PLA-7642, Draft Written Examination and Operating Test Outlines (Folder 2)
| ML18109A320 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/05/2017 |
| From: | Sivaraman M Susquehanna |
| To: | Peter Presby Operations Branch I |
| Shared Package | |
| ML17138A109 | List: |
| References | |
| CAC U01950, PLA-7642 | |
| Download: ML18109A320 (33) | |
Text
Manu Sivaraman Assistant Operations Manager Mr. Peter Presby USNRC Chief Examiner Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.1431 Fax 570.542.1504 Manu.Sivaraman@TalenEnergy.com U.S. Nuclear Regulatory Commission - Region 1 2100 Renaissance Blvd - Suite 100 King of Prussia, PA 19406-2713 SUSQUEHANNA STEAM ELECTRIC STATION TALEN----~
ENERGY 10 CFR 55.40 LOC 29 NRC INITIAL OPERATING EXAMINATION SUBMITTAL UNIT 1 LICENSE NO. NPF-14 Docket No. 50-387 UNIT 2 LICENSE NO. NPF-22 50-388 PLA-7642 Enclosed are the examination outlines, supporting the Initial License Examination scheduled for the weeks of March 5 and 12, 2018 at Susquehanna Steam Electric Station.
This submittal includes all appropriate Examination Standard forms and outlines in accordance with NUREG-1021, "Operator Licensing Examination Standards," Revision
- 11.
In accordance with NUREG-1021, Revision 11, Section ES-201, "Initial Operator Licensing Examination Process," please ensure that these materials are withheld from public disclosure until after the examinations are complete.
Should you have any questions regarding this submittal, please contact Mr. Jason Jennings, Manager-Nuclear Regulatory Affairs at (570) 542-3155.
This ()0 no new regulatory commitments.
M. Sivaraman Enclosures to this letter contain confidential information to be withheld from public disclosure under 10 CFR 2.390
Enclosures:
- Examination Security Agreements (Form ES-201-3) - Administrative Topics Outline(s) (Form ES-301-1) - Control Room/In-Plant Systems Outline (Form ES-301-2) - BWR Examination Outline (Form ES-401-1) - Generic Knowledge and Abilities Outline (Tier 3) (Form ES-401-3) - Scenario Outlines (Form ES-D-1) - Record of Rejected K/As (Form ES-401-4) - Examination Outline Quality Checklist (Form ES-201-2) - Transient and Event Checklist (Form ES-301-5)
Copy: (without attachments)
NRC Document Control Desk L. Micewski, NRC Senior Resident Inspector T. E Hood, NRC Project Manager USNRC - Region I PLA-7642 Enclosures to this letter contain confidential information to be withheld from public disclosure under 10 CFR2.390
Electronic Copy:
B. Berryman J. R. Goodbred, Jr.
J. R. Jennings D. J. LaMarca N. D. Pagliaro M. Sivaraman C. Breitman A. Rodgers M. Wilcox DCS SSES w/o attachments USNRC - Region I PLA-7642 Enclosures to this letter contain confidential information to be withheld from public disclosure under 10 CFR2.390
ES-201 Examination Outline Quality Checklist Form ES-201-2 Facility: ss~s Date of Examination: ('-1 II /LC-i./
--z_otB Initials Item Task Description a
b*
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- 1.
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- b. Assess whether the outline was systematically and randomly prepared in accordance with
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R Section D.1 of ES-401 or ES-401 N and whether all KIA categories are appropriately sampled.
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C. Assess whether the outline overemphasizes any systems, evolutions, or generic topics.
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- d. Assess whether the justifications for deselected or rejected KIA statements are appropriate.
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- 2.
- a.
Using Form ES-301-5, verify that the proposed scenario sets cover the required number of
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normal evolutions, instrument and component failures, technical specifications, and major Qa ~
s transients.
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- b. Assess whether there are enough scenario sets (and spares) to test the projected number and u
mix of applicants in accordance with the expected crew composition and rotation schedule
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L without compromising exam integrity, and ensure that each applicant can be tested using at
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least one new or significantly modified scenario, that no scenarios are duplicated from the T
applicants' audit test(s), and that scenarios will not be repeated on subsequent days.
0 To the extent possible, assess whether the oulline(s) conforms with the qualitative and R
C.
quantitative criteria specified on Form ES-301-4 and described in Appendix D and in w (jJ vh Section D.5, "Specific Instructions for the 'Simulator Operating Test,"' of ES-301 (including overlap).
- 3.
- a. Verify that the systems walkthrough outline meets the criteria specified on Form ES-301-2:
(1)
The outline(s) contains the required number of control room and in-plant tasks distributed w
among the safety functions as specified on the form.
A (2)
Task repetition from the last two NRG examinations is within the limits specified on the form.
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(3)
No tasks are duplicated from the applicant's audit test(s).
K (4)
The number of new or modified tasks meets or exceeds the minimums specified on the form.
T (5)
The number of alternate-path, low-power, emergency, and radiologically controlled area H
tasks meets the criteria on the form.
R 0
- b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:
u (1)
The tasks are distributed among the topics as specified on the form.
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(2)
At least one task is new or significantly modified.
H (3)
No more than one task is repeated from the last two NRG licensing examinations.
C.
Determine whether there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on subsequent days.
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- 4.
- a.
Assess whether plant-specific priorities (including probabilistic risk assessment and individual w tn /t' plant examination insights) are covered in the appropriate exam sections.
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- b. Assess whether the 10 CFR 55.41, 55.43, and 55.45 sampling is appropriate.
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N Ensure that KIA importance ratings (except for plant-specific priorities) are at least 2.5.
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- d. Check for duplication and overlap among exam sections and the last two NRG exams.
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- e. Check the entire exam for balance of coverage.
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- f.
Assess whether the exam fits the appropriate job level (RO or SRO).
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Printed Name/Signature w~
Date
- a. Author J\\.11,e,.1,1 A,!. L t,_) l~ ~ ')<c l
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- b.
Facility Reviewer(*)
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NRG Supervisor n.._£ lol \\....,,/_,'., /nJ,/ lJ': -;, ~ -
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- Not applicable for NRG-prepared examination outlines.
- The independent NRG reviewer initials items in column "c"; the chief examiner's concurrence is required.
