CP-202000261, Inservice Inspection (ISI) and Inservice Testing (IST) Program Relief Requests

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Inservice Inspection (ISI) and Inservice Testing (IST) Program Relief Requests
ML20099D059
Person / Time
Site: Comanche Peak Luminant icon.png
Issue date: 04/07/2020
From: Sewell S
Luminant, TXU Energy, Vistra Operations Company
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-202000261, TXX-20026
Download: ML20099D059 (23)
v • d • e


Text

Steven K. Sewell Senior Director, Engineering & Regulatory Affairs Luminant P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043 o 254.897.6113 CP-202000261 TXX-20026 U. S. Nuclear Regulatory Commission Ref 10 CFR 50.55a ATTN: Document Control Desk Washington, DC 20555-0001 04/07/2020

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT (CPNPP)

DOCKET NO. 50-446 INSERVICE INSPECTION (ISI) AND INSERVICE TEST (IST) PROGRAM RELIEF REQUESTS

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(z)(2), Hardship without a compensating increase in quality and safety, Vistra Operations Company LLC (Vistra OpCo) hereby requests Nuclear Regulatory Commission (NRC) approval to extend scheduled CPNPP inservice inspection (ISI) and inservice test (IST) examinations and testing, from the Unit 2 spring 2020 refueling outage (2RF18) to the Unit 2 fall 2021 refueling outage (2RF19) due to COVID-19 issues.

On March 13, 2020, President Donald Trump declared the Coronavirus (COVID-19) pandemic a national emergency. In addition, Texas Governor Greg Abbott declared a state of disaster due to the COVID-19 pandemic on March 13, 2020. The U.S. Center for Disease Control (CDC) has determined that COVID-19 poses a serious public health risk. The CDC identified the majority of U.S. states reporting community spread of COVID-19.

Currently CPNPP is operating in accordance with the CPNPP Pandemic Response Guideline. Due to the COVID-19 pandemic, there is a desire to minimize the potential of inadvertently spreading the COVID-19 virus to CPNPP personnel from outside contractors who perform examinations and testing for the ISI and IST programs. Due to the potential spread of COVID-19 to CPNPP personnel, Vistra OpCo has identified performance of these examinations and testing as a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2). As an alternative, Vistra OpCo is proposing to delay the inspections for one refueling cycle from spring 2020 (2RF18) to fall 2021 (2RF19).

These augmented examinations in the Unit 2 ISI program are delineated by American Society of Mechanical Engineers (ASME) section IWA-2430(c)(3). These examinations and tests in the Unit 2 IST program are delineated by American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) sections ISTC-5221(c)(3), I-1320(a), I-1350, and I-1350(a).

TXX-20026 Page 2 of 2 CPNPP has identified relief requests that will be required to minimize the spread of the COVID-19 virus. The identified relief requests are listed in the Attachment. Each relief request is included as an Enclosure. CPNPP requests approval of these relief requests on or before April 12, 2020.

This communication contains no new commitments.

Should you have any questions, please contact James Barnette at (254) 897-5866 or james.barnette@luminant.com.

Sincerely,

$K~

Steven K. Sewell Attachment Relief Request List Enclosures 1. CPNPP Unit 2, ISi DEFERRAL RELIEF REQUEST 2A3-2

2. CPNPP Unit 2, 1ST RELIEF REQUEST NUMBER V-2, REVISION 0 c- Scott Morris, Region IV [Scott.Morris@nrc.gov]

Dennis Galvin, NRR [Dennis.Galvin@nrc.gov]

John Ellegood, Senior Resident Inspector, CPNPP [John.Ellegood@nrc.gov]

Neil Day, Resident Inspector, CPNPP [Neil.Day@nrc.gov]

Attachment to TXX-20026 Relief Request List 2RF18 Inservice Inspection Deferral (2A3-2)

Regulation Number 10 CFR 50.55a Regulation Name Codes and Standards Regulation Exemption 10 CFR 50.12, Specific Exemptions being requested under Basis for Exemption 10 CFR 50.12(a)(1) and (a)(2)

