT-1 (CT-12) - Manually actuate main steamline isolation before a Red path to either subcriticality or the integrity CFST, or transition to LOSC-2, Uncontrolled Depressurization of All Steam Generators.
SAFETY SIGNIFICANCE -- Failure to close the MSIVs under the postulated plant conditions causes challenges to CSFs beyond those irreparably introduced by the postulated conditions.
Additionally, such an omission constitutes a failure by the crew to demonstrate (the ability to) recognize a failure or an incorrect automatic actuation of an ESF system or component, and to take one or more actions that would prevent a challenge to plant safety.
In the typical FSAR, the analysis for a large steamline break assumes steamline isolation within a short time frame, on the order of seconds. The analysis typically assumes a steam system piping failure in which a single SG blows down completely. That is, the analysis assumes a fault that can be isolated from all but one SG.
However, in the plant conditions postulated for this critical task, the break is located downstream of the MSIVs. Thus, closure of all MSIVs would terminate all uncontrolled blowdown. In this case, there is no reason for even a single SG to completely depressurize. If the crew allows all MSIVs to remain open, then all SGs depressurize uncontrollably and unnecessarily.
Uncontrolled depressurization of all SGs causes an excessive rate of RCS cooldown, well beyond the conditions typically analyzed in the FSAR. The excessive cooldown rate creates large thermal stresses in the reactor pressure vessel and causes rapid insertion of a large amount of positive reactivity.
Thus, failure to close the MSIVs under the postulated conditions can result in challenges to the following CSFs:
Integrity Subcriticality Cues:
[Indication that main steamline isolation is required]6 AND Indication that main steamline isolation has not actuated automatically
- MSIVs indicate open
- Indication of uncontrolled depressurization of all SGs Measurable Performance Standard:
Manually actuate main steamline isolation before a Red path to either subcriticality or the integrity CFST, or transition to LOSC-2, Uncontrolled Depressurization of All Steam Generators.
MSIVs undergo fast-closure (see Endnote 1 for an expanded discussion of fast vs. slow MSIV closure). This can be accomplished by the Fast Closure pushbuttons on 2CC2 or using the Loops 21-24 MSLI on 2CC1 Safeguards bezels.
MSIVs indicate closed
Appendix D Scenario Outline Form ES-D-1 Feedback:
Steam flow indication from all SGs decreases to zero All SGs stop depressurizing RCS cooldown stops CT-2 (CT-43) - Establish feed flow to one SG before RCS bleed and feed is required SAFETY SIGNIFICANCE -- Failure to establish feedwater flow to any SG results in the crews having to rely upon the lower-priority action of establishing RCS bleed and feed to minimize core uncovery. This constitutes incorrect performance that fails to prevent degradation of any barrier to fission product release.
The analyses presented in the ERG Background Document for FR-H.1 demonstrate that a complete loss of heat sink occurs when the SG inventories deplete (dry out). Unless some form of SG inventory is restored, the SG dryout deteriorates primary-to-secondary heat transfer, allowing core decay heat to increase the RCS temperature and pressure. The increasing RCS pressure automatically forces the pressurizer PORVs to open, which creates a small-break LOCA and simultaneously degrades the RCS fission-product barrier. As long as the RCS pressure remains high, the flow out the PORVs exceeds the ECCS flow into the RCS, which depletes RCS inventory. Eventually the core starts to uncover, degrading the core cooling CSF.
Once the core is uncovered, fuel temperatures increase rapidly until severe fuel damage occurs, unless some form of core cooling is restored. Fuel over-heating constitutes severe degradation of a fission-product barrier (fuel matrix/clad).
Establishing feedwater flow into the SGs offers the most effective recovery action to restore the heat sink. The introduction of feedwater flow immediately restores SG inventory and re-establishes primary-to-secondary heat transfer, decreasing RCS pressure and cooling the core. The RCS pressure decrease then precludes the opening of the PORVs and degradation of the RCS barrier.
Cues:
Extreme (RED path) challenge to the heat sink CSF AND Indication that RCS pressure remains above the pressure of all SGs AND Indication that RCS temperature is above the temperature for placing the RHR system in service AND Indication and/or annunciation that no AFW flow is available after repeated attempts to establish AND
[Indication that RCS bleed and feed is not required]4 Measurable Performance Standard:
Establish feed flow to one SG before RCS bleed and feed is required.
Appendix D Scenario Outline Form ES-D-1 Feedback:
Indication of feedwater flow into at least one SG Indication of increasing water level in at least one SG
Appendix D Scenario Outline Form ES-D-1 Facility: _____Salem______
Scenario No.:
____ESG-4_____
Op-Test No.: __20-01 NRC__
Examiners:
Operators:
Initial Conditions: IC-243: Unit 2 is at 2% power, BOL; 21 SGFP in service.