ES-201, Page 30 of 32
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
SSES Units 1 and 2 Date of Examination:
March 2018 Examination Level: RO IZJ SRO D Operating Test Number:
LOC29 NRC Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations Conduct of Operations Equipment Control Radiation Control D,S D,P,R 2016 NRC D,R N,R Implement On-Site Class 1 E Operability Test for Inoperable Diesel Generator (24.S0.1475.202)
S0-024-013, KA2.1.31 (4.6)
Complete Aborted Evolution Log OP-133-001, KA2.1.20 (4.6)
Describe Reactor Protection System Response to APRM Voter #1 Upscale Vote Using Prints M1-C72-22, KIA 2.2.41 (3.5)
Determine Radiological and Heat Stress Requirements - Steam Leak in RCIC Room NDAP-QA-0626, SP-00-305, KIA 2.3.7 (3.5)
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 (S 3 for ROs; s 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (=== 1)
(P)revious 2 exams (S 1, randomly selected)
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
SSES Units 1 and 2 Date of Examination:
March 2018 Examination Level: RO D SRO
~
Operating Test Number:
LOC29 NRC Administrative Topic (see Note)
Type Describe activity to be performed Code*
Implement On-Site Class 1 E Operability Test Conduct of Operations D,S for Inoperable Diesel Generator (24.S0.1475.202)
S0-024-013, KA 2.1.31 (4.3)
Authorize Bypassing Rod Block Monitor Conduct of Operations D,R NDAP-QA-0338, KA 2.1.37 (4.6)
Describe Reactor Protection System Response to APRM Voter #1 Upscale Vote Using Prints Equipment Control D,R M1-C72-22, KIA 2.2.41 (3.9)
Determine Radiological and Heat Stress Radiation Control N,R Requirements - Steam Leak in RCIC Room NDAP-QA-0626, SP-00-305, KIA 2.3.7 (3.6)
D, P, R Classify an Emergency Condition and Complete Emergency Plan 2016 Emergency Notification Report NRC EP-PS-100, EP-RM-004, KA 2.4.41 (4.6)
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 (S 3 for ROs; s 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (:::: 1)
(P)revious 2 exams (S 1, randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
SSES Units 1 and 2 Date of Examination:
March 2018 Exam Level: RO
[XI SRO-I
[XI SRO-U D Operating Test Number:
LOC29 NRC Control Room Systems: 8 for RO, 7 for SRO-I System/JPM Title Type Code*
Safety Function
- a. Drain MSIV Leakage to Main Condenser Post-LOCA D,S 9
KA 239003 A4.01 (3.2/3.2), OP-184-001
D,S 2
KA 259002 A4.01 (3.8/3.6), OP-145-001
KA 295016 AA1.08 (4.0/4.0), ON-CREVAC-101
- d. Start RCIC in Pressure Control Mode; Auto Isolation Signal Fails M,A,L,EN,S 4
KA 217000 A4.04 (3.6/3.6), OP-150-001
KA295032 EA1.03 (3.7/3.7), ES-134-003 EN,S 2016 NRC
- f. Synchronize the Main Generator; Auto Synchronization Fails D,A,S 6
KA 262001 A4.04 (3.6/3.7), OP-198-001
- g. Perform Weekly RPS Surveillance N,A,S 7
KA 212000 A4.02 (3.6/3.7), S0-158-001, ON-CRD-101
KA 400000 A2.01 (3.3/3.4), OP-114-001, G0-100-014 In-Plant Systems: 3 for RO, 3 for SRO-I
- i. Shift CRD Flow Stations from A to B D,R 1
201001 A2.07 (3.2/3.1), OP-155-001
- j. Place RHR in Suppression Pool Cooling at RSDP P, D, E, R 5
KA 219000 A2.13 (3.5/3.7), OP-249-005 2016 NRC
KA 264000 A2.09 (3.7/4.1), OP-024-004 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.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- Type Codes Criteria for R /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank
- 5 9/:5 8/:5 4 (E)mergency or abnormal in-plant
.:1/.:1/.:1 (EN)gineered safety feature
.: 1/.: 1/.: 1 (control room system)
(L)ow-Power/Shutdown
.:1/.:1/.:1 (N)ew or (M)odified from bank including 1 (A)
.: 2/.: 2/.: 1 (P)revious 2 exams
- 5 3/:5 3/:5 2 (randomly selected)
(R)CA
.:1/.:1/.:1 (S)imulator
Appendix D Scenario Outline Form ES-D-1 Facility: SSES Units 1 and 2 Scenario No.:
NRC-2 Op-Test No.: LOC29 Examiners:
Operators:
Initial Conditions: The plant is operating at approximately 100% power. IAC B and SLC pump B are out of service for maintenance.
Turnover: Start TBCCW pump Band secure TBCCW pump A per OP-115-001 section 2.2.
Critical Tasks: See Page 2 Event Malf.
Event Event No.
No.
Type*
Description 1
N/A N-BOP, Swap TBCCW Pumps SRO OP-115-001 C-BOP, Loss of Extraction Steam to Feedwater Heating 2
cmfMV05 SRO HV10242A ON-FWHTG-101, Technical Specifications R-ATC 3
cmITR02_P I-SRO Turbine First Stage Pressure Instrument Failure T14201A AR-Technical Specifications fx1RRPA 1 I-ATC, Recirculation Pump Speed Rises 4
RRPASTD.
OUT SRO ON-PWR-101 C-BOP, Fuel Failure 5
mfRR17900 SRO 3
ON-MSLRAD-101, ON-SCRAM-101, E0-000-102 R-ATC 6
mfMS18300 M-AII Main Steam Leak into Turbine Building 8
E0-000-102, E0-000-105 cmfAV06_H V141F022A 7
(B)(C)(D)
C-AII MSIVs Fail Open cmfAV06_H E0-000-105 V141F028A (B)(C)(D) 8 cmfPM03_1 C-AII Turbine Building HVAC Trips V104A(B)
E0-000-105, E0-000-112
{N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Facility: SSES Units 1 and 2 Scenario No.: NRC-2 Op-Test No.: LOC29
- 1. Malfunctions after EOP entry (1-2) 2 Events 7, 8
- 2. Abnormal events (2-4) 3 Events 2, 4, 5
- 3. Major transients (1-2) 1 Event6
- 4. EOPs entered/requiring substantive actions (1-2) 2 E0-000-102, E0-000-105
- 5. EOP contingencies requiring substantive actions (0-2) 1 E0-000-112
- 6. Preidentfied Critical tasks (> 2) 2 CRITICAL TASK DESCRIPTIONS:
CRITICAL TASK JUSTIFICATION:
CT-1.0: Given a fuel failure causing Main Steam Line radiation levels to rise, Manually scramming the Reactor when manually scram the Reactor when Main Steam Line Hi Hi Rad setpoint is Main Steam Line radiation levels approached or exceeded.
approach or exceed predetermined values is necessary to limit the production and release of fission products outside of the Reactor coolant system and Primary Containment. Continued Reactor operation with fuel damage causing high Main Steam Line radiation levels will result in increased production and release of fission products. Plant release rates will rise, resulting in an elevated dose to the public.