IWA-2430(c)(3)

Code Allowance Normal compliance will be restored following CPNPP, Mitigations Measures Unit 2, Cycle 19 in 2RF19 Date Exemption Needed April 12, 2020 2RF18 IST Relief Request Number V-2, Revision 0, Valve Testing Deferral Regulation Number 10 CFR 50.55a Regulation Name Codes and Standards Regulation Exemption 10 CFR 50.12, Specific Exemptions being requested under Basis for Exemption 10 CFR 50.12(a)(1) and (a)(2)

Code Allowance ISTC-5221(c)(3), I-1320(a), I-1350, & I-1350(a)

Normal compliance will be restored following CPNPP, Mitigations Measures Unit 2, Cycle 19 in 2RF19 Date Exemption Needed April 12, 2020 to TXX-20026 Page 1 of 9 Comanche Peak Unit 2 Relief Request Number 2A3-2 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

--Hardship without a compensating increase in quality and safety--

1. ASME Code Component(s) Affected Comanche Peak Unit 2 ASME Class 1, 2, and 3 components and component supports.
2. Applicable Code Edition and Addenda ASME Section XI, 2007 Edition through the 2008 Addenda, Rules for Inservice Inspection of Nuclear Power Plant Components. (Ref. 1)
3. Applicable Requirement The Code of Federal Regulations, Title 10, Part 50, Section 55a (Ref. 2) requires United States Nuclear plants to follow the requirements of ASME BPV Code. This requirement is mandated in paragraph (b), which states in part:

Use and conditions on the use of standards. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code and the ASME OM Code as specified in this paragraph (b).

ASME Section XI, 2007 Edition through the 2008 Addenda, subparagraph IWA-2430(c)(3), which states:

(c) For components inspected under the Inspection Program, the following shall apply:

(3) That portion of an inspection interval described as an inspection period may be reduced or extended by as much as 1 year. This adjustment shall not alter the requirements for scheduling inspection intervals.

4. Reason for Request The U.S. Federal Government made a COVID-19 declaration of emergency pursuant to the Stafford Act on March 13, 2020. The U.S. Center for Disease Control (CDC) has determined that COVID-19 poses a serious public health risk. The CDC has identified many U.S. States reporting community spread of COVID-19 with the anticipation of conditions worsening over the coming weeks.

In response to the COVID-19 Pandemic, Vistra Operations Company LLC (Vistra OpCo) has established the following guidelines and restrictions at Comanche Peak Nuclear Power Plant (CPNPP):

1. Employees who do not have a critical need to be at CPNPP facilities must work remotely.
2. Employees who must work from a CPNPP facility are to practice strict social distancing.

1 to TXX-20026 Page 2 of 9 Comanche Peak Unit 2 Relief Request Number 2A3-2

3. Spring 2020 refueling outage (2RF18) scope shall be reduced to limit the number of supporting contract personnel.

These guidelines and restrictions were established to eliminate the potential of inadvertently spreading the COVID-19 virus to CPNPP personnel and the surrounding community from outside contractors who perform examinations and tests at CPNPP. Also, due to recent travel restrictions and quarantine requirements imposed by both the U.S.

Government and the State of Texas, the availability of outside contractors to provide the NDE services may be limited. Further complicating this situation is the Vistra OpCo reliance on outside NDE resources to supplement the small CPNPP NDE Staff.

Many of the planned 2RF18 Outage activities have been postponed until future outages based on these guidelines and restrictions; however, the 2RF18 ASME Section XI examinations and tests mandated by the Code of Federal Regulations in Title 10, Part 50, Section 55a, (10CFR50.55a) cannot be postponed due to the following circumstances and ASME Section XI requirements.

CPNPP Unit 2 is currently in the Second Period of the Third ISI Inspection Interval. 2RF18 is the last outage of the Second Period and is currently scheduled to begin on April 19, 2020. Completion of the 2RF18 Section XI examinations and tests listed in Table 1 is required to meet IWB-2411, IWC-2411, IWD-2411 and IWF-2410 period completion percentages. Because the Second Period of the Third ISI Inspection Interval ends on August 3, 2020, 2RF18 is the last scheduled opportunity to perform Section XI examinations and tests within the Period.

