IR 05000445/2018010

NRC Design Bases Assurance Inspection (Teams) Report 05000445/2018010 and 05000446/2018010

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

August 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

Enclosure U.S. NUCLEAR REGULATORY COMMISSION

Inspection Report

Docket Numbers:

05000445, 05000446

License Numbers:

NPF-87, NPF-89

Report Numbers:

05000445/2018010 and 05000446/2018010

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

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-cutting Aspect Report Section Mitigating Systems

Green NCV 05000445/2018010-01; 05000446/2018010-01 Closed None 71111.21M 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-cutting Aspect Report Section Mitigating Systems

Green NCV 05000445/2018010-02; 05000446/2018010-02 Closed None 71111.21M 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 Aspect Report Section Mitigating Systems

Green NCV 05000445/2018010-01; 05000446/2018010-01 Closed None 71111.21M

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 Aspect Report Section Mitigating Systems Green NCV 05000445/2018010-02; 05000446/2018010-02 Closed None 71111.21M

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

212783

5582853

5538622

5609896

240872

261005

3604038

5186334

3659824

273830

4610645

149566

5179229

211820

4598867

4598838

5465353

149564

399132

27245

4842555

4881066

4967904

5174262

4977059

5198308

5542215

5494586

5198308

4064912

5460952

5574642

Procedures

Number

Title

Revision

ABN-301

Instrument Air System Malfunction

ABN-502A

Component Cooling Water System Malfunctions

ABN-602

Response to 6900/480V System Malfunction

ABN-803A

Response to a Fire in the Control Room or Cable

Spreading Room

Procedures

Number

Title

Revision

ABN-803B

Response to a Fire in the Control Room or Cable

Spreading Room

ABN-804A

Response to Fire in the Safeguards Building

ABN-804B

Response to Fire in the Safeguards Building (Unit 2)

ABN-805A

Response to Fire in the Auxiliary Building or the Fuel

Building (Unit 1)

ABN-805B

Response to Fire in the Auxiliary Building or the Fuel

Building (Unit 2)

ABN-806A

Response to Fire in the Electrical and Control

Building (Unit 1)

ABN-806B

Response to Fire in the Electrical and Control

Building (Unit 2)

ABN-807A

Response to Fire in the Containment Building

(Unit 1)

ABN-807B

Response to Fire in the Containment Building

(Unit 2)

ABN-808A

Response to Fire in Service Water Intake Structure

ALM-0032A

Alarm Procedure 1-ALB-3B (Unit 1)

ALM-0032B

Alarm Procedure 1-ALB-3B (Unit 2)

ALM-0102A

Alarm Procedure 1-ALB-10B

ECA 3.1A

Steam Generator Tube Rupture with Loss of Reactor

Coolant Subcooled Recovery Desired (Unit 1)

ECA 3.1B

Steam Generator Tube Rupture with Loss of Reactor

Coolant Subcooled Recovery Desired (Unit 2)

EOP 0.0A

Reactor Trip or Safety Injection (Unit 1)

EOP 0.0B

Reactor Trip or Safety Injection (Unit 2)

EOP 3.0A

Steam Generator Tube Rupture (Unit 1)

EOP 3.0B

Steam Generator Tube Rupture (Unit 2)

EOP-0.0B

Reactor Trip or Safety Injection

INC-7756B

Channel Calibration Reactor Coolant System Wide

Range Pressure and RHR Isolation Valve Interlock

Test

IPO-002A

Startup from Hot Standby

IPO-003A

Power Operations

MSE-C0-6305

6.9KV 7.5 HK Circuit Breaker Enhanced

Maintenance

MSE-GO-6300

Breaker Removal and Installation

Procedures

Number

Title

Revision

MSE-P0-5304

GE DC Switchboards Inspection and Testing

MSE-P0-6000

6.9 KV Switchgear Clean and Inspection

MSE-P0-6305

Station Transformer Maintenance (Dry Type)