CT-2.0: Given a radiological release, perform a Emergency Depressurization In order to minimize radiation exposure to prior to EPB projected dose/ dose rates reaching General Emergency the public, Emergency Depressurization declaration criteria.
of RPV is required if a primary system is discharging and the radioactivity release rate cannot be controlled below release rate that requires a General Emergency.
Appendix D Scenario Outline Form ES-D-1 SCENARIO
SUMMARY
The crew assumes the shift with the plant operating at approximately 100% power. IAC B and SLC pump B are out of service for maintenance.
The crew will begin by starting TBCCW pump Band securing TBCCW pump A per OP-115-001. During this evolution, extraction steam to Feedwater Heater 5A will isolate. The loss of Feedwater heating will cause Reactor power to rise. The crew will respond per ON-FWHTG-101, Feedwater Heating Off Normal Operation, and multiple other off-normal procedures. The crew will be required to reduce Reactor power to S71 %. They will initiate a Recirculation run back to limiter #2. The crew will then isolate Feedwater Heater string A. The SRO will determine the Technical Specification impact.
Next, a failure of the Turbine First Stage Pressure instrumentation will occur. The SRO will determine the Technical Specification impact.
Recirculation pump A speed will rise over approximately 5 minutes. Reactor power will rise to a maximum of approximately 104% if nothing is done. The crew will enter ON-PWR-101, Reactor Power.
The crew will perform the immediate operator actions to lock Recirculation pump A scoop tube and lower Reactor power to below the license limit. The crew may dispatch an operator to manually control Recirculation pump A scoop tube.
The reactivity excursion will cause fuel damage. Off-gas and Main Steam Line radiation levels will rise.
The crew will execute ON-MSLRAD-101, Rising Offgas MSL Rad Levels. The crew will lower Reactor power in an attempt to reduce radiation levels. The crew will eventually scram the Reactor and attempt to isolate the MSIVs. The MSIVs will stick in mid-position.
After the scram, a Main Steam Line break will develop in the Turbine Building. With the MSIVs stuck mid-position, this is an un-isolable primary system discharging outside of the primary containment. Turbine Building exhaust fan A will trip. Turbine Building exhaust fan B will trip approximately 1 minute after being started. The loss of Turbine Building HVAC will lead to an un-monitored release from the Turbine Building. The crew will execute E0-000-105, Radioactivity Release Control. Off-site dose assessment will report dose rates approaching the General Emergency level. The crew will execute E0-000-112, Emergency Depressurization, and open all ADS valves. The crew will control Reactor injection to restore
/ maintain Reactor water level during and after the emergency depressurization.
The scenario will be terminated when 6 SRVs are open and Reactor water level is being restored to or controlled in the assigned band above -161".
Appendix D Scenario Outline Form ES-D-1 Facility: SSES Units 1 and 2 Scenario No.:
NRC-3 Op-Test No.: LOC29 Examiners:
Operators:
Initial Conditions: The plant is operating at approximately 45-46% power. IAC B and SLC pump B are out of service for maintenance.
Turnover: Insert control rods to lower the rod line below 60% per the Reactivity Manipulation Package, OP-156-001, and G0-100-012. Then, remove Recirculation pump B from service per OP-164-001 section 2.6.
Critical Tasks: See Page 2 Event Malf.
Event Event No.
No.
Type*
Description R-Lower Reactor Power with Control Rod Insertion 1
N/A
N/A
- ATC, SRO OP-164-001, Technical Specifications cmfRD02 Refuel Floor High Exhaust Radiation Monitor Fails Downscale 3
RED121NO I-SRO 15A AR-112-G02, Technical Specifications cmfPM03_1 C-CRD Pump A Trip with One Inoperable CRD Accumulator P132A 4
- BOP, mfRD15501 SRO ON-CRD-101, Technical Specifications 93431 C-Main Generator Auto Voltage Regulator Failure 5
Override
- ATC, aiHS10001 SRO ON-GENGRID-101 C-Main Turbine Lube Oil Controller Fails to Minimum Cooling in Auto 6
cmfCN02 T
- BOP, IC10955 SRO AR-123-H05 mfTU19300 Main Turbine Bearing #4 High Temperature and Vibration 7D 7
C-AII mfTU19300 AR-105-C05, AR-105-E05, ON-SCRAM-101 8D 8
mfRD15501 M-AII Hydraulic ATWS 7
E0-000-102, E0-000-113 9
cmfPM02_1 C-AII SLC Pump Trip P208A E0-000-113 (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Facility: SSES Units 1 and 2 Scenario No.: NRC-3 Op-Test No.: LOC29
- 1. Malfunctions after EOP entry (1-2) 1 Event9
- 2. Abnormal events (2-4) 4 Events 4, 5, 6, 7
- 3. Major transients (1-2) 1 Event 8
- 4. EOPs entered/requiring substantive actions (1-2) 2 E0-000-102, E0-000-103
- 5. EOP contingencies requiring substantive actions (0-2) 1 E0-000-113
- 6. Preidentfied Critical tasks (> 2) 2 CRITICAL TASK DESCRIPTIONS:
CRITICAL TASK JUSTIFICATION:
CT-1.0: Given a failure of the reactor to scram with power >5%, Lower RPV High Reactor power after a scram is level less than -60" but greater than -161" to reduce power IAW E0-000-113, attempted indicates a challenge to Level/Power Control.
nuclear fuel and to plant heat sinks. In the event of a loss of the normal heat sink, this may result in adding heat to the Suppression Pool and challenging the Primary Containment. Lowering Reactor power reduces these challenges.
CT-2.0: Given a failure of the reactor to scram, reduce reactor power by Inserting control rods lowers Reactor inserting control rods or injecting Boron IAW E0-100-113 power, which reduces challenges to the plant during a failure to scram.
Additionally, inserting control rods ultimately provides a long-term, stable core shutdown. Boron injection will lower power rapidly, however, alone may not provide a stable shutdown condition.
Appendix D Scenario Outline Form ES-D-1 SCENARIO
SUMMARY
The crew assumes the shift with the plant operating at approximately 45-46% power. IAC B and SLC pump B are out of service for maintenance.
The crew will begin by lowering Reactor power with control rod insertion. The crew will insert control rods to lower rod line below 60% for securing Recirculation pump B. Then, the crew will secure Recirculation pump B per OP-164-001. The crew will establish single loop operation and the SRO will determine the Technical Specification requirements.