Turnover: The crew is directed to continue power ascension to 10% reactor power IAW S2.OP-IO.ZZ-0003 using control rods, steam dumps, and turbine load control.
Critical Tasks:
- 1. Isolate feed and steam flow to ruptured SG before transition to SGTR-3 occurs
- 2. Cooldown RCS to target temperature so that transition from SGTR-1 does not occur Event No.
Malf. No.
Event Type*
Event Description 1
N/A ATC (R)
CRS (N)
BOP(N)
Continue power ascension to 10% IAW IOP-3 and enter MODE 1.
2 PR0018B ATC (C)
CRS (C,TS) 2PR2 PZR PORV leakage.
3 CW0350E BOP (C)
CRS (C)
High DP across 23A CW Traveling Screen.
4 SG0078C ATC (C)
CRS (C,TS) 23 SG Tube Leak (35 gpm).
5 SG0078C ALL (M) 23 SG Tube Rupture (650 gpm).(CT-1 & CT-2) 6 RP318D1 ALL (I) 21 CFCU fails to start in LOW Speed.
7 PR0019B ATC (C)
CRS (C)
PZR Spray Valve 2PS3 fails to close during depressurization.
ABs IOP-3 AB.PZR-1 AR.ZZ-10 AB.CW-1 AB.SG-1 EOPs TRIP-1 SGTR-1 with depressurization and a failed open spray valve (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Scenario No.: 4 Target Quantitative Attributes per Scenario (See Section D.5.d)
Actual Attributes Event No.
- 1. Total malfunctions (5-8) 6 2-7
- 2. Malfunctions after EOP entry (1-2) 2 6,7
- 3. Abnormal events (2-4) 3 2,3,4
- 4. Major transients (1-2) 1 5
- 5. EOPs entered/requiring substantive actions (1-2) 1 5 (SGTR-1)
- 6. Entry into a contingency EOP with substantive actions ( 1 per scenario set) 0 NA
- 7. Preidentified critical tasks (2) 2 5
- 8. Tech Specs exercised ( 2) 2 2,4
Appendix D Scenario Outline Form ES-D-1 Event Summary (Scenario #4):
Event #1: The crew will raise control rods to raise reactor power to 10%. When reactor power is above 5%, the CRS will enter MODE 1. After the crew enters MODE 1, Event #2 will be entered.
Event #2: 2PR2 PZR PORV valve will experience leakage. RCS pressure will lower and spray valves will eventually close. Subsequently, OHA E-28, PZR HTR ON PRESS LO, will actuate to alert the crew of lowering RCS pressure. The crew will assess plant conditions and determine that PZR PORV tailpipe temperature is elevated indicating a leaking PORV. The CRS will enter S2.OP-AB.PZR-0001, Pressurizer Pressure Malfunction, and take actions to isolate and identify that 2PR2 is leaking and close the associated block valve, 2PR7. The CRS will evaluate Tech Specs for a leaking PZR PORV and enter TS 3.4.5 action a. After Tech Specs has been evaluated, Event #3 will be entered.
Event #3: OHA K-1, 21-23 A CW SCRNWSH TRBL, will actuate and indications for high differential pressure across 23A CW pump traveling screen. 23A traveling screen will be running in fast speed. CW operator will report that heavy grassing on the screens. The DP across the traveling screen will continue to rise until exceeding the emergency trip criteria of > 8 feet. The will crew will emergency trip 23A CW and enter S2.OP-AB.CW-0001, CW System Malfunction. After the CW pump is removed from service, Event #4 will be entered.
Event #4: 23 SG will experience a 25-30 gpm tube leak. The crew will receive various RMS alarms (2R53C, 2R15, 2R41D, and 2R19C) and enter S2.OP-AB.SG-0001, SG Tube Leak. The crew will transfer to a centrifugal charging pump and determine leak rate. The crew will be able to stabilize PZR pressure. The CRS will evaluate Tech Spec and enter TS 3.4.7.2.c action a.
After the CRS evaluates Tech Spec, Event #5 will be entered.
Events #5, #6, and #7: The tube leak on 23 SG will worsen to a design basis 650 gpm tube rupture and the crew will manually trip the reactor and actuate SI. The crew will enter EOP-TRIP-1, Reactor Trip of Safety Injection, and perform the following; (1) isolate feed flow to 23 SG (CT#1), and (2) recognize that 21 CFCU failed to start in LOW speed. The crew will block and reset 2A SEC and manually start 21 CFCU in LOW speed. The CRS will transition to EOP-SGTR-1, Steam Generator Tube Rupture, and isolate the steam side of 23 SG (CT#1),
cooldown to target RCS temperature (CT#2), and then depressurize the RCS to stop the primary to secondary leakage. When the crew attempts to stop the depressurization, 2PS3 spray valve will fail to close requiring the crew to stop 21 and 23 RCPs. The scenario can be terminated at this point.