MSE-S0-6301

6.9KV Air Circuit Breaker Inspection and Cleaning

MSE-S0-6303

Molded Case Circuit Breaker Test and Inspection

MSE-S1-0602A

Unit 1 train A Electrical UV Relay Test, Response

Time Test and Bus Transfer Test

MSE-S1-0603A

Unit 1 train A UV Relay Calibration and Response

Time Surveillance Test

OPT-108A-2

RSP/STP Switch and Controller Lineup Verification

Data Sheet

OPT-216A

Remote Shutdown Operability Test

OPT-430A

train A Integrated Test Sequence

OPT-512B

ECCS Operability

OPT-512B

Residual Heat Removal and SI Valve Subsystem

Valve Test

OPT-612B

Reactor Coolant System Pressure Boundary

Leakage Test For Loop 1 CL Injection Valves

PPT-S0-6000

Motor Operated Valve Risk-Informed IST

SOP-102B

Residual Heat Removal System

SOP-302A

Feedwater System

SOP-304A

Auxiliary Feedwater System

SOP-304B

Auxiliary Feedwater System

SOP-506

Spent Fuel Pool Cooling and Cleanup System

SOP-815B

Safety Chilled Water System

STA-716

Modification Process

STI-426.02

Processing important OE

TSP-509

Predictive Maintenance Thermographic Analysis

Program

Calculations

Number

Title

Revision

or Date

2-EE-0011

Protection and Ampacity of Electrical Containment

Penetration

2-ME-0071

Unit 2 Component Cooling Water Heat Loads and

Temps for Various Operating Modes

2-ME-0121

Determine Available NPSH(A)

Calculations

Number

Title

Revision

or Date

2-ME-0177

Component Cooling Water Flow Distribution

EE01E-2EB3-2

Cable Sizing Report - Voltage

EE-1E-2EB4-2

Cable Sizing Report - Voltage

EE-1E-BT1ED1

25V DC Battery and Charger Sizing Calculation

EE-CA-0008-0871

Protective Relay Settings for Safeguard Buses

OV/UV Relays and Associated Time Delay Relays

EE-CA-0008-157

Coordination Study of 6.9KV Power Distribution

EE-CA-0008-182

Coordination Study - 125V DC Class 1E Power

Distribution System

EE-SC-U1-1E

Unit 1 and Unit 2 Class 1E System Short Circuit

Study with Unit 1 Preferred Source Lineup

EE-VP-U1-1E

Unit 1 Class 1E System Voltage Profile

ER-ME-089

Resolution of NRC Information Notice IN-92-018

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

ME(3)-073

Component Cooling Water Surge Tank Volume

ME(B)-0267

Component Cooling Water Flow Distribution

ME(B)-071

Component Cooling Water Pump NPSH for MELB

ME(B)-093

Hydraulic Analysis of Component Cooling Water

ME-CA-0000-5478

Fire Safe Shutdown Analysis - MS) - Refueling

Water Storage Tank Gravity Drain Down Time

(to Containment Sumps)