At the end of this evolution, Refuel Floor High exhaust radiation monitor A will fail downscale. The SRO will determine the Technical Specification impact.
Next, CRD pump A will trip. The crew will respond per ON-CRD-101 by placing the CRD flow control valve in manual and fully closing it. Then the crew will start CRD pump B, open the CRD flow control valve, and place the valve back in automatic. One CRD accumulator will alarm with low nitrogen pressure. The low nitrogen pressure condition will continue even after other CRD parameters are restored. The SRO will determine the Technical Specification impact of the inoperable accumulator.
The Main Generator voltage regulator will fail to maximum demand while in automatic. The crew will respond per ON-GENGRID-101. The crew may attempt to fix the automatic voltage regulator demand signal, but will eventually place the manual voltage regulator in service and lower reactive load.
The Main Turbine Lube Oil temperature controller will fail to minimum cooling while in automatic. The crew will respond by placing the temperature controller in manual and lowering oil temperature. Main Turbine bearing #4 temperature and vibration will continue to rise even after cooling is restored, indicating bearing damage from the initial temperature transient. Damage to Main Turbine seals will result in rising Main Condenser air in-leakage. Bearing temperature will eventually require the crew to insert a manual Reactor scram.
A hydraulic failure to scram will occur. The crew will execute E0-000-113, Level/Power Control, to control Reactor power, level, and pressure. The crew will attempt to inject SLC, but SLC pump A will trip after
-30 seconds and SLC pump B is out of service for maintenance. The crew will lower Reactor water level.
The crew will be able to insert control rods using RMCS and by repeated manual scrams. If all control rods are inserted during the scenario, the crew will restore and maintain Reactor water level to the normal level control band.
The scenario will be terminated when control rod insertion is in progress or when all control rods are inserted and Reactor water level is being restored to or controlled in the assigned band above -161 ".
Appendix D Scenario Outline Form ES-D-1 Facility: SSES Units 1 and 2 Scenario No.:
NRC4 Op-Test No.: LOC29 Examiners:
Operators:
Initial Conditions: The plant is operating at approximately 100% power. IAC B and SLC pump B are out of service for maintenance.
Turnover: Reduce reactor power to 95% using Recirc Flow. Then perform RCIC valve exercising per S0-150-004. The procedure is in progress up to step 5.1.8.
Critical Tasks: See Page 2 Event Malf.
Event Event No.
No.
Type*
Description 1
N/A R-ATC, Lower Reactor Power Using Recirc Flow SRO 2
N/A N-BOP, Perform Quarterly RCIC Valve Exercising SRO S0-150-004 C-BOP, RCIC Turbine Exhaust to Suppression Pool Valve Fails to Re-3 diHS14959 Open A
SRO S0-150-004, Technical Specifications C-BOP, Loss of Power to Instrument Bus 1Y125 4
rfDC102114 SRO ON-YPNL-101, Technical Specifications 5
cmfEB01_1 C-AII Electrical Fault on ESS Bus 1A (1A201)
A201 ON-4KV-101, Technical Specifications 6
mfMS18300 M-AII Steam Leak in Drywell 7
ON-DWLEAK-101, ON-SCRAM-101, E0-000-102, E0-000-103 7
mfHP15201 C-AII HPCI Trips 5
E0-000-102 8
cmfPM03_1 C-AII RHR Pumps B and D Trip P202B(D)
E0-000-103, OP-116-001 cmfAV04 P Failed Open Suppression Chamber to Drywell Vacuum Breaker 9
SV1570481 C-AII (2)
E0-000-103, E0-000-112 (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Facility: SSES Units 1 and 2 Scenario No.: NRC-4 Op-Test No.: LOC29
- 1. Malfunctions after EOP entry (1-2) 3 Events 6, 7, 8
- 2. Abnormal events (2-4) 3 Events 2, 3, 4
- 3. Major transients (1-2) 1 Event 5
- 4. EOPs entered/requiring substantive actions (1-2) 2 E0-000-102, E0-000-103
- 5. EOP contingencies requiring substantive actions (0-2) 1 E0-000-112
- 6. Preidentfied Critical tasks (> 2) 2 CRITICAL TASK DESCRIPTIONS:
CRITICAL TASK JUSTIFICATION:
CT-1.0: Spray the Drywell with RHRSW when Suppression Chamber Initiating Containment Sprays reduces pressure exceeds 13 psig.
Primary Containment pressure. This reduces stresses on the Drywell and Suppression Chamber, assists in avoiding "chugging" that may cause fatigue failure of the LOCA downcomers, and avoids the need for a blowdown.
These benefits reduce challenges to the fuel cladding, the RPV, and the Primary Containment.
CT-2.0: Perform Emergency Depressurization when Suppression Chamber A Slowdown is required to limit further Pressure cannot be maintained below the Pressure Suppression Limit.
release of energy into the Primary Containment and to ensure that the RPV is depressurized while pressure suppression capability is still available.
This protects the integrity of the Primary Containment.
Appendix D Scenario Outline Form ES-D-1 SCENARIO
SUMMARY
The crew assumes the shift with the plant operating at approximately 100% power. IAC B and SLC pump B are out of service for maintenance.
The crew will begin by lowering reactor power using recirc flow, then performing quarterly RCIC valve exercising per S0-150-004. The second valve to be exercised will be the Turbine Exhaust to Suppression Pool Isolation Valve. This valve will close, but fail to re-open. The surveillance will be placed on hold and the SRO will determine the Technical Specification impact.
Next, power will be lost to Instrument Bus 1Y125. This bus supplies power to multiple Reactor and ECCS indicators. The SRO will determine the Technical Specification impact. The crew will respond per ON-YPNL-101, Loss of Instrument Bus, and restore power to the various instruments from the alternate supply.
Then, an electrical fault will cause ESS Bus 1A to de-energize. This results in the loss of Core Spray pump A, RHR pump A, and RHR loop A Containment Spray ability. The crew will respond per ON-4KV-101, Loss of 4KV ESS Bus, and multiple other off-normal procedures. The crew will cross-tie Instrument Air to Containment Instrument Gas, place RPS Bus A on the alternate supply, reset a half scram and half isolation, and start the 1 B CRD pump. The SRO will determine the Technical Specification impact.
A steam leak will develop inside the Primary Containment. The crew will execute ON-SCRAM-101, Reactor Scram, E0-000-102, RPV Control, and E0-000-103, Primary Containment Control. The crew will scram the Reactor, initiate Suppression Chamber spray, and attempt to initiate Drywell spray. HPCI will trip upon start, requiring use of other systems for Reactor water level control.