Appendix D Scenario Outline Form ES-D-1 Critical Tasks:
CT-1 (CT-18) - Isolate feed and stem flow to ruptured SG before transition to SGTR-3, SGTR with LOCA - Subcooled Recovery, occurs SAFETY SIGNIFICANCE -- Failure to isolate the ruptured SG causes a loss of differential pressure between the ruptured SG and the intact SGs. The fact that the crew allows the differential pressure to dissipate and, as a result, are then forced to transition to a contingency ERG constitutes an incorrect performance that necessitates the crew taking compensating action that would complicate the event mitigation strategy....
The analyses presented in the ERG Background Document for E-3 demonstrate that a SGTR violates the RCS fission-product barrier because the 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 PORV 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 PORV or safety valve(s) or to SG overfill, which seriously compromises the SG as a fission-product barrier and complicates mitigation. To stop the primary-to-secondary leakage, the crew must intervene to mitigate excessive inventory increase in the ruptured SG.
To mitigate excessive inventory increase, the crew must take the following actions:
Identify and isolate the ruptured SG Cool down to establish RCS subcooling margin Depressurize RCS to restore inventory Terminate SI to stop primary-to-secondary leakage The RCS depressurization decreases the RCS leakage into the SG, which helps to mitigate the excessive increase in SG inventory. The RCS depressurization also helps the ECCS restore RCS inventory, which in turn allows SI termination. SI termination eliminates the remaining cause of leakage from the RCS into the SG, mitigating the increase in SG inventory.
However, the RCS depressurization and SI termination cannot occur until the crew establishes RCS subcooling margin. To establish subcooling margin, the crew must cool down the RCS to a target temperature. But the crew cannot start the RCS cooldown until the ruptured SG is completely isolated. (Isolation means that all steam flow from the SG and all feedwater flow into the SG must be stopped.)
Isolating the ruptured SG maintains a differential pressure between the ruptured SG and the intact SGs. The differential pressure (250 psi) ensures that minimum RCS subcooling remains after RCS depressurization.
Without steam isolation, the ruptured SG pressure decreases to less than 250 psi above the intact SG as the cooldown occurs. When the crew cannot maintain the 250 psi differential, the ERGs require a transition to contingency ERG ECA-3.1. This transition unnecessarily delays the sequence of actions leading to RCS depressurization and SI termination.
For the feedwater, isolation must occur after the ruptured SG level exceeds minimum indication,
Appendix D Scenario Outline Form ES-D-1 delaying isolation until after the SG tubes are covered. The feedwater coverage of the tubes places a water barrier between the tubes and the steam in the upper portion of the SG. Failure to maintain the water barrier allows the SG steam to contact the tubes. When the tube temperature decreases during the subsequent RCS cooldown, the tubes condense the hot steam, decreasing the SG pressure. The decreasing SG pressure decreases the differential pressure between the ruptured SG and the intact SGs to less than 250 psi. This forces the crew to transition to contingency ERG ECA-3.1, which delays RCS depressurization and SI termination.
Any delay in the feedwater isolation allows the ruptured SG level to increase as the feedwater adds additional inventory along with the primary-to-secondary leakage. Too long a delay prevents the crew from depressurizing the RCS and terminating SI before excessive inventory seriously compromises the SG as a fission-product barrier, which complicates mitigation. The delay in feedwater isolation cannot be measured in terms of SG water level. But the delay can be measured in terms of the crews inability to complete the RCS depressurization or SI termination before excessive SG inventory accumulates.
Thus, when the crew fails to isolate steam and feedwater when it is possible to do so (as in the postulated conditions), it constitutes the following:
An incorrect action that necessitates the crew to take compensating actions that would complicate the event mitigation AND A significant reduction of safety margin beyond that irreparably introduced by the scenario Cues:
Indication and/or annunciation of SGTR in one SG
- Increasing SG water level
- Radiation AND Indication and/or annunciation of reactor trip AND Indication and/or annunciation of SI Measurable Performance Standard:
Isolate feed and steam flow to ruptured SG before transition to SGTR-3, SGTR with LOCA -
Subcooled Recovery, occurs
[Main steam isolation valve position lamps indicate closed Main steam isolation bypass valve position lamps indicate closed PORV setpoint adjusted to ERG Footnote O.03 Blowdown isolation valve position lamps indicate closed Steam isolation valve to TDAFW pump position lamps indicate closed AFW valve position lamps and/or indicators indicate closed Feedwater isolation valve position lamps indicate closed]2
Appendix D Scenario Outline Form ES-D-1 Feedback:
Indication of stable or increasing pressure in the ruptured SG Indication of decreasing or zero feedwater flow rate in the ruptured SG CT2 (CT-19) - Cooldown RCS to target temperature so that transition from SGTR-1 does not occur SAFETY SIGNIFICANCE -- Failure to establish and maintain the correct RCS temperature during a SGTR leads to a transition from E-3 to a contingency ERG. This failure constitutes an incorrect performance that necessitates the crew taking compensating action that would complicate the event mitigation strategy....