ME-CA-0000-5483

Fire Safe Shutdown Analysis - MSO - HBC-0 Stop

Nut Evaluation in SMB-000 Actuators under stall

conditions

ME-CA-0206-5543

TDAFW Pump Crimped Exhaust Stack Evaluation

ME-CA-0206-5545

TDAFW Pump Crimped Flash Tank Vent Evaluation

ME-CA-0229-5127

The Concerns Raised by SMF-1999-001334 on

Calculation ME(B)-255 Revision 1

ME-CA-0260-5471

RHR Temperature Limits

ME-CA-1100-3356

Component Cooling Water Flow Balance for LOCA

with Flows Throttled

TE-93-56

Component Cooling Water Pump IST Basis

TNE-EE-CA-0008-

265

Selection and Settings of Relays and CTs for Unit 1

and Unit 2

Drawings

Number

Title

Revision

50020445

Penetration Assy Low Voltage Power

T

DDVEN-PL-7551-

1000

Conax Penetration BOM

A

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

Material

CP-89

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

Residual Heat Removal Pump

CP-9

E1-0061, Sheet 23

Motor Operated Valve 1-8811B Sump to Number 2

Residual Heat Removal Pump

CP-10

E1-0061, Sheet 4

Motor Operated Valve 1-8110 Charging Pump

Miniflow Isolation

CP-10

E1-0061, Sheet 5

Motor Operated Valve 1-8111 Charging Pump

Miniflow Isolation

CP-9

E1-0061, Sheet 66

Motor Operated Valve 1-8351A Seal Water Injection

Isolation

CP-5

E1-0062, Sheet 24

Motor Operated Valve 1-8812A Refueling Water

Storage Tank to RHR Pump 1 Isolation

CP-8

E1-0062, Sheet 25

Motor Operated Valve 1-8812B Refueling Water

Storage Tank to RHR Pump 2 Isolation

CP-9

E1-0063, Sheet 2

Motor Operated Valve 1-8701B Residual Heat

Removal Loop 2 Inlet Isolation Valve

CP-7

E1-0063, Sheet 4

Motor Operated Valve 1-8702B Residual Heat

Removal Loop 2 Inlet Isolation Valve

CP-8

E1-2400, Sheet

134

Protective Device Settings - 6.9 kV Safeguard

Buses

CP-1

E1-2400, Sheet

2

Protective Device Settings 6.9KV Safeguard Buses

CP-6

E1-2400, Sheet

153

Protective Device Settings 6.9KV Safeguard Buses

CP-8

E1-2400, Sheet

20

Protective Device Settings 480V Safeguard Buses

CP-6

Drawings

Number

Title

Revision

E1-2400, Sheet

21

Protective Device Settings 480V Safeguard Buses

CP-6

E1-2400, Sheet

2

Protective Device Settings 480V Safeguard Buses

CP-5

E2-0024, Sheet 4

Device Level One Line Diagram Fuse/Breaker Bill of

Material

CP-48

E2-0061, Sheet 4

Motor Operated Valve 2-8110 Charging Pump

Miniflow Isolation

CP-6

E2-0061, Sheet 5

Motor Operated Valve 2-8111 Charging Pump

Miniflow Isolation

CP-8

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-

000172-01-00

Flow Diagram Chemical and Volume Control System

Charging and Positive Displacement Pump Trains

SK-0003-10-

000172-01-01

Chemical and Volume Control

SK-0009-10-

000172-01-01

Vents and Drains System Flow Diagram Auxiliary

Building Leak-offs

Miscellaneous

Number

Title

Revision

or Date

23-ES-012A

Specification Electrical Penetration Assemblies

59EV-2010-

000172-01-00

50.59 Evaluation

April 11, 2012

59SC-2010-

000172-0-03

50.59 Screening

June 3, 2013

59SC-2013-

000185-01-00

Replace 2SI-8819A

February 4,

2014

59SC-2015-

000089-01-00

50.59 Screen for FDA-2015-000089-01

CP-0080B-002

Hermetic Turbopak Safety-Related Chillers

CP-0425-001

6.9kV Metal Clad Switchgear

CP-0430-002

Indoor Low Voltage Metal Enclosed Switchgear

D102601X012

Manual - 67C Series Instrument Supply Regulators

March 2017

DBD-EE-040

6.9kV Electrical Power System

DBD-EE-051

Protection Philosophy

DBD-EE-062

Containment Electrical Penetration Assemblies

FDA 2014-FDA

000130

Design Change UV Setpoints

FDA-2010-000172-

36-07

Multiple Spurious Operations Drain Down of

Refueling Water Storage Tank

FDA-2013-000185-

Lift Check Valve 2SI-8819A Requires Replacement

with a Nozzle Check Valve due to Excessive

Leakage Past the Seat

February 4,

2014

FDA-2014-000134-

01-06

Unfused DC Ammeter Circuits

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

October 2,

2015

Fire Watch Map

Fire Watch No. 18-0007

June 21, 2018

PQE ID:229

Qualification Evaluation: Elec Penetration

System Health

Report

AC Distribution 480 MCCs

4th Qtr 2017

System Health

Report

Switchyard Equipment (EPA, EPB, IPC, EP)

2nd Qtr 2018

TSN-468698

Pressure Regulator 0-60 psig

June 7, 2018

TSN-468699

Pressure Regulator 0-60 psig

June 7, 2018

Miscellaneous

Number

Title

Revision

or Date

VTMR-001-802-

004

Testing and Maintenance of Molded Case Circuit

Breakers

VTMR-001-802-

150

Installation and Maintenance Instructions AV-Line

Switchboards

WCAP-11736-A

Residual Heat Removal System Autoclosure

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

Addressing NSAL 07-11

Design Bases

Documents Number

Title

Revision

DBD-EE-044

DC Power Systems

DBD-EE-051

Protection Philosophy

DBD-ME-229

Component Cooling Water System

DBD-ME-260

Residual Heat Removal System

DBD-ME-261

Safety Injection System

DBD-ME-311

Safety Chilled Water System

ADDITIONAL REQUEST FOR INFORMATION

SUPPLEMENTAL REQUEST FOR INFORMATION

DETAILED RISK EVALUATION

ML18232A057

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