When Drywell spray is initiated, RHR pumps Band D will trip. RHR loop A is unavailable for Drywell spray due to earlier electrical losses. The crew will then place alternate Containment Spray in service using RHRSW.
Once alternate spray is in service, a Suppression Chamber to Drywell vacuum breaker will stick open and the steam leak will worsen. Containment pressure will rise and the Pressure Suppression Limit will be violated. The crew will execute E0-000-112, Emergency Depressurization, and open 6 ADS valves.
The scenario will be terminated when 6 SRVs are open, RHRSW is spraying the Drywell, and Reactor water level is being restored to or controlled in the assigned band above -161".
ES-401 1
Form ES-401-1 Facility:
Susquehanna Units 1 & 2 Date of Exam:
03/15/2018 Tier Group RO KIA Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
1 3
4 4
3 3
3 20 3
4 7
Emergency and 2
2 1
1 N/A 1
1 N/A 1
7 2
1 3
Abnormal Plant Evolutions Tier Totals 5
5 5
4 4
4 27 5
5 10
- 2.
1 2
3 2
3 3
2 2
2 3
1 3
26 3
2 5
Plant 2
1 0
1 1
2 1
1 1
1 2
1 12 0 I 2 1
3 Systems Tier Totals 3
3 3
4 5
3 3
3 4
3 4
38 5
3 8
- 3. Generic Knowledge and Abilities 1
2 3
4 10 I
1 I
2 I
3 I
4 I
7 I
Categories 2
3 2
3 1
2 2
2 Note: 1.
Ensure that at least two topics from every applicable KIA 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 KIA category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it is replaced by a KIA 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 NRG 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 KIA 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 KIAs having an importance rating (IR) of 2.5 or higher. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
- 7. The generic (G) Kl As in Tiers 1 and 2 shall be selected from Section 2 of the KIA 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 KIAs.
- 8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' I Rs 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 KIA catalog, and enter the KIA numbers, descriptions, !Rs, and point totals (#) on Form ES-401-3. Limit SRO selections to Kl As that are linked to 10 CFR 55.43.
G* Generic KIAs These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the KIA catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan.
ES-401 2
Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO/SRO)
E/APE # / Name/ Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR Q#
295001 (APE 1) Partial or Complete Loss of X
G2.2.22, Knowledge of limiting conditions 4.0 1
Forced Core Flow Circulation/ 1 & 4 for operations and safety limits.
295003 (APE 3) Partial or Complete Loss of X
AK2.04, Knowledge of the interrelations 3.4 2
AC Power/6 between PARTIAL OR COMPLETE LOSS OF A.G. POWER and the following: A.G.
electrical loads 295004 (APE 4) Partial or Complete Loss of X
AK1.05, Knowledge of the operational 3.3 3
DC Power I 6 implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Loss of breaker protection 295005 (APE 5) Main Turbine Generator Trip /
X AK2.04, Knowledge of the interrelations 3.3 4
3 between MAIN TURBINE GENERATOR TRIP and the following: Main generator protection 295006 (APE 6) Scram I 1 X
G2.4.6, Knowledge of EOP mitigation 3.7 5
strategies.
X AA2.04, Ability to determine and/or 4.1 76 interpret the following as they apply to SCRAM: Reactor pressure 295016 (APE 16) Control Room Abandonment X
AA2.03, Ability to determine and/or interpret 4.3 6
17 the following as they apply to CONTROL ROOM ABANDONMENT: Reactor pressure AA2.02, Ability to determine and/or X
interpret the following as 4.3 77 they apply to CONTROL ROOM ABANDONMENT: Reactor water level 295018 (APE 18) Partial or Complete Loss of X
AK1.01, Knowledge of the operational 3.5 7
CCW/8 implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on component/system operations 295019 (APE 19) Partial or Complete Loss of X
AA 1.04, Ability to operate and/or monitor the 3.3 8
Instrument Air/ 8 following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:
Service air isolations valves: Plant-Specific 295021 (APE 21) Loss of Shutdown Cooling/
X AA 1.03, Ability to operate and/or monitor the 3.1 9
4 following as they apply to LOSS OF SHUTDOWN COOLING: Component cooling water systems: Plant-Specific X
G2.4.31, Knowledge of annunciator 4.1 78 alarms, indications, or response procedures.
295023 (APE 23) Refueling Accidents / 8 X
AK2.02, Knowledge of the interrelations 2.9 10 between REFUELING ACCIDENTS and the following: Fuel pool cooling and cleanup system X
AA2.04, Ability to determine and/or 4.1 79 interpret the following as they apply to REFUELING ACCIDENTS: Occurrence of fuel handling accident
ES-401 3
Form ES-401-1 295024 High Drywell Pressure/ 5 X
EK1.01, Knowledge of the operational 4.1 11 implications of the following concepts as they apply to HIGH DRYWELL PRESSURE:
Drvwell inteoritv: Plant-Specific 295025 (EPE 2) High Reactor Pressure/ 3 X
EK3.03, Knowledge of the reasons for the 3.8 12 following responses as they apply to HIGH REACTOR PRESSURE: HPCI operation:
Plant-Specific X
G2.1.7, Ability to evaluate plant 4.7 80 performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
295026 (EPE 3) Suppression Pool High Water X
EK3.02, Knowledge of the reasons for the 3.9 13 Temperature/ 5 following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool coolino 295028 (EPE 5) High Drywell Temperature X
EA2.06, Ability to determine and/or interpret 3.4 14 (Mark I and Mark II only)/ 5 the following as they apply to HIGH DRYWELL TEMPERATURE:
Torus/suppression chamber air space temperature: Plant-Specific 295030 (EPE 7) Low Suppression Pool Water X
EA 1.01, Ability to operate and/or monitor the 3.6 15 Level/ 5 following as they apply to LOW SUPPRESSION POOL WATER LEVEL:
ECCS systems (NPSH considerations):
Plant-Specific X
G2.4.47, Ability to diagnose and 4.2 81 recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
295031 (EPE 8) Reactor Low Water Level/ 2 X
EA2.03, Ability to determine and/or interpret 4.2 16 the following as they apply to REACTOR LOW WATER LEVEL: Reactor pressure 295037 (EPE 14) Scram Condition Present X
EK2.09, Knowledge of the interrelations 4.0 17 and Reactor Power Above APRM Downscale between SCRAM CONDITION PRESENT or Unknown / 1 AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following: Reactor water level X
G2.4.34, Knowledge of RO tasks 4.1 82 performed outside the main control room during an emergency and the resultant operational effects.
