IR 05000445/2018010
| ML18232A057 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/20/2018 |
| From: | Thomas Farnholtz Division of Reactor Safety IV |
| To: | Peters K Vistra Operations Company |
| Farnholtz T | |
| References | |
| IR 2018010 | |
| Download: ML18232A057 (26) | |
Text
ust 20, 2018
SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT - NRC DESIGN BASES ASSURANCE INSPECTION (TEAMS) REPORT 05000445/2018010 and 05000446/2018010
Dear Mr. Peters:
On July 12, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Comanche Peak Nuclear Power Plant, Units 1 and 2, and discussed the results of this inspection with Mr. T. McCool, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
Both of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC resident inspector at the Comanche Peak Nuclear Power Plant. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Thomas R. Farnholtz, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-445 and 50-446 License Nos. NPF-87 and NPF-89 Enclosure:
Inspection Report 05000445/2018010 and 05000446/2018010 w/ Attachments:
1. Additional Request for Information 2. Supplemental Request for Information 3. Detailed Risk Evaluation
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Numbers: 05000445, 05000446 License Numbers: NPF-87, NPF-89 Report Numbers: 05000445/2018010 and 05000446/2018010 Enterprise Identifier: I-2018-010-0042 Licensee: Vistra Operations Company, LLC Facility: Comanche Peak Nuclear Power Plant, Units 1 and 2 Location: Glen Rose, Texas Inspection Dates: June 25, 2018, to July 12, 2018 Inspectors: J. Braisted, PhD, Reactor Inspector, Team Lead B. Correll, Reactor Inspector C. Speer, Resident Inspector D. Reinert, PhD, Resident Inspector M. Bloodgood, Emergency Response Specialist R. Deese, Senior Reactor Analyst Accompanying C. Baron, Contractor, Beckman and Associates Personnel: S. Gardner, Contractor, Beckman and Associates Approved By: T. Farnholtz, Chief Engineering Branch 1 Division of Reactor Safety Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting Inspection Procedure 71111.21M, Design Bases Assurance (Teams), at Comanche Peak Nuclear Power Plant, Units 1 and 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. NRC-identified and self-revealed findings, violations, and additional items are summarized in the table below.
List of Findings and Violations Failure to Establish Test Program to Verify Residual Heat Removal Suction Valve Capability Cornerstone Significance Cross- Report cutting Section Aspect Mitigating Green None 71111.21M Systems NCV 05000445/2018010-01; 05000446/2018010-01 Closed The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XI, Test Control, for the failure to establish a test program to ensure that residual heat removal suction isolation valves would perform adequately in service.
Failure to Provide Procedural Guidance for the Failure of a Component Cooling Water Surge Tank Makeup Valve Cornerstone Significance Cross- Report cutting Section Aspect Mitigating Green None 71111.21M Systems NCV 05000445/2018010-02; 05000446/2018010-02 Closed The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, for the failure to provide procedural guidance for the failure of a component cooling water surge tank makeup valve.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs)in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html.
Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
===71111.21MDesign Bases Assurance Inspection (Teams)
The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience during the weeks of June 25 to June 29, 2018, and July 9 to July 12, 2018:
Component ===
- (1) 125 VDC Switchboard 1ED1 a) Component system health and history reports to verify the monitoring of potential degradation.
b) Calculations for electrical distribution, system load flow/voltage drop, short-circuit, and electrical protection to verify that bus capacity and voltages remain within minimum acceptable limits.
c) Protective device settings and circuit breaker ratings to ensure adequate selective protection coordination of connected equipment during worst-case short circuit conditions.
d) Procedures for circuit breaker preventive maintenance, inspection, and testing to compare maintenance practices against industry and vendor guidance.
- (2) Safety-Related Chiller 2-06 a) Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.
b) Calculations for heat loading and thermal performance under accident conditions.
c) Operations procedures for system loading under accident conditions.
d) Preventative maintenance and testing program documents.