The analyses presented in the ERG Background Document for E-3 demonstrate that a SGTR violates the RCS fission-product barrier because the 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 PORV or the safety valve(s) open, releasing radioactivity to the environment. If the leakage continues, the SG inventory increase leads to water release through the PORV or safety valve(s) or to SG overfill, which seriously compromises the SG as a fission-product barrier and complicates mitigation. To stop the primary-to-secondary leakage, the crew must intervene to mitigate excessive inventory increase in the ruptured SG.
To mitigate excessive inventory increase, the crew must take the following actions:
Identify and isolate the ruptured SG Cool down to establish RCS subcooling margin Depressurize RCS to restore inventory Terminate SI to stop primary-to-secondary leakage The RCS depressurization decreases the RCS leakage into the SG, which helps to mitigate the excessive increase in SG inventory. The RCS depressurization also helps the ECCS restore RCS inventory, which in turn allows SI termination. SI termination eliminates the remaining cause of leakage from the RCS into the SG, mitigating the increase in SG inventory.
However, the RCS depressurization and SI termination cannot occur until the crew establishes RCS subcooling margin. To establish subcooling margin, the crew must cool down the RCS to a target temperature. 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 ERG ECA-3.1, thereby delaying the RCS depressurization and SI termination. Such a delay allows the excessive inventory increase of the ruptured SG to continue until the SG overpressure components release water or until SG overfill occurs.
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
Appendix D Scenario Outline Form ES-D-1 could force the RCS temperature low enough to challenge the integrity CSF. The crew must then transition to one of the integrity FRGs. The transition also delays RCS depressurization and SI termination.
For plants without the BIT (BAT for LP plants) or with reduced BIT (BAT) boron concentration, late termination of cooldown could force the RCS temperature low enough to challenge the subcriticality CSF. Also, the crews transition delays RCS depressurization and SI termination.
In addition to avoiding challenges to the CSFs during the cooldown, the crew must maintain the RCS temperature high enough to avoid similar challenges.
Thus, when the crew fails to establish and maintain the correct RCS temperature when it is possible to do so (as in the postulated conditions) without transition from E-3, it constitutes the following:
An incorrect action that necessitates the crew to take compensating actions that would complicates the event mitigation AND A significant reduction of safety margin beyond that irreparably introduced by the scenario Cues:
Indication and/or annunciation of SGTR in one SG
- Increasing SG water level
- Radiation AND Indication and/or annunciation of reactor trip AND Indication and/or annunciation of SI AND Indication of ruptured SG pressure [greater than minimum required pressure]4 Measurable Performance Standard:
Cooldown RCS to target temperature so that transition from SGTR-1 does not occur Steam dump valve position lamps and/or indicators indicate closed SG PORV valve position lamps and/or indicators indicate closed Feedback:
Indication of steam flow rate greater than zero Indication of RCS temperature decreasing OR Indication of RCS temperature less than target temperature
ES-401 PWR Examination Outline Form ES-401-2 Facility: Salem Nuclear Power Plant Date of Exam: February 14 - 24, 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
2 3
3 N/A 4
4 N/A 2
18 3
3 6
2 1
2 2
2 1
1 9
2 2
4 Tier Totals 3
5 5
6 5
3 27 5
5 10
- 2.
Plant Systems 1
3 2
4 4
2 2
2 3
2 2
2 28 2
3 5
2 1
0 1
2 1
0 1
1 0
2 1
10 0
1 2
3 Tier Totals 4
2 5
6 3
2 3
4 2
4 3
38 3
5 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
3 2
2 3
2 2
1 2
Note: 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
3.
Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
4.
Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics 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 2
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000007 (EPE 7; BW E02&E10; CE E02)
Reactor Trip, Stabilization, Recovery / 1 02 EK2.02 - Knowledge of the interrelations between a reactor trip and the following: Breakers, relays and disconnects.
(CFR: 41.7 / 45.7) 2.6 1
000008 (APE 8) Pressurizer Vapor Space Accident / 3 02 AK2.02 - Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Sensors and detectors.
(CFR 41.7 / 45.7) 2.7 2
000009 (EPE 9) Small Break LOCA / 3 08 EA1.08 - Ability to operate and monitor the following as they apply to a small break LOCA: Containment isolation system.
(CFR: 41.7 / 45.5 / 45.6) 4.0 3
000011 (EPE 11) Large Break LOCA / 3 01 EK1.01 - Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA: Natural circulation and cooling, including reflux boiling.