ASME Section XI, 2007 Edition through the 2008 Addenda, subparagraph IWA-2430(c)(3) allows period extensions and states:

(c) For components inspected under the Inspection Program, the following shall apply:

(3) That portion of an inspection interval described as an inspection period may be reduced or extended by as much as 1 year. This adjustment shall not alter the requirements for scheduling inspection intervals.

Should CPNPP Unit 2 elect to utilize the extension allowed by IWA-2430(c)(3) above, the Second Period of the Third ISI Inspection Interval could be extended as far out as August 3, 2021. However, the next scheduled refueling outage for CPNPP Unit 2 is 2RF19, currently scheduled to begin on October 3, 2021. Therefore, the 1 year extension in IWA-2430(c)(3) does not provide CPNPP the flexibility it needs to defer the required Section XI examinations and tests from 2RF18 to 2RF19.

5. Proposed Alternative and Basis for Use Proposed Alternative Due to the hardship caused by potential spread of COVID-19 to CPNPP personnel and the surrounding community as well as the travel restrictions and quarantine requirements affecting outside contractors, Luminant is proposing that the 1 year inspection period extension as allowed by IWA-2430(c)(3) be increased in accordance with 10 CFR 50.55a(z)(2). As an alternative, Luminant is proposing that the extension allowed by IWA-2430(c)(3) be increased to 18 months due the hardship caused by COVID-19 Pandemic.

2 to TXX-20026 Page 3 of 9 Comanche Peak Unit 2 Relief Request Number 2A3-2 Basis for Use As verified by the Inservice Inspection Program at CPNPP Unit 2 there were no significant flaws identified during the previous examination of the items listed in Table 1 that exceeded the applicable acceptance standards or that were outside of allowable evaluation requirements of ASME Section XI. In addition, there have been no known major indications in any ASME Section XI Safety Systems or Components during any ISI examination that had an adverse effect on the safe operation of CPNPP Unit 2. Most recently, this history is supported by the Luminant submittal to the NRC of CPNPP Unit 2 Third Interval OAR-1 Forms for Outages 2RF15, 2RF16 and 2RF17. OAR-1 Forms are Owners Activity Reports prepared in accordance with ASME Code Case N-532-4; this Code Case is an alternative to Section XI reporting requirements and is approved for use by the NRC in Regulatory Guide 1.147.

Prior to the COVID-19 Pandemic, the Second Period of the Third Interval for CPNPP Unit 2 was on track to be completed as required by ASME Section XI. As a result of the COVID-19 circumstances, extension of the Second Period beyond that allowed by Section XI is necessary. The impact of extending the CPNPP Unit 2 Second Period is detailed below.

All examinations and tests required to be completed by ASME Section XI within the Third ISI Inspection Interval will be completed prior to the end of the Interval. With the approval of this Proposed Alternative, all applicable ASME Section XI period completion percentages will be maintained, including the following requirements:

ASME Section XI, 2007 Edition through the 2008 Addenda, subparagraph IWA-2430(c)(2) states:

(c) For components inspected under the Inspection Program, the following shall apply:

(2) Examinations may be performed to satisfy the requirements of the extended period or interval in conjunction with examinations performed to satisfy the requirements of the successive period or interval. However, an examination performed to satisfy requirements of either the extended period or interval or the successive period or interval shall not be credited to both periods or intervals.

Therefore, if this Proposed Alternative were to be approved, examinations and tests for both the Second Period and Third Period of the Third Interval could be completed concurrently during 2RF19 as long as any given examination is not credited to both periods.

Operations Monitoring Technical Specifications (TS) 3.4.13, RCS Operational LEAKAGE, for each Unit, limits system operation in the presence of leakage from Reactor Coolant System (RCS) components to leakage amounts that do not compromise safety. Surveillance Requirement (SR) 3.4.13.1 requires the performance of RCS water inventory balance to verify RCS leakage is within limits to ensure that the integrity of the Reactor Coolant Pressure Boundary (RCPB) is maintained.