- (3) Component Cooling Water (CCW) Pump 2-02 a) Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.
b) Calculations for system flow, system flow balance, net positive suction head, surveillance test acceptance criteria minimum flow, and runout flow.
c) The impact of minimum and maximum allowable electrical power supply frequency on pump performance and net positive suction head.
d) Procedures for operation of the CCW system under accident conditions.
e) Design of the safety-related makeup flowpath to the CCW system.
f) Procedures related to cross-tying the CCW system between units.
- (4) 6900 VAC Switchgear 1EA1 a) Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.
b) Calculations for electrical distribution, system load flow/voltage drop, short-circuit, and electrical protection to verify that bus capacity and voltages remained within minimum acceptable limits.
c) Protective device settings and circuit breaker ratings to ensure adequate selective protection coordination of connected equipment during worst-case short circuit conditions.
d) Procedures for circuit breaker preventive maintenance, inspection, and testing to compare maintenance practices against industry and vendor guidance.
e) Corrective actions associated with a non-cited violation involving undervoltage relay settings documented in the 2013 Component Design Bases Inspection report (ML13214A346).
- (5) 6900/480 VAC Transformer T1EB4 a) Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.
b) Calculations for electrical distribution and electrical protection to verify that transformer capacity and voltages remained within minimum acceptable limits.
c) The protective device settings and circuit breaker ratings to ensure adequate selective protection coordination of connected equipment during worst-case short circuit conditions.
d) Procedures for transformer preventive maintenance, inspection, and testing to compare maintenance practices against industry and vendor guidance.
Component Large Early Release Frequency (LERF) (1 Sample)
- (1) Residual Heat Removal Valve 2-8701B a) Procedures for valve operation during normal, shutdown, and post-accident conditions.
b) Calculations for valve pressure interlock setpoints and interlock surveillance test records.
c) Motor operated valve program calculations for required and available voltage during normal and alternate electrical lineups.
Permanent Modification (5 Samples)
- (1) FDA-2010-000172-01-01, Replace Manual Valve 1-8401A with a Motor Operated Valve
- (2) FDA-2010-000172-36-07, Multiple Spurious Operations Cause Refueling Water Storage Tank Drain Down
- (3) FDA-2013-000185-01-00, Lift Check Valve 2SI-8819A Requires Replacement with a Nozzle Check Valve due to Excessive Leakage Past the Seat
- (4) FDA-2014-000134-01-06, Install 6 amp Fuses in 1E DC Battery Supply
- (5) FDA-2015-000089-01-00, This FDA Validates That 67 CFR Pressure Regulators may be used in Locations where the Design Basis Event is Seismic or Environmentally
Harsh Operating Experience (3 Samples)
- (1) NRC Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire
- (2) NRC Information Notice 2014-04, Potential for Teflon Material Degradation in Containment Penetrations, Mechanical Seals, and Other Components
- (3) NRC Information Notice 2015-05, Inoperability of Auxiliary and Emergency Feedwater Auto-Start Circuits on Loss of Main Feedwater Pumps Evaluation of Inspection Sample Related Operator Procedures and Actions
- (1) Control room operator actions resulting from a simulated steam generator tube rupture (SGTR) accident followed by a post reactor trip loss-of offsite power with a single failure of an intact steam generator atmospheric relief valve.
a) Control room crew was expected to enter procedures for standard post trip actions and SGTR.
b) Following the failure of an intact steam generator atmospheric relief valve, the crew was expected to cooldown using the two remaining atmospheric relief valves.
- (2) In plant operator actions resulting from a loss of instrument air.
a) In plant operators were expected to manually fill the CCW surge tank.
b) Following the loss of instrument air to the CCW surge tank fill valves, the operators were expected to manually operate the fill valves.
INSPECTION RESULTS
Failure to Establish Test Program to Verify Residual Heat Removal Suction Valve Capability Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Green None 71111.21M Systems NCV 05000445/2018010-01; 05000446/2018010-01 Closed
Introduction:
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to establish a test program to ensure that residual heat removal suction isolation valves would perform adequately in service.