(CFR: 41.8 / 41.10 / 45.3) 4.1 4
000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 08 AK2.08 - Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: CCWS (CFR 41.7 / 45.7) 2.6 5
000022 (APE 22) Loss of Reactor Coolant Makeup / 2 04 AA1.04 - Ability to operate and/or monitor the following as they apply to the Loss of Reactor Coolant Makeup: Speed demand controller and running indicators (positive displacement pump).
(CFR 41.7 / 45.5 / 45.6) 3.3 6
000025 (APE 25) Loss of Residual Heat Removal System / 4 02 AK3.02 - Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat Removal System: Isolation of RHR low-pressure piping prior to pressure increase above specified level.
(CFR 41.5,41.10 / 45.6 / 45.13) 3.3 7
000026 (APE 26) Loss of Component Cooling Water / 8 04 AA2.04 - Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The normal values and upper limits for the temperatures of the components cooled by CCW.
(CFR: 43.5 / 45.13) 2.5 8
000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 03 AK3.03 - Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: Actions contained in AOP for PZR PCS malfunction.
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.7 9
000029 (EPE 29) Anticipated Transient Without Scram / 1 04.
06 G2.4.6 - Knowledge of EOP mitigation strategies.
(CFR: 41.10 / 45.3 / 45.13) 3.7 10 000038 (EPE 38) Steam Generator Tube Rupture / 3 33 EA1.33 - Ability to operate and monitor the following as they apply to a SGTR: Use of S/G for natural circulation cooldown.
(CFR: 41.7 / 45.5 / 45.6) 4.4 11 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 01 AK3.01 - Knowledge of the reasons for the following responses as they apply to the Steam Line Rupture:
Operation of steam line isolation valves.
(CFR 41.5,41.10 / 45.6 / 45.13) 4.2 12 000054 (APE 54; CE E06) Loss of Main Feedwater /4 01 AK1.01 - Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): MFW line break depressurizes the S/G (similar to a steam line break).
(CFR 41.8 / 41.10 / 45.3) 4.1 13 000055 (EPE 55) Station Blackout / 6
ES-401 3
Form ES-401-2 000056 (APE 56) Loss of Offsite Power / 6 03 AA2.03 - Ability to determine and interpret the following as they apply to the Loss of Offsite Power:
Operational status of safety injection pump.
(CFR 43.5 / 45.13) 3.8 14 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 01 AA2.01 - Ability to determine and interpret the following as they apply to the Loss of DC Power:
That a loss of dc power has occurred; verification that substitute power sources have come on line.
(CFR: 43.5 / 45.13) 3.7 15 000062 (APE 62) Loss of Nuclear Service Water / 4 02 AA2.02 - Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The cause of possible SWS loss.
(CFR: 43.5 / 45.13) 2.9 16 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 01.
27 G2.1.27 - Knowledge of system purpose and/or function.
(CFR: 41.7) 3.9 17 (W E04) LOCA Outside Containment / 3 01 W E04: EA1.1 - Ability to operate and/or monitor the following as they apply to the (LOCA Outside Containment): Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(CFR: 41.7 / 45.5 / 45.6) 4.0 18 (W E11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals:
2 3
3 4
4 2
Group Point Total:
18
ES-401 4
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 03 AA2.03 - Ability to determine and interpret the following as they apply to the Dropped Control Rod: Dropped rod, using in-core /
ex-core instrumentation, in-core or loop temperature measurements.
(CFR: 43.5 / 45.13) 3.6 19 000005 (APE 5) Inoperable/Stuck Control Rod / 1 02 AK2.02 - Knowledge of the interrelations between the Inoperable / Stuck Control Rod and the following: Breakers, relays, disconnects, and control room switches.
(CFR 41.7 / 45.7) 2.5 20 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 01 AK1.01 - Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions and the following:
PZR reference leak abnormalities.
(CFR 41.8 / 41.10 / 45.3) 2.8 21 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 04.
46 G2.4.46 - Ability to verify that the alarms are consistent with the plant conditions.
(CFR: 41.10 / 43.5 / 45.3 / 45.12) 4.2 22 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 02 AK2.02 - Knowledge of the interrelations between the Accidental Gaseous Radwaste Release and the following:
Auxiliary building ventilation system.
(CFR 41.7 / 45.7) 2.7 23 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 06 EA1.06 - Ability to operate and monitor the following as they apply to a Inadequate Core Cooling: RCPs.
(CFR: 41.7 / 45.5 / 45.6) 3.6 24 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3
ES-401 5
Form ES-401-2 (W E01 & E02) Rediagnosis & SI Termination / 3 02 W E01: EK3.2 - Knowledge of the reasons for the following responses as they apply to the (Reactor Trip or Safety Injection/Rediagnosis): Normal, abnormal, and emergency operating procedures associated with (Reactor Trip or Safety Injection/Rediagnosis).