295038 (EPE 15) High Offsite Radioactivity X
EK3.03, Knowledge of the reasons for the 3.7 18 Release Rate / 9 following responses as they apply to HIGH OFF-SITE RELEASE RATE: Control room ventilation isolation: Plant-Specific 600000 (APE 24) Plant Fire On Site I 8 X
AK3.04, Knowledge of the reasons for the 2.8 19 following responses as they apply to PLANT FIRE ON SITE: Actions contained in the abnormal procedure for plant fire on site 700000 (APE 25) Generator Voltage and X
G2.4.8, Knowledge of how abnormal 3.8 20 Electric Grid Disturbances / 6 operating procedures are used in conjunction with EOPs.
KIA Category Totals:
3 4
4 3
3/3 3/4 RO/SRO Group Point Total:
20/7
ES-401 4
Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions-Tier 1/Group 2 (RO/SRO)
E/APE #/Name I Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR Q#
295002 (APE 2) Loss of Main Condenser X
AA2.01, Ability to determine and/or 2.9 21 Vacuum/ 3 interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM:
Condenser vacuum/absolute pressure 295007 (APE 7) High Reactor Pressure/ 3 X
AK2.04, Knowledge of the interrelations 3.2 22 between HIGH REACTOR PRESSURE and the following: LPCS 295008 (APE 8) High Reactor Water Level/ 2 X
AA2.01, Ability to determine and/or 3.9 83 interpret the following as they apply to HIGH REACTOR WATER LEVEL:
Reactor water level 295012 (APE 12) High Drywell Temperature/
X AA 1.02, Ability to operate and/or monitor 3.8 23 5
the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell cooling system 295013 (APE 13) High Suppression Pool X
AK1.04, Knowledge of the operational 2.9 24 Temperature. I 5 implications of the following concepts as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Complete condensation 295017 (APE 17) High Offsite Release Rate/ 9 X
AA2.05, Ability to determine and/or 3.8 84 interpret the following as they apply to HIGH OFF-SITE RELEASE RATE:
Meteorological data 295020 (APE 20) Inadvertent Containment X
G2.1.20, Ability to interpret and execute 4.6 85 Isolation / 5 & 7 procedure steps.
295029 (EPE 6) High Suppression Pool Water X
EK3.03, Knowledge of the reasons for the 3.4 25 Level/ 5 following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL:
Reactor SCRAM 295034 (EPE 11) Secondary Containment X
EK1.01, Knowledge of the operational 3.8 26 Ventilation High Radiation / 9 implications of the following concepts as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Personnel protection 500000 (EPE 16) High Containment Hydrogen X
G2.4.18 Knowledge of the specific bases 3.3 27 Concentration I 5 for EOPs KIA Cateoory Point Totals:
2 1
1 1
1/2 1/1 RO/SRO Group Point Total:
7/3
ES-401 5
Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems-Tier 2/Grou :> 1 (RO/SRO)
System # I Name K
K K
K K
K A A
A A
G KIA Topic(s)
IR Q#
1 2
3 4
5 6
1 2
3 4
203000 (SF2, SF4 RHR/LPCI)
X K3.02, Knowledge of the effect that a loss or 3.5 28 RHR/LPCI: Injection Mode malfunction of the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) will have on following: Suppression pool level 205000 (SF4 SCS) Shutdown Cooling X G2.1.27, Knowledge of system purpose 3.9 29 and/or function.
X K3.05, Knowledge of the effect that a loss or 2.6 30 malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: Fuel pool cooling assist: Plant-Specific 206000 (SF2, SF4 HPCIS)
X A 1.05, Ability to predict and/or monitor 4.1 31 High-Pressure Coolant Injection changes in parameters associated with operating the HIGH PRESSURE COOLANT INJECTION SYSTEM controls including:
Suppression pool temperature: BWR-2,3,4 X
K5.08, Knowledge of the operational 3.0 32 implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM: Vacuum breaker operation: BWR-2,3,4 209001 (SF2, SF4 LPCS)
X A3.06, Ability to monitor automatic operations 3.6 33 Low-Pressure Core Spray of the LOW PRESSURE CORE SPRAY SYSTEM including: Lights and alarms X G2.4.50, Ability to verify system alarm 4.0 86 setpoints and operate controls identified in the alarm response manual.
211000 (SF1 SLCS) Standby Liquid X G2.1.28, Knowledge of the purpose and 4.1 34 Control function of major system components and controls.
212000 (SF7 RPS) Reactor X
K4.12, Knowledge of REACTOR 3.9 35 Protection System PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Bypassing of selected SCRAM signals (manually and automatically): Plant-Specific X
K5.02, Knowledge of the operational 3.3 36 implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM:
Specific logic arrangements 215003 (SF7 IRM)
X K6.02, Knowledge of the effect that a loss or 3.6 37 Intermediate-Range Monitor malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM)
SYSTEM: 24/48 volt D.C. power: Plant-Specific
ES-401 6
Form ES-401-1 215004 (SF? SRMS) Source-Range X
A2.02, Ability to (a) predict the impacts of the 3.4 38 Monitor following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and {b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
SRM inop condition 215005 (SF? PRMS) Average Power X
K5.04, Knowledge of the operational 2.9 39 Range Monitor/Local Power Range implications of the following concepts as they Monitor apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: LPRM detector location and core symmetry X
A2.03, Ability to (a) predict the impacts of 3.8 87 the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inoperative trip (all causes) 217000 (SF2, SF4 RCIC) Reactor X
K4.04, Knowledge of REACTOR CORE 3.0 40 Core Isolation Cooling ISOLATION COOLING SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following: Prevents turbine damage: Plant-Specific 218000 (SF3 ADS) Automatic X
A4.08, Ability to manually operate and/or 3.7 41 Depressurization System monitor in the control room: Suppression pool level 223002 (SF5 PCIS) Primary X
A3.01, Ability to monitor automatic operations 3.4 42 Containment Isolation/Nuclear Steam of the PRIMARY CONTAINMENT Supply Shutoff ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including: System indicating lights and alarms X
A2.06, Ability to (a) predict the impacts of 3.2 88 the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Containment instrumentation failures 239002 (SF3 SRV) Safety Relief X
K2.01, Knowledge of electrical power 2.8 43 Valves supplies to the following: SRV solenoids X
K6.02, Knowledge of the effect that a loss or 3.4 44 malfunction of the following will have on the RELIEF/SAFETY VALVES: Air (Nitrogen) supply: Plant-Specific
ES-401 7
Form ES-401-1 259002 (SF2 RWLCS) Reactor Water X
A1.03, Ability to predict and/or monitor 3.8 45 Level Control changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including:
Reactor power X
A2.02, Ability to (a) predict the impacts of 3.4 89 the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of any number of reactor feedwater flow inputs 261000 (SF9 SGTS) Standby Gas X
K1.01, Knowledge of the physical 3.4 46 Treatment connections and/or cause-effect relationships between STANDBY GAS TREATMENT SYSTEM and the following: Reactor building ventilation system 262001 (SF6 AC) AC Electrical X
A2.10, Ability to (a) predict the impacts of the 2.9 47 Distribution following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Exceeding current limitations X G2.2.36, Ability to analyze the effect of 4.2 90 maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
262002 (SF6 UPS) Uninterruptable X
A3.01, Ability to monitor automatic operations 2.8 48 Power Supply (AC/DC) of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including: Transfer from preferred to alternate source X G2.1.23, Ability to perform specific system 4.3 49 and integrated plant procedures during all modes of plant operation.