Description:
The inspectors reviewed the design and testing associated with the residual heat removal (RHR) suction isolation valves. Each RHR suction line is equipped with two redundant motor operated valves that isolate the higher pressure reactor coolant system from the lower pressure RHR system during normal plant operation. Following a design basis accident, licensed operators open the valves to initiate cooldown using the RHR system.
As discussed in final safety analysis report (FSAR) Appendix 5A, the RHR system is designed to bring the plant from hot shutdown to cold shutdown in a reasonable period of time, assuming the most limiting single failure. To address the limiting single failure of one emergency power train, the two valves in each RHR suction line are powered from different emergency power trains. This arrangement allows that, even with a single failure of an emergency electrical train, both RHR suction lines can maintain their isolation capability.
However, the failure of either emergency power train will prevent the initiation of RHR cooling in the normal manner.
In the event of such a failure, the affected valve can be opened using proceduralized operator actions outside the control room. Normally, valve 8701B is supplied from the train A power supply and valve 8702A from the B power supply. If either of these valves cannot be opened using their normal power supplies, power and control cables for either valve can be swapped to its alternate, unaffected emergency power train. Several abnormal operating procedures include the use of this alternate power lineup for valves 8701B and 8702A.
The inspectors reviewed the periodic testing associated with these motor operated isolation valves and determined that not all valves were being tested in all potential post-accident configurations. Specifically, the licensee was not periodically testing to assure that valve 8701B could be opened using its alternate power supply. A latent failure within the alternate power lineup would result in RHR suction isolation valve 8701B failing to open and could cause a loss of RHR system function.
Corrective Actions: The licensee verified that individual active components within the alternate power supply lineup, including the motor control center breaker and valve operator, are routinely tested. The licensee also initiated an action to test the valves from their alternate power supplies during the next refueling outage.
Corrective Action Reference: CR-2018-004665.
Performance Assessment:
Performance Deficiency: The licensees failure to establish a test program to assure that RHR suction isolation valve 8701B would perform satisfactorily in service, as required by 10 CFR Part 50, Appendix B, Criterion XI, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the mitigating systems cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the lack of testing affected the objective because there was no method to determine the capability of the valve to perform its function in the event of a postulated single failure of an emergency electrical train during an accident which could affect the residual heat removal function.
Significance: The inspectors assessed the significance of the finding using NRC Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, dated October 7, 2016. Using Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability.
Cross-cutting Aspect: No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Contrary to the above, since initial plant startup until July 11, 2018, the licensee failed to establish a test program to assure that testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures. Specifically, the licensee did not establish a test program to assure that RHR suction isolation valve 8701B would perform satisfactorily when powered from its alternate power source.
Disposition: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Provide Procedural Guidance for the Failure of a Component Cooling Water Surge Tank Makeup Valve Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Green None 71111.21M Systems NCV 05000445/2018010-02; 05000446/2018010-02 Closed
Introduction:
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to provide procedural guidance for the failure of a CCW surge tank makeup valve.
Description:
The inspectors reviewed the design of the CCW system and source of makeup to the CCW system. Through a single flowpath, the reactor makeup water system provides the only safety-related makeup to the CCW surge tank in order to accommodate CCW system leakage, to ensure CCW pumps have sufficient net positive suction head, to allow for thermal expansion and contraction of the CCW system, and to provide a means of CCW system overpressure protection.
Valve 4500-1 is a safety-related, fail-open, air-operated valve in this single flowpath and is considered part of the CCW system. This valve is normally closed. During a design basis accident, when level in the CCW surge tank reaches the lo-lo setpoint, the safeguards loops automatically isolate and an alarm response procedure directs the operators to ensure valve 4500-1 is open. If valve 4500-1 were to fail in the closed position, or if any other component in the single flowpath were to fail, there are currently no instructions or procedures to provide alternate makeup methods to the CCW surge tank.
As discussed in CCW FSAR Section 9.2.2.2.1, the failure or malfunction of any single active or passive component does not prevent fulfillment of the CCW system safeguards functions.