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.0 25 (W E13) Steam Generator Overpressure / 4 02 W E13: EK3.2 - Knowledge of the reasons for the following responses as they apply to the (Steam Generator Overpressure):
Normal, abnormal and emergency operating procedures associated with (Steam Generator Overpressure).
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 2.9 26 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 01 W E03: EA1.1 - Ability to operate and/or monitor the following as they apply to the (LOCA Cooldown and Depressurization):
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(CFR: 41.7 / 45.5 / 45.6) 4.0 27 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:
1 2
2 2
1 1
Group Point Total:
9
ES-401 6
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 003 (SF4P RCP) Reactor Coolant Pump 07 K4.07 - Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following: Minimizing RCS leakage (mechanical seals).
(CFR: 41.7) 3.2 28 004 (SF1; SF2 CVCS) Chemical and Volume Control 11 01.
32 G2.1.32 - Ability to explain and apply system limits and precautions.
(CFR: 41.10 / 43.2 / 45.12)
K4.11 - Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following: Temperature/pressure control in letdown line: prevent boiling, lifting reliefs, hydraulic shock, piping damage, and burst.
(CFR: 41.7) 3.8 3.1 29 30 005 (SF4P RHR) Residual Heat Removal 01 05 K3.01 - Knowledge of the effect that a loss or malfunction of the RHRS will have on the following: RCS.
(CFR: 41.7 / 45.6)
K5.05 - Knowledge of the operational implications of the following concepts as they apply to the RHRS: Plant response during solid plant pressure change due to the relative incompressibility of water.
(CFR: 41.5 / 45.7) 3.9 2.7 31 32 006 (SF2; SF3 ECCS) Emergency Core Cooling 06 K5.06 - Knowledge of the operational implications of the following concepts as they apply to ECCS: Relationship between ECCS flow and RCS pressure.
(CFR: 41.5 / 45.7) 3.5 33 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 03 A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Overpressurization of the PZR.
(CFR: 41.5 /43.5/ 45.3/45.13) 3.6 34 008 (SF8 CCW) Component Cooling Water 04 02 A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High/low surge tank level.
(CFR: 41.5 /43.5/ 45.3/45.13)
K1.04 - Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: RCS, in order to determine source(s) of RCS leakage into the CCWS.
(CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.2 3.3 35 36 010 (SF3 PZR PCS) Pressurizer Pressure Control 07 A1.07 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including:
RCS pressure.
(CFR: 41.5 / 45.5) 3.7 37
ES-401 7
Form ES-401-2 012 (SF7 RPS) Reactor Protection 02 K3.02 - Knowledge of the effect that a loss or malfunction of the RPS will have on the following: T/G.
(CFR: 41.7 / 45.6) 3.2 38 013 (SF2 ESFAS) Engineered Safety Features Actuation 07 01 K6.01 - Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS: Sensors and detectors.
(CFR: 41.7 / 45.5 to 45.8)
K4.07 - Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: Power supply loss.
(CFR: 41.7) 2.7 3.7 39 40 022 (SF5 CCS) Containment Cooling 04 A4.04 - Ability to manually operate and/or monitor in the control room: Valves in the CCS.
(CFR: 41.7 / 45.5 to 45.8) 3.1 41 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray 01 K2.01 - Knowledge of bus supplies to the following: Containment spray pumps.
(CFR: 41.7) 3.4 42 039 (SF4S MSS) Main and Reheat Steam 07 A4.07 - Ability to manually operate and/or monitor in the control room: Steam dump valves.
(CFR: 41.7 / 45.5 to 45.8) 2.8 43 059 (SF4S MFW) Main Feedwater 19 02 K4.19 - Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Automatic feedwater isolation of MFW.
(CFR: 41.7)
A3.02 - Ability to monitor automatic operation of the MFW, including:
Programmed levels of the S/G.
(CFR: 41.7 / 45.5) 3.2 2.9 44 45 061 (SF4S AFW) Auxiliary/Emergency Feedwater 01 K6.01 - Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Controllers and positioners.
(CFR: 41.7 / 45.7) 2.5 46 062 (SF6 ED AC) AC Electrical Distribution 02 K3.02 - Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following: ED/G.
(CFR: 41.7 / 45.6) 4.1 47 063 (SF6 ED DC) DC Electrical Distribution 02 01.
30 K3.02 - Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following: Components using DC control power.
(CFR: 41.7 / 45.6)
G2.1.30 - Ability to locate and operate components, including local controls.
(CFR: 41.7 / 45.7) 3.5 4.4 48 49 064 (SF6 EDG) Emergency Diesel Generator 08 A1.08 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including: Maintaining minimum load on ED/G (to prevent reverse power).