3 to TXX-20026 Page 4 of 9 Comanche Peak Unit 2 Relief Request Number 2A3-2 Per LCO 3.4.13 RCS operational LEAKAGE shall be limited to no pressure boundary LEAKAGE and 1 gpm unidentified LEAKAGE.

Procedure OPT-303, Reactor Coolant System Water Inventory, Attachment 10.2, RCS Live Leakrate Action Levels and Response Guidelines, provides the steps that would be taken to satisfy TS SR 3.4.13.1 by performance of an RCS water inventory balance. A summary of the steps is seen below in A through E:

A. Action Levels based on Absolute Value of Unidentified RCS Inventory Balance (From Surveillance Data)

Entrance Criteria ACTION LEVEL One seven (7) day rolling average of LEVEL 1 Unidentified RCS Inventory Balance values > 0.1 gpm Two consecutive Unidentified RCS LEVEL 2 Inventory values > 0.15 gpm One Unidentified RCS Inventory Balance LEVEL 3 value > 0.3 gpm B. Action Levels based on Deviation from the Baseline Mean:

Entrance Criteria ACTION LEVEL Nine (9) consecutive Unidentified RCS LEVEL 1 Inventory Balance values > baseline mean

[]

Two (2) of three (3) consecutive LEVEL 2 Unidentified RCS Inventory Balance Values > [ + 2], where is the baseline standard deviation One (1) Unidentified RCS Inventory LEVEL 3 Balance > [ + 3]

C. Exit Criteria following Action Level Entry:

Exit Criteria

  • Location of RCS leak has been identified AND Leak terminated (isolated or stopped) - confirmed by RCS leak rate and VCT level trends OR
  • A minimum of 2 daily leak rates with indicated leakage less than the applicable Action Level AND Ops Department and Engineering staff recommend exiting the applicable Action Level D. For any ACTION LEVEL response:
1) Run confirmatory leak rate calculation 4

to TXX-20026 Page 5 of 9 Comanche Peak Unit 2 Relief Request Number 2A3-2

2) Confirm indication
3) Evaluate trend of affected parameters
4) Check for abnormal trend of other leakage indicators E. IF confirmed:
1) Increase monitoring of leakage indicators
2) Initiate a Condition Report to document investigation and results in an evaluation
3) Commence a leak investigation
  • IF ACTION LEVEL 1, search for sources as resources allow
  • IF ACTION LEVEL 2, search for sources, continue 24 hour/day, 7 days/week until cause determined
  • IF ACTION LEVEL 3, search for sources, continue 24 hour/day, 7 days/week until cause determined. At 0.50 gpm sustained leakage, the Shift Manager should consult with the Duty Manager to consider an orderly shutdown to Mode 3.
  • Initiate log book detailing all systems and rooms inspected for leakage with findings. All rooms should be inspected with a thermographic camera (Fire Brigade or Engineering camera).
  • Track all inspections on copies of system drawings, denoting piping and valves inspected.
  • Review recent plant evolutions to determine any suspect source(s).
  • Check any components or flow paths recently changed or placed in service, shutdown, vented, drained, filled, etc.
  • Check any maintenance activity that may have resulted in increasing leakage.
  • Check RCS and seal injection filter seals, vents, and drains for leakage (most common source of leakage).
  • Check any filters recently alternated or changed for leakage from their vents or drains, inspect filter housings for gasket leaks, check seal injection filters and reactor coolant filters for signs of leakage.
  • Scrutinize sump pump run times, sample trends, and rad monitor readings.
  • Divide the primary plant into several large groups for investigation. (e.g.

Leakage to atmosphere inside containment, S/G primary to secondary leakage, leakage to atmosphere outside containment, inter system leakage)

  • Scrutinize tank levels and trends (eg., RHUT, RCDT, PRT and SRST),

utilizing temporary indication if necessary.