263000 (SF6 DC) DC Electrical X
K2.01, Knowledge of electrical power 3.1 50 Distribution supplies to the following: Major D.C. loads 264000 (SF6 EGE) Emergency X
K1.04, Knowledge of the physical 3.2 51 Generators (Diesel/Jet) connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEL/JET) and the following: Emergency generator cooling water system 300000 (SF8 IA) Instrument Air X
K4.03, Knowledge of INSTRUMENT AIR 2.8 52 SYSTEM design feature(s) and or interlocks which provide for the following: Securing of IAS upon loss of cooling water 400000 (SF8 CCS) Component X
K2.02, Knowledge of electrical power 2.9 53 Cooling Water supplies to the following: CCW valves KIA Category Point Totals:
2 3
2 3
3 2
2 2/ 3 1 3/ RO/SRO Group Point Total:
26/5 3
2
ES-401 8
Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems-Tier 2/Group 2 (RO/SRO)
System # I Name K K K K K K A A
A A G*
KIA Topic(s)
IR Q#
1 2
3 4
5 6
1 2
3 4
201001 (SF1 CRDH) CRD Hydraulic X
A 1.09, Ability to predict and/or 2.9 54 monitor changes in parameters associated with operating the CONTROL ROD DRIVE HYDRAULIC SYSTEM controls including: CRD drive water flow 201002 (SF1 RMCS) Reactor Manual Control X G2.4.49, Ability to perform without 4.6 55 reference to procedures those actions that require immediate operation of system components and controls.
201003 (SF1 CROM) Control Rod and Drive X
K5.01, Knowledge of the operational 2.6 56 Mechanism implications of the following concepts as they apply to CONTROL ROD AND DRIVE MECHANISM: Hydraulics 202001 (SF1, SF4 RS) Recirculation System X G2.1.7, Ability to evaluate plant 4.7 91 performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
202002 (SF1 RSCTL) Recirculation Flow X
A4.03, Ability to manually operate 3.1 57 Control System and/or monitor in the control room:
LiQhts and alarms.
215001 (SF7 TIP) Traversing In-Core Probe X
K6.04, Knowledge of the effect that 3.1 58 a loss or malfunction of the following will have on the TRAVERSING IN-CORE PROBE: Primary containment isolation system: Mark-1&11 (Not-BWR1) 215002 (SF7 RBMS) Rod Block Monitor X
K5.01, Knowledge of the 2.6 59 operational implications of the following concepts as they apply to ROD BLOCK MONITOR SYSTEM:
Trip reference selection: Plant-Specific 223001 (SF5 PCS) Primary Containment and X
A2.08, Ability to (a) predict the 3.1 60 Auxiliaries impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Compressor trips (loss of air): Plant-Specific 234000 (SF8 FH) Fuel-Handling Equipment X
A4.01, Ability to manually operate 3.7 61 and/or monitor in the control room:
Neutron monitorinQ system 241000 (SF3 RTPRS) Reactor/Turbine X
K3.08, Knowledge of the effect that 3.7 62 Pressure Regulating System a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: Control/governor valves
ES-401 9
Form ES-401-1 259001 (SF2 FWS) Reactor Feedwater X
A3.06, Ability to monitor automatic 3.1 63 System operations of the REACTOR FEEDWATER SYSTEM including:
Pump discharge pressure 271000 (SF9 OG) Offgas X
A2.04, Ability to (a) predict the 4.1 92 impacts of the following on the OFFGAS SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Offgas system high radiation 272000 (SF?, SF9 RMS) Radiation Monitoring X
K1.08, Knowledge of the physical 3.6 64 connections and/or cause-effect relationships between RADIATION MONITORING SYSTEM and the following: Reactor protection system 286000 (SF8 FPS) Fire Protection X
K4.03, Knowledge of FIRE 3.3 65 PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following:
Maintenance of fire header pressure 290001 (SF5 SC) Secondary Containment X
A2.03, Ability to (a) predict the 3.6 93 impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High area radiation
_,_., Point Totals:
1 0
1 1
2 1
1 1/ 1 2
1/ RO/SRO Group Point Total:
2 1
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility: Susquehanna Units 1 & 2 Date of Exam: 03/15/2018 Category KIA#
Topic RO SRO-only IR Q#
IR Q#
2.1.29 Knowledge of how to conduct system lineups, such as 4.1 66 valves, breakers, switches, etc.
- 1. Conduct of 2.1.39 Knowledge of conservative decision making 4.3 94 practices.
Operations 2.1.3 Knowledge of shift or short-term relief turnover 3.7 67 practices Subtotal lff;ti}i:)
2
- *c:;itt.~,
1 2.2.12 Knowledge of surveillance procedures 4.1 95 2.2.17 Knowledge of the process for managing 3.8 96 maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system
- 2. Equipment operator.
Control 2.2.35 Ability to determine Technical Specification Mode of 3.6 68 Operation.
2.2.13 Knowledqe of taaainq and clearance procedures.
4.1 69 2.2.41 Ability to obtain and interpret station electrical and 3.5 70 mechanical drawinqs.
Subtotal 1:;;*'\\l::'::';i:.
3 2
2.3.4 Knowledge of radiation exposure limits under 3.7 97 normal or emerQencv conditions.
2.3.5 Ability to use radiation monitoring systems, such as 2.9 71 fixed radiation monitors and alarms, portable survey instruments, personnel monitorinq equipment, etc.
2.3.13 Knowledge of radiological safety procedures pertaining 3.4 72
- 3. Radiation to licensed operator duties, such as response to Control radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
2.3.14 Knowledge of radiation or contamination hazards 3.8 98 that may arise during normal, abnormal, or emergency conditions or activities.
Subtotal
- i::,t::i***
2 2
2.4.4 Ability to recognize abnormal indications for system 4.5 73 operating parameters that are entry-level conditions for emerqencv and abnormal operatinq procedures.