(CFR: 41.5 / 45.5) 3.1 50 073 (SF7 PRM) Process Radiation Monitoring 02 A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure.
(CFR: 41.5 /43.5/ 45.3/45.13) 2.7 51
ES-401 8
Form ES-401-2 076 (SF4S SW) Service Water 16 K1.16 - Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems: ESF (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.6 52 078 (SF8 IAS) Instrument Air 01 K2.01 - Knowledge of bus power supplies to the following: Instrument air compressor.
(CFR: 41.7) 2.7 53 103 (SF5 CNT) Containment 01 01 A3.01 - Ability to monitor automatic operation of the containment system, including: Containment isolation.
(CFR: 41.7 / 45.5)
K1.01 - Knowledge of the physical connections and/or cause-effect relationships between the containment system and the following systems: CCS.
(CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.9 3.6 54 55 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:
3 2
4 4
2 2
2 3
2 2
2 Group Point Total:
28
ES-401 9
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 001 (SF1 CRDS) Control Rod Drive 04 K5.04 - Knowledge of the following operational implications as they apply to the CRDS: Rod insertion limits.
(CFR: 41.5 / 45.7) 4.3 56 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 07 K4.07 - Knowledge of NIS design feature(s) and/or interlock(s) which provide for the following: Permissives.
(CFR: 41.7) 3.7 57 016 (SF7 NNI) Nonnuclear Instrumentation 12 K3.12 - Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: S/G.
(CFR: 41.7 / 45.6) 3.4 58 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 03 A4.03 - Ability to manually operate and/or monitor in the control room: CIRS fans.
(CFR: 41.7 / 45.5 to 45.8) 3.3 59 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 01 A4.01 - Ability to manually operate and/or monitor in the control room: HRPS controls.
(CFR: 41.7 / 45.5 to 45.8) 4.0 60 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 05 K1.05 - Knowledge of the physical connections and/or cause-effect relationships between the Spent Fuel Pool Cooling System and the following systems: RWST.
(CFR: 41.2 to 41.9 / 45.7 to 45.8) 2.7 61 034 (SF8 FHS) Fuel-Handling Equipment 02 K4.02 - Knowledge of design feature(s) and/or interlock(s) which provide for the following:
Fuel movement.
(CFR: 41.7) 2.5 62 035 (SF4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 01 A1.01 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SDS controls including: T-ave., verification above low/low setpoint.
(CFR: 41.5 / 45.5) 2.9 63 045 (SF4S MTG) Main Turbine Generator 01.
23 G2.1.23 - Ability to perform specific system and integrated plant procedures during all modes of plant operation.
(CFR: 41.10 / 43.5 / 45.2 / 45.6) 4.3 64 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 04 A2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of condensate pumps.
(CFR: 41.5 /43.5/ 45.3/45.13) 2.6 65 068 (SF9 LRS) Liquid Radwaste
ES-401 10 Form ES-401-2 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 (SF8 FPS) Fire Protection 050 (SF9 CRV*) Control Room Ventilation K/A Category Point Totals:
1 0
1 2
1 0
1 1
0 2
1 Group Point Total:
10
ES-401 11 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000007 (EPE 7; BW E02&E10; CE E02)
Reactor Trip, Stabilization, Recovery / 1 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient Without Scram / 1 04.
18 G2.4.18 - Knowledge of the specific bases for EOPs.
(CFR: 41.10 / 43.1 / 45.13) 4.0 76 000038 (EPE 38) Steam Generator Tube Rupture / 3 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 000054 (APE 54; CE E06) Loss of Main Feedwater /4 000055 (EPE 55) Station Blackout / 6 04 EA2.04 - Ability to determine and interpret the following as they apply to a Station Blackout:
Instruments and controls operable with only dc battery power available.
(CFR: 43.5 / 45.13) 4.1 77 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 02.
25 G2.2.25 - knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits (CFR: 41.5 / 41.7 / 43.2) 4.2 78 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8 04.
31 G2.4.31 - Knowledge of annunciator alarms, indications, or response procedures.
(CFR: 41.10 / 45.3) 4.1 79 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 (W E04) LOCA Outside Containment / 3 (W E11) Loss of Emergency Coolant Recirculation / 4 01 W E11: EA2.1 - Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
(CFR: 43.5 / 45.13) 4.2 80
ES-401 12 Form ES-401-2 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 02 W E05: EA2.2 - Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.
(CFR: 43.5 / 45.13) 3.7 81 K/A Category Totals:
0 0
0 0
3 3
Group Point Total:
6
ES-401 13 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 08 AA2.08 - Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:
Intermediate range channel operability.
(CFR: 43.5 / 45.13) 3.4 82 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 04.