  • Initiate outside Containment walk-downs of various portions of potentially affected systems.
  • Notify System Engineer to obtain input/assistance.
4) IF increased leak rate is indicated inside Containment, THEN:
  • Begin planning for Containment entry while carrying out other actions; obtain proper approval for Containment entry.
  • Obtain a Containment Sump sample (during pump out) and analyze for activity, a larger than expected boric acid concentration and other unexpected chemicals.
  • Evaluate other systems for indications of leakage.
  • Obtain a Containment atmosphere sample for indications of RCS leakage.
  • Perform a Containment entry to search for signs of leakage.
  • If necessary, utilize robot to perform loop room inspections.

5 to TXX-20026 Page 6 of 9 Comanche Peak Unit 2 Relief Request Number 2A3-2

5) Identify the source of the increase in leakage.
6) Quantify the leakage.
7) Initiate plan to correct the leak.
8) Monitor Containment airborne radiation levels as well as area radiation monitors and sample Containment atmosphere for indications of RCS leakage.
9) Monitor other Containment parameters (temperature, pressure, humidity, etc.).
10) If the leak source is found and isolated or stopped, re-perform RCS leak rate calculation.

In addition to surveillance required monitoring, Operations continually monitors the RCS leak rate through control board metering and trend graphing capability. Instrumentation such as pressurizer level, sump levels, containment pressure and humidity instruments, and many others are monitored on a continuous basis and are logged and reviewed each shift.

Administrative controls Administrative procedures require monitoring of RCS leakage under the Boric Acid Corrosion Control Program (BACCP) on a per cycle basis. The program also addresses abnormal trends in RCS primary system leakage indicators, which may provide indication of leaks much smaller than TS and RCS leakage levels. CPNPP monitors the following containment building/system parameters during the operating cycle to determine any potential leakage from borated or radioactive systems containing boric acid:

Unidentified Reactor Coolant System Leakage:

This parameter monitors unidentified reactor coolant system leakage both inside and outside of containment. The purpose is to review the data graph for trends in leakage and assess if any increases are indication of borated system leakage in Unit 2 containment.

Containment Air Particulate and Gaseous Concentration:

Radiochemistry analysis are performed of the containment atmosphere for each vent process. The vent process is performed approximately every three days and can identify changes in containment activity. The isotope for detecting increase leakage is noble gas Argon-41. Argon-41 is a short-lived isotope that can be readily monitored to detect small increases or upward trends in leakage. From this parameter, an increase in borated system leakage in the Unit 2 containment can be detected.

Containment Air Cooler Condensate Flow and Containment Sump Level/Pump Rates:

The containment air cooler condensate flow is not separately measured, but flows to the containment sumps. Therefore, the reported sump flow rates include the containment air cooler condensate flows. Review of monthly Unit 2 sump pump volumes for this operating cycle are performed to determine if any upward trend in daily pump volume rates are observed. An increased pump rate could indicate leakage from borated systems in Unit 2 containment.

6 to TXX-20026 Page 7 of 9 Comanche Peak Unit 2 Relief Request Number 2A3-2 Containment Humidity/Dew Point Temperatures:

The Unit 2 containment has five area dew point temperature sensors at containment elevations 832, 860, and 905, as well as the Control Rod Drive Mechanism shroud and the reactor coolant pipe penetrations. Weekly dew point temperature measurements during plant operation are performed to monitor changes in temperature that could be indicative of leakage from borated systems in containment. Minor changes in temperature, both up and down, are gradual and are considered to be normal due to seasonal variations.

Continued monitoring of these parameters will provide early indication of any abnormal unidentified leakage during the upcoming cycle following 2RF18.

Other Considerations Visual examinations of borated systems will be performed by the Boric Acid Corrosion Control Program (BACCP) during 2RF18.

The ISI Pressure Testing Program will implement the Class 1 System Leakage Test required by ASME Section XI Examination Category B-P, Item Number B15.10 at NOP/NOT during heatup from 2RF18; VT-2 Visual Examination will be performed during the test.

Operations looks for leakage during their shiftly tours which includes ASME Section XI piping/components located outside of containment.