2.4.30 Knowledge of events related to system 4.1 99 operation/status that must be reported to internal organizations or external agencies, such as the
- 4. Emergency State, the NRC, or the transmission system Procedures/Plan operator.
2.4.19 Knowledge of EOP layout, symbols, and icons.
3.4 74 2.4.40 Knowledge of SRO responsibilities in emergency 4.5 100 plan implementation.
2.4.50 Ability to verify system alarm setpoints and operate 4.2 75 controls identified in the alarm response manual.
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Subtotal 2
Tier 3 Point Total 7
ES-401 1 / 1 1 / 2 2 I 1 2 / 1 2/2 2/2 295027 High Containment Temperature 295011 High Containment Temperature 207000 Isolation (Emergency)
Condenser 209002 HPCS 201004 RSCS 201005 RCIS 239003 MSIV 2 I 2 Leakage Control 1 / 1 Question 6 295016 Control Room Abandonment AA2.07 - Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT:
Suppression chamber pressure Record of Rejected K/As Form ES-401-4 Reason for Rejection This topic applies to plants with Mark Ill containments only.
The facility has a Mark II containment.
This topic applies to plants with Mark Ill containments only.
The facility has a Mark II containment.
This system is not installed at the facility.
This system is not installed at the facility.
This system is no longer installed at the facility.
This system is not installed at the facility.
This system is no longer installed at the facility.
An acceptable question could not be developed for the randomly sampled KIA due to lack of procedural guidance for control or monitoring of Suppression Chamber pressure relative to a Control Room Abandonment and lack of installed instrumentation at the Remote Shutdown Panel.
Randomly resampled KIA 295016 Control Room Abandonment AA2.03 - Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Reactor pressure.
ES-401 Record of Rejected K/As Form ES-401-4 2 / 1 Question 29 An acceptable question could not be developed for the 205000 Shutdown randomly sampled KIA due to limited EOP bases related to Cooling Shutdown Cooling.
2.4.18 -
Randomly resampled KIA 205000 Shutdown Cooling 2.1.27 -
Knowledge of the Knowledge of system purpose and/or function.
specific bases for EOPs.
2 / 1 Question 36 An acceptable question could not be developed for the 212000 RPS randomly sampled KIA K5.01 -
Randomly resampled KIA 212000 RPS K5.02 - Knowledge of Knowledge of the the operational implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM: Specific logic operational arrangements.
implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM: Fuel thermal time constant 2 / 1 Question 49 An acceptable question could not be developed for the 262002 randomly sampled KIA due to lack of indications on the plant Uninterruptable computer for UPS.
Power Supply Randomly resampled KIA 262002 Uninterruptable Power (AC/DC)
Supply (AC/DC) 2.1.23 - Ability to perform specific system 2.1.19 - Ability to and integrated plant procedures during all modes of plant use plant operation.
computers to evaluate system or component status.
ES-401 Record of Rejected K/As Form ES-401-4 2/2 Question 57 An acceptable question could not be developed for the 202002 randomly sampled KIA because the facility does not have Recirculation Flow Hydraulic power units in the Recirculation Flow Control Control System System.
K2.02 -
Randomly resampled KIA 202002 Recirculation Flow Control Knowledge of System A4.03 - Ability to manually operate and/or monitor in electrical power the control room: Lights and alarms.
supplies to the following:
Hydraulic power unit: Plant-Specific 3
Question 67 An acceptable question could not be developed for the 2.1.42 -
randomly sampled KIA at the RO license level because this is Knowledge of new an SRO area of responsibility (1 OCFR55.43(b)(7)).
and spent fuel Randomly resampled KIA 2.1.3 - Knowledge of shift or short-movement term relief turnover practices.
procedures.
3 Question 69 An acceptable Tier 3 question could not be developed for the 2.2.37 - Ability to randomly sampled KIA at the RO license level because this is determine an SRO area of responsibility.
operability and/or Randomly resampled KIA 2.2.13 - Knowledge of tagging and availability of clearance procedures.
safety related equipment.
3 Question 74 An acceptable question could not be developed for the 2.4.32 -
randomly sampled KIA due to lack of specific procedural Knowledge of guidance for a loss of all annunciators at the facility.
operator response Randomly resampled KIA 2.4.19 - Knowledge of EOP layout, to loss of all symbols, and icons.
ES-401 Record of Rejected K/As Form ES-401-4 1 / 2 Question 83 An acceptable question could not be developed for the 295008 High randomly sampled KIA due to lack of a plant specific link Reactor Water between the evolution and heatup rate.
Level Randomly resampled KIA 295008 High Reactor Water Level AA2.04 - Ability to AA2.01 - Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL: Reactor determine and/or water level.
interpret the following as they apply to HIGH REACTOR WATER LEVEL:
Heatup rate:
Plant-Specific 1 / 2 Question 85 An acceptable question could not be developed for the 295020 randomly sampled KIA due to lack of sufficient alarm Inadvertent response manual guidance for an SRO level question for this Containment evolution.
Isolation Randomly resampled KIA 295020 Inadvertent Containment 2.4.50 - Ability to Isolation 2.1.20 - Ability to interpret and execute procedure verify system steps.
alarm setpoints and operate controls identified in the alarm response manual.
2 I 1 Question 90 An acceptable question could not be developed for the 262001 A.C.
randomly sampled KIA at the SRO level due to lack of Electrical suitable EOP entry conditions and immediate action steps Distribution related to the system.
2.4.1 - Knowledge Randomly resampled KIA 262001 A.C. Electrical Distribution 2.2.36 - Ability to analyze the effect of maintenance activities, of EOP entry such as degraded power sources, on the status of limiting conditions and conditions for operations.
immediate action steps.
ES-401 Record of Rejected K/As Form ES-401-4 2/2 Question 92 An acceptable question could not be developed for the 271 000 Off gas randomly sampled KIA at the SRO level without overlapping Question 21.
A2.01 - Ability to Randomly resampled KIA 271000 Offgas A2.04 - Ability to (a)
(a) predict the predict the impacts of the following on the OFF GAS impacts of the SYSTEM; and (b) based on those predictions, use following on the procedures to correct, control, or mitigate the consequences OFFGAS of those abnormal conditions or operations: Offgas system SYSTEM; and (b) high radiation.
based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low condenser vacuum 3
Question 95 An acceptable question could not be developed for the 2.2.15 - Ability to randomly sampled KIA at the SRO level and is better suited determine the to testing on the operating exam.
expected plant Randomly resampled KIA 2.2.12 - Knowledge of surveillance configuration procedures.
using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.