41 W E16: G2.4.41 - Knowledge of the emergency action level thresholds and classifications.
(CFR: 41.10 / 43.5 / 45.11) 4.6 83 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4
ES-401 14 Form ES-401-2 (BW E08; W E03) LOCA CooldownDepressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 02 W E10: EA2.2 - Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.
(CFR: 43.5 / 45.13) 3.9 84 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 04.
21 W E08: G2.4.21 - Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
(CFR: 41.7 / 43.5 / 45.12) 4.6 85 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:
0 0
0 0
2 2
Group Point Total:
4
ES-401 15 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 003 (SF4P RCP) Reactor Coolant Pump 02 A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP.
(CFR: 41.5 /43.5/ 45.3/45.13) 3.9 86 004 (SF1; SF2 CVCS) Chemical and Volume Control 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 04.
30 G2.4.30 - Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
(CFR: 41.10 / 43.5 / 45.11) 4.1 87 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)
Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution 063 (SF6 ED DC) DC Electrical Distribution 064 (SF6 EDG) Emergency Diesel Generator 02.
36 G2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
(CFR: 41.10 / 43.2 / 45.13) 4.2 89 073 (SF7 PRM) Process Radiation Monitoring
ES-401 16 Form ES-401-2 076 (SF4S SW) Service Water 01 A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of SWS.
(CFR: 41.5 /43.5/ 45.3/45.13) 3.7 90 078 (SF8 IAS) Instrument Air 02.
44 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 88 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:
0 0
0 0
0 0
0 2
0 0
3 Group Point Total:
5
ES-401 17 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 02.
40 G2.2.40 - Ability to apply Technical Specifications for a system.
(CFR: 41.10 / 43.2 / 43.5 / 45.3) 4.7 91 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 04 A2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Misaligned rod.
(CFR: 41.5 /43.5/ 45.3/45.13) 3.9 92 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF4S MTG) Main Turbine Generator 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 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 (SF8 FPS) Fire Protection 02.
38 G2.2.38 - Knowledge of conditions and limitations in the facility license.
(CFR: 41.7 / 41.10 / 43.1 / 45.13) 4.5 93 050 (SF9 CRV*) Control Room Ventilation K/A Category Point Totals:
0 0
0 0
0 0
0 1
0 0
2 Group Point Total:
3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility: Salem Nuclear Power Plant Date of Exam: February 14 - 24, 2022 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.1 G2.1.1 - Knowledge of conduct of operations requirements.
(CFR: 41.10 / 45.13) 3.8 66 2.1.3 G2.1.3 - Knowledge of shift or short-term relief turnover practices.
(CFR: 41.10 / 45.13) 3.7 67 2.1.26 G2.1.26 - Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).
(CFR: 41.10 / 45.12) 3.4 68 2.1.35 G2.1.35 - Knowledge of fuel-handling responsibilities of SROs.
(CFR: 41.10 / 43.7) 3.9 94 2.1.37 G2.1.37 - Knowledge of procedures, guidelines, or limitations associated with reactivity management.
(CFR: 41.1 / 43.6 / 45.6) 4.6 95 Subtotal 3
2
- 2. Equipment Control 2.2.6 G2.2.6 - Knowledge of the process for making changes to procedures.
(CFR: 41.10 / 43.5 / 45.13) 3.0 69 2.2.13 G2.2.13 - Knowledge of tagging and clearance procedures.
(CFR: 41.10 / 45.13) 4.1 70 2.2.11 G2.2.11 - Knowledge of the process for controlling temporary design changes.
(CFR: 41.10 / 43.3 / 45.13) 3.3 96 2.2.17 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) 3.8 97 Subtotal 2
2
- 3. Radiation Control 2.3.4 G2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.
(CFR: 41.12 / 43.4 / 45.10) 3.2 71 2.3.12 G2.3.12 - Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
(CFR: 41.12 / 45.9 / 45.10) 3.2 72 2.3.6 G2.3.5 - Ability to approve release permits.
(CFR: 41.13 / 43.4 / 45.10) 3.8 98 Subtotal 2
1
- 4. Emergency Procedures/Plan 2.4.17 G2.4.17 - Knowledge of EOP terms and definitions.
(CFR: 41.10 / 45.13) 3.9 73 2.4.39 G2.4.39 - Knowledge of RO responsibilities in emergency plan implementation.
(CFR: 41.10 / 45.11) 3.9 74
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 2.4.43 G2.4.43 - Knowledge of emergency communications systems and techniques.
(CFR: 41.10 / 45.13) 3.2 75 2.4.26 G2.4.26 - Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage.
(CFR: 41.10 / 43.5 / 45.12) 3.6 99 2.4.38 G2.4.38 - Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.
(CFR: 41.10 / 43.5 / 45.11) 4.4 100 Subtotal 3
2 Tier 3 Point Total 10 7