Any gross leakage (i.e. water, boric acid, insulation deformation) or structural deformities would be identified during the above examinations/activities and captured in the Corrective Action Program.

6. Duration of Proposed Alternative When approved, the proposed alternative will be implemented at CPNPP Unit 2 starting from the existing Third Interval, Second Period end date of August 3, 2020 to the end of the approved 18-month extension.
7. Precedents Relief Request, I4R-23, was submitted by Exelon Generation Company, LLC for deferral of Section XI requirements at Limerick Generating Station, Unit 1 due to COVID-19 issues, see Accession No. ML20088B022 and Verbal Authorization per Accession No. ML20089A007.
8. References
1. American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section Xl, 2007 Edition through 2008 Addenda
2. 10CFR50.55a, Code and standards, January 2020.

7

Enclosure 1 to TXX-20026 Page 8 of 9 RELIEF REQUEST 2A3-2 TABLE 1 - PREVIOUS EXAMINATION RESULTS NDE ASME ASME PREV.

COMPONENT ID DESCRIPTION PREVIOUS OUTAGE RESULTS TYPE CAT ITEM OUTAGE TCX-1-2100-10-WA MT B-K B10.10 PRESSURIZER SUPPORT SKIRT WELD 2RF12 NO INDICATIONS TCX-1-2100-11 UT B-D B3.110 PRESSURIZER SURGE NOZZLE TO VESSEL WELD 2RF10 NO INDICATIONS TCX-1-2100-11IR UT B-D B3.120 PRSSURIZER SURGE NOZZLE INNER RADIUS 2RF10 NO INDICATIONS TCX-1-3100-2-1 UT B-B B2.40 SG2 CHANNEL HEAD TO TUBESHEET WELD 2RF10 NO INDICATIONS TCX-1-3100-2A UT B-D B3.140 SG2 INLET NOZZLE INNER RADIUS 2RF10 NO INDICATIONS TCX-1-3100-2B UT B-D B3.140 SG2 OUTLET NOZZLE INNER RADIUS 2RF10 NO INDICATIONS TCX-1-3100-4SN1 VT-3 F-A F1.40 S.G. HYD. SNUBBER 2RF12 NO INDICATIONS TCX-1-3100-4SN2 VT-3 F-A F1.40 S.G. HYD. SNUBBER 2RF12 NO INDICATIONS TCX-1-3100-4US1 VT-3 F-A F1.40 S.G. UPPER LATERAL SUPPORT 2RF12 NO INDICATIONS TCX-1-3100-4US2 VT-3 F-A F1.40 S.G. UPPER LATERAL SUPPORT 2RF12 NO INDICATIONS TCX-1-3100-4US3 VT-3 F-A F1.40 S.G. UPPER LATERAL SUPPORT 2RF12 NO INDICATIONS TCX-1-4100-2 UT R-A R1.20 SAFE END TO PIPE 2RF15 NO INDICATIONS PT / VT-TCX-1-4109-4 R-A R1.11 PIPE TO ELBOW (SOCKET WELD) 2RF12 NO INDICATIONS 2