However, the only safety-related source of makeup to the CCW surge tank is a single flowpath from the reactor makeup water system. Because the CCW system would be required to operate in the long term following a design basis accident, a source of makeup water would be required to accommodate isolation valve leakage, among other purposes. A postulated single failure in this flowpath could prevent fulfillment of the CCW system safeguards functions.
Additionally, as discussed in CCW design basis document DBD-ME-229, Section 5.4.2, and CCW FSAR Table 9.2-5, if the reactor makeup valve 4500-1 fails in the closed position as a result of an electrical or mechanical single failure within the valve, an operator action to open the valve by venting the diaphragm and/or forcing the valve open may be required. There were no instructions or procedures directing the operators to take these actions or to establish an alternate source of makeup water to the CCW surge tank to ensure functionality of the CCW system.
Corrective Actions: The licensee implemented a compensatory measure, failing open valve 4500-1 by removing air to it, until permanent corrective actions are accomplished.
Corrective Action Reference: IR-2018-004603 and IR-2018-004701.
Performance Assessment:
Performance Deficiency: The licensees failure to provide procedural guidance for the failure of a CCW surge tank makeup valve, as required by 10 CFR Part 50, Appendix B, Criterion V, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the mitigating systems cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, given a postulated single failure of valve 4500-1, or another component in the single makeup flowpath, the lack of procedural guidance for ensuring makeup to the CCW surge tank during an accident could affect the ability of the CCW system to perform its safeguards function.
Significance: The inspectors assessed the significance of the finding using NRC Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, dated October 7, 2016. Using Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding required a detailed risk evaluation because the finding involved a deficiency affecting the design and qualification of a mitigating structure, system, or component that lost its operability or functionality and represented a loss of system function. A Region IV senior reactor analyst performed a detailed risk evaluation and determined that the bounding increase in core damage frequency for this issue was 7.9E-8/year for both units, and the finding was therefore of very low safety significance (Green). Additional information regarding the detailed risk evaluation is found in 3 of this report.
Cross-cutting Aspect: No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, required, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Contrary to the above, since initial plant startup until July 12, 2018, the licensee failed to prescribe by documented instructions, procedures, or drawings, of a type appropriate to the circumstances activities affecting quality. Specifically, the licensee failed to provide procedural guidance for the failure of CCW surge tank makeup valve 4500-1, or the failure of another component, in the single safety-related makeup flowpath.
Disposition: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
On July 12, 2018, the inspectors presented the results of this design bases assurance inspection to Mr. T. McCool, Site Vice President, and other members of the licensee staff. The inspectors verified no proprietary information was retained or documented in this report.
DOCUMENTS REVIEWED
71111.21MDesign Bases Assurance Inspection (Teams)
Condition Reports (CRs) (Reviewed)
CR-2013-006252 CR-2010-004244 CR-2014-010113 CR-2017-001489
CR-2016-010346 CR-2011-001742 CR-2016-008215 CR-2018-004696
CR-2010-005563 CR-2015-008517 CR-2018-001530 CR-2015-007625
CR-2015-009942 CR-2015-009839 CR-2014-004995 CR-2013-008401
CR-2015-011497 CR-2015-010339 CR-2015-010000 CR-2015-009979
CR-2017-000269 CR-2016-007653 CR-2016-003348 CR-2015-011913
EV-CR-2014-003591 CR-2016-010346 TR-2014-009407 TR-2018-003301
CR-2008-000089 CR-2017-007437 CR-2018-001532 CR-2014-010279
CR-2015-004579 CR-2017-012024 CR-2018-000941 CR-2018-004367
CR-2018-001372 CR-2018-000940 CR-2017-011633 CR-2017-004995
CR-2018-004259 CR-2018-003136 CR-2018-002879 CR-2018-001671
CR-2017-002493 EV-CR-2012-007312 IR-2018-004369 CR-2017-002493
CR-2012-007312 OER-2017-004566
Condition Reports (CRs) (Issued)
IR-2018-004602 IR-2018-004612 IR-2018-004637 CR-2018-004638
IR-2018-004367 CR-2018-004369 CR-2018-004448 IR-2018-004390
CR-2018-004403 IR-2018-004447 CR-2018-004597 CR-2018-004660
CR-2018-004665 IR-2018-004603 IR-2018-004624 IR-2018-004649
IR-2018-004660 IR-2018-004665 CR-2018-004447 IR-2018-004701
Work Orders
4747766 5180554 5438272 5464313
5588117 5063996 5212783 5582853
5538622 5609896 4240872 5261005
3604038 5186334 3659824 5273830
4610645 149566 5179229 5211820
4598867 4598838 5465353 149564
399132 4827245 4842555 4881066
4967904 5174262 4977059 5198308
5542215 5494586 5198308 4064912
5460952 5574642
Procedures
Number Title Revision
ABN-301 Instrument Air System Malfunction 13
ABN-502A Component Cooling Water System Malfunctions 9
ABN-602 Response to 6900/480V System Malfunction 8
ABN-803A Response to a Fire in the Control Room or Cable 13
Spreading Room
Procedures
Number Title Revision
ABN-803B Response to a Fire in the Control Room or Cable 10
Spreading Room
ABN-804A Response to Fire in the Safeguards Building 6
ABN-804B Response to Fire in the Safeguards Building (Unit 2) 4
ABN-805A Response to Fire in the Auxiliary Building or the Fuel 8
Building (Unit 1)
ABN-805B Response to Fire in the Auxiliary Building or the Fuel 7
Building (Unit 2)
ABN-806A Response to Fire in the Electrical and Control 8
Building (Unit 1)
ABN-806B Response to Fire in the Electrical and Control 6
Building (Unit 2)
ABN-807A Response to Fire in the Containment Building 8
(Unit 1)
ABN-807B Response to Fire in the Containment Building 6
(Unit 2)
ABN-808A Response to Fire in Service Water Intake Structure 6
ALM-0032A Alarm Procedure 1-ALB-3B (Unit 1) 7
ALM-0032B Alarm Procedure 1-ALB-3B (Unit 2) 3
ALM-0102A Alarm Procedure 1-ALB-10B 12
ECA 3.1A Steam Generator Tube Rupture with Loss of Reactor 9
Coolant Subcooled Recovery Desired (Unit 1)
ECA 3.1B Steam Generator Tube Rupture with Loss of Reactor 9
Coolant Subcooled Recovery Desired (Unit 2)
EOP 0.0A Reactor Trip or Safety Injection (Unit 1) 9
EOP 0.0B Reactor Trip or Safety Injection (Unit 2) 9
EOP 3.0A Steam Generator Tube Rupture (Unit 1) 9
EOP 3.0B Steam Generator Tube Rupture (Unit 2) 9
EOP-0.0B Reactor Trip or Safety Injection 9
INC-7756B Channel Calibration Reactor Coolant System Wide 4
Range Pressure and RHR Isolation Valve Interlock
Test
IPO-002A Startup from Hot Standby 21
IPO-003A Power Operations 30
MSE-C0-6305 6.9KV 7.5 HK Circuit Breaker Enhanced 3
Maintenance
MSE-GO-6300 Breaker Removal and Installation 3
Procedures
Number Title Revision
MSE-P0-5304 GE DC Switchboards Inspection and Testing 2
MSE-P0-6000 6.9 KV Switchgear Clean and Inspection 7
MSE-P0-6305 Station Transformer Maintenance (Dry Type) 6
MSE-S0-6301 6.9KV Air Circuit Breaker Inspection and Cleaning 6
MSE-S0-6303 Molded Case Circuit Breaker Test and Inspection 8
MSE-S1-0602A Unit 1 train A Electrical UV Relay Test, Response 2
Time Test and Bus Transfer Test
MSE-S1-0603A Unit 1 train A UV Relay Calibration and Response 7
Time Surveillance Test
OPT-108A-2 RSP/STP Switch and Controller Lineup Verification 13
Data Sheet
OPT-216A Remote Shutdown Operability Test 14
OPT-430A train A Integrated Test Sequence 7
OPT-512B ECCS Operability 6
OPT-512B Residual Heat Removal and SI Valve Subsystem 11
Valve Test
OPT-612B Reactor Coolant System Pressure Boundary 3
Leakage Test For Loop 1 CL Injection Valves
PPT-S0-6000 Motor Operated Valve Risk-Informed IST 3
SOP-102B Residual Heat Removal System 15
SOP-302A Feedwater System 19
SOP-304A Auxiliary Feedwater System 17
SOP-304B Auxiliary Feedwater System 13
SOP-506 Spent Fuel Pool Cooling and Cleanup System 21
SOP-815B Safety Chilled Water System 11
STA-716 Modification Process 26
STI-426.02 Processing important OE 0
TSP-509 Predictive Maintenance Thermographic Analysis 6
Program
Calculations Revision
Number Title or Date
2-EE-0011 Protection and Ampacity of Electrical Containment 11
Penetration
2-ME-0071 Unit 2 Component Cooling Water Heat Loads and 1
Temps for Various Operating Modes
2-ME-0121 Determine Available NPSH(A) 0
Calculations Revision
Number Title or Date
2-ME-0177 Component Cooling Water Flow Distribution 0
EE01E-2EB3-2 Cable Sizing Report - Voltage 7
EE-1E-2EB4-2 Cable Sizing Report - Voltage 6
EE-1E-BT1ED1 125V DC Battery and Charger Sizing Calculation 7
EE-CA-0008-0871 Protective Relay Settings for Safeguard Buses 18
OV/UV Relays and Associated Time Delay Relays
EE-CA-0008-157 Coordination Study of 6.9KV Power Distribution 4
EE-CA-0008-182 Coordination Study - 125V DC Class 1E Power 3
Distribution System
EE-SC-U1-1E Unit 1 and Unit 2 Class 1E System Short Circuit 5
Study with Unit 1 Preferred Source Lineup
EE-VP-U1-1E Unit 1 Class 1E System Voltage Profile 5
ER-ME-089 Resolution of NRC Information Notice IN-92-018 0
Potential Loss of Remote Shutdown Capability
Following Control Room Fire
FSD/SS-TBX-340 Residual Heat Removal Initiation Window April 29, 1982
IC(B)-064 Main Steam Valve Air Pressure 1
ME(3)-073 Component Cooling Water Surge Tank Volume 3
ME(B)-0267 Component Cooling Water Flow Distribution 1
ME(B)-071 Component Cooling Water Pump NPSH for MELB 3
ME(B)-093 Hydraulic Analysis of Component Cooling Water 1
ME-CA-0000-5478 Fire Safe Shutdown Analysis - MS) - Refueling 0
Water Storage Tank Gravity Drain Down Time
(to Containment Sumps)
ME-CA-0000-5483 Fire Safe Shutdown Analysis - MSO - HBC-0 Stop 1
Nut Evaluation in SMB-000 Actuators under stall
conditions
ME-CA-0206-5543 TDAFW Pump Crimped Exhaust Stack Evaluation 0
ME-CA-0206-5545 TDAFW Pump Crimped Flash Tank Vent Evaluation 0
ME-CA-0229-5127 The Concerns Raised by SMF-1999-001334 on 0
Calculation ME(B)-255 Revision 1
ME-CA-0260-5471 RHR Temperature Limits 0
ME-CA-1100-3356 Component Cooling Water Flow Balance for LOCA 0
with Flows Throttled
TE-93-56 Component Cooling Water Pump IST Basis 0
TNE-EE-CA-0008- Selection and Settings of Relays and CTs for Unit 1 4
265 and Unit 2
Drawings
Number Title Revision
50020445 Penetration Assy Low Voltage Power T
DDVEN-PL-7551- Conax Penetration BOM A
1000
E1-0001 Plant One Line Diagram CP-33
E1-0004 6.9 KV Auxiliaries One Line Diagram CP-41
E1-0024, Sheet 4 Device Level One Line Diagram Fuse/Breaker Bill of CP-89
Material
E1-0031, Sheet 1 6.9 KV Switchgear Bus 1EA1 CP-10
E1-0031, Sheet 21 6.9 KV Switchgear Bus 1EA1 Diesel Breaker CP-11
E1-0031, Sheet 3 6.9 KV Switchgear Bus 1EA1 Breaker 1EA1-2 CP-19
E1-0061, Sheet 22 Motor Operated Valve 1-8811A Sump to Number 1 CP-9
E1-0061, Sheet 23 Motor Operated Valve 1-8811B Sump to Number 2 CP-10
E1-0061, Sheet 4 Motor Operated Valve 1-8110 Charging Pump CP-10
Miniflow Isolation
E1-0061, Sheet 5 Motor Operated Valve 1-8111 Charging Pump CP-9
Miniflow Isolation
E1-0061, Sheet 66 Motor Operated Valve 1-8351A Seal Water Injection CP-5
Isolation
E1-0062, Sheet 24 Motor Operated Valve 1-8812A Refueling Water CP-8
Storage Tank to RHR Pump 1 Isolation
E1-0062, Sheet 25 Motor Operated Valve 1-8812B Refueling Water CP-9
Storage Tank to RHR Pump 2 Isolation
E1-0063, Sheet 2 Motor Operated Valve 1-8701B Residual Heat CP-7
Removal Loop 2 Inlet Isolation Valve
E1-0063, Sheet 4 Motor Operated Valve 1-8702B Residual Heat CP-8
Removal Loop 2 Inlet Isolation Valve
E1-2400, Sheet Protective Device Settings - 6.9 kV Safeguard CP-1
134 Buses
E1-2400, Sheet Protective Device Settings 6.9KV Safeguard Buses CP-6
2
E1-2400, Sheet Protective Device Settings 6.9KV Safeguard Buses CP-8
153
E1-2400, Sheet Protective Device Settings 480V Safeguard Buses CP-6
20
Drawings
Number Title Revision
E1-2400, Sheet Protective Device Settings 480V Safeguard Buses CP-6
21
E1-2400, Sheet Protective Device Settings 480V Safeguard Buses CP-5
2
E2-0024, Sheet 4 Device Level One Line Diagram Fuse/Breaker Bill of CP-48
Material
E2-0061, Sheet 4 Motor Operated Valve 2-8110 Charging Pump CP-6
Miniflow Isolation
E2-0061, Sheet 5 Motor Operated Valve 2-8111 Charging Pump CP-8
Miniflow Isolation
M1-0229 Flow Diagram Component Cooling Water System CP-23
M1-0229, Sheet A Flow Diagram Component Cooling Water System CP-21
M1-0229, Sheet B Flow Diagram Component Cooling Water System CP-25
M1-0307, Sheet A Flow Diagram Chilled Water System CP-8
M1-0307, Sheet B Flow Diagram Chilled Water System CP-8
M1-0307, Sheet C Flow Diagram Chilled Water System CP-4
M2-0229 Flow Diagram Component Cooling Water System CP-19
M2-0229, Sheet A Flow Diagram Component Cooling Water System CP-14
M2-0229, Sheet B Flow Diagram Component Cooling Water System CP-15
M2-0263 Flow Diagram Safety Injection System CP-17
M2-0263, Sheet A Flow Diagram Safety Injection System CP-7
M2-0263, Sheet B Flow Diagram Safety Injection System CP-13
M2-0263, Sheet C Flow Diagram Safety Injection System CP-7
M2-0307, Sheet A Flow Diagram Chilled Water System CP-14
M2-0311 Flow Diagram Safety Chilled Water System CP-9
M2-0311, Sheet A Flow Diagram Safety Chilled Water System CP-6
SK-0001-10- Flow Diagram Chemical and Volume Control System 00
000172-01-00 Charging and Positive Displacement Pump Trains
SK-0003-10- Chemical and Volume Control 01
000172-01-01
SK-0009-10- Vents and Drains System Flow Diagram Auxiliary 01
000172-01-01 Building Leak-offs
Miscellaneous Revision
Number Title or Date
23-ES-012A Specification Electrical Penetration Assemblies 0
59EV-2010- 50.59 Evaluation April 11, 2012
000172-01-00
59SC-2010- 50.59 Screening June 3, 2013
000172-0-03
59SC-2013- Replace 2SI-8819A February 4,
000185-01-00 2014
59SC-2015- 50.59 Screen for FDA-2015-000089-01 0
000089-01-00
CP-0080B-002 Hermetic Turbopak Safety-Related Chillers 19
CP-0425-001 6.9kV Metal Clad Switchgear 28
CP-0430-002 Indoor Low Voltage Metal Enclosed Switchgear 29
D102601X012 Manual - 67C Series Instrument Supply Regulators March 2017
DBD-EE-040 6.9kV Electrical Power System 18
DBD-EE-051 Protection Philosophy 44
DBD-EE-062 Containment Electrical Penetration Assemblies 17
FDA 2014-FDA Design Change UV Setpoints 1
000130
FDA-2010-000172- Multiple Spurious Operations Drain Down of 7
36-07 Refueling Water Storage Tank
FDA-2013-000185- Lift Check Valve 2SI-8819A Requires Replacement February 4,
with a Nozzle Check Valve due to Excessive 2014
Leakage Past the Seat
FDA-2014-000134- Unfused DC Ammeter Circuits 6
01-06
FDA-2015-000089- This FDA validates that 67 CFR pressure regulators October 2,
01-00 may be used in locations where the design basis 2015
event is seismic or environmentally harsh
Fire Watch Map Fire Watch No. 18-0007 June 21, 2018
PQE ID:229 Qualification Evaluation: Elec Penetration 1
System Health AC Distribution 480 MCCs 4th Qtr 2017
Report
System Health Switchyard Equipment (EPA, EPB, IPC, EP) 2nd Qtr 2018
Report
TSN-468698 Pressure Regulator 0-60 psig June 7, 2018
TSN-468699 Pressure Regulator 0-60 psig June 7, 2018
Miscellaneous Revision
Number Title or Date
VTMR-001-802- Testing and Maintenance of Molded Case Circuit
004 Breakers
VTMR-001-802- Installation and Maintenance Instructions AV-Line
150 Switchboards
WCAP-11736-A Residual Heat Removal System Autoclosure 0
Interlock Removal Report for the Westinghouse
Owners Group
White Paper Evaluation of Timing Associated with Refueling
Water Storage Tank Drain Down through a
Spuriously Open Containment Sump Isolation Valve
WPT-17834 Steam Generator Tube Rupture Margin to Overfill 0
Addressing NSAL 07-11
Design Bases
Documents Number Title Revision
DBD-EE-044 DC Power Systems 27
DBD-EE-051 Protection Philosophy 44
DBD-ME-229 Component Cooling Water System 41
DBD-ME-260 Residual Heat Removal System 29
DBD-ME-261 Safety Injection System 36
DBD-ME-311 Safety Chilled Water System 18
ADDITIONAL REQUEST FOR INFORMATION
SUPPLEMENTAL REQUEST FOR INFORMATION
DETAILED RISK EVALUATION
SUNSI Review: ADAMS: Non-Publicly Available Non-Sensitive Keyword: NRC-002
By: JDB Yes No Publicly Available Sensitive
OFFICE RI:EB1 RI:EB2 RI:PBD RI:PBD SRA:PSB2 ERC:RCB C:EB1
NAME JBraisted BCorrell DReinert CSpeer RDeese MBloodgood TFarnholtz
SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ /RA/
DATE 08/14/2018 07/25/2018 07/25/2018 07/26/2018 07/30/2018 08/16/2018 08/16/2018
OFFICE C:PBA C:EB1
NAME MHaire TFarnholtz
SIGNATURE /RA/ /RA/
DATE 08/16/2018 08/20/2018