PT / VT-TCX-1-4109-10 R-A R1.11 TEE TO PIPE (SOCKET WELD) 2RF12 NO INDICATIONS 2

TCX-1-4200-2 UT R-A R1.20 SAFE END TO PIPE 2RF15 NO INDICATIONS TCX-1-4300-2 UT R-A R1.20 SAFE END TO PIPE 2RF15 NO INDICATIONS TCX-1-4302-H1 VT-3 F-A F1.10 SNUBBER 2RF11 NO INDICATIONS TCX-1-4302-H3 VT-3 F-A F1.10 STRUT 2RF11 NO INDICATIONS TCX-1-4303-11 UT R-A R1.20 PIPE TO ELBOW 2RF11 NO INDICATIONS TCX-1-4303-H1 VT-3 F-A F1.10 SPRING 2RF11 NO INDICATIONS TCX-1-4303-H1-WA PT B-K B10.20 WELDED ATTACHMENT 2RF11 NO INDICATIONS TCX-1-4400-2 UT R-A R1.20 SAFE END TO PIPE 2RF15 NO INDICATIONS TCX-1-4505-19 UT R-A R1.11 PIPE TO PIPE 2RF12 NO INDICATIONS TCX-1-4505-21 UT R-A R1.20 PIPE TO PIPE 2RF12 NO INDICATIONS TCX-1-4505-23 UT R-A R1.20 PIPE TO PIPE 2RF12 NO INDICATIONS TCX-1-4505-H1 VT-3 F-A F1.10 BOX 2RF07 NO INDICATIONS TCX-1-4505-H27 VT-3 F-A F1.10 STRUT 2RF11 NO INDICATIONS TCX-1-4505-H28 VT-3 F-A F1.10 STRUT 2RF11 NO INDICATIONS TCX-2-1120-1-1 UT C-A C1.20 RHR HX1 HEAD TO SHELL WELD 2RF14 NO INDICATIONS TCX-2-1120-1-1WS-WA PT C-C C3.10 RHR HX1 WELDED SUPPORT 2RF14 NO INDICATIONS TCX-2-1120-1-2 UT C-A C1.10 RHR HX1 SHELL TO FLANGE WELD 2RF14 NO INDICATIONS TCX-2-1120-1-2WS-WA PT C-C C3.10 RHR HX1 WELDED SUPPORT 2RF14 NO INDICATIONS TCX-2-1120-1-3 UT / PT C-B C2.21 RHR HX1 INLET NOZZLE TO SHELL WELD 2RF14 NO INDICATIONS TCX-2-1120-1-4 UT / PT C-B C2.21 RHR HX1 OUTLET NOZZLE TO SHELL WELD 2RF14 NO INDICATIONS TCX-2-1120-1S1 VT-3 F-A F1.40 RHR HX 01 TOP SUPPORTS 2RF14 NO INDICATIONS TCX-2-1120-1S1W-WA VT-1 D-A D1.10 RHR HX 01 TOP SUPPORTS WELDED ATTACHMENT 2RF14 NO INDICATIONS TCX-2-1120-1S2 VT-3 F-A F1.40 RHR HX 01 INTERMEDIATE SUPPORTS 2RF14 NO INDICATIONS RHR HX 01 INTERMEDIATE SUPPORTS WELDED TCX-2-1120-1S2W-WA VT-1 D-A D1.10 2RF14 NO INDICATIONS ATTACHMENT TCX-2-1120-1S3 VT-3 F-A F1.40 RHR HX 01 FOUNDATION SUPPORTS 2RF14 NO INDICATIONS TCX-2-2536-H1 VT-3 F-A F1.20 SNUBBER 2RF12 NO INDICATIONS TCX-2-2536-H1-WA PT C-C C3.20 WELDED ATTACHMENT 2RF12 NO INDICATIONS TCX-2-2575-H1 VT-3 F-A F1.20 BOX 2RF12 NO INDICATIONS 8

Enclosure 1 to TXX-20026 Page 9 of 9 RELIEF REQUEST 2A3-2 TABLE 1 - PREVIOUS EXAMINATION RESULTS NDE ASME ASME PREV.

COMPONENT ID DESCRIPTION PREVIOUS OUTAGE RESULTS TYPE CAT ITEM OUTAGE TCX-2-2575-H5 VT-3 F-A F1.20 STRUT 2RF12 NO INDICATIONS TCX-2-2575-H6 VT-3 F-A F1.20 STRUT 2RF12 NO INDICATIONS TCX-2-2575-H7 VT-3 F-A F1.20 STRUT 2RF12 NO INDICATIONS TCX-2-2575-H8 VT-3 F-A F1.20 SPRING 2RF12 NO INDICATIONS TCX-2-2575-H9 VT-3 F-A F1.20 STRUT 2RF12 NO INDICATIONS TCX-2-2575-H10 VT-3 F-A F1.20 STRUT 2RF12 NO INDICATIONS TCX-2-2577-H5 VT-3 F-A F1.20 STRUT 2RF12 NO INDICATIONS TCX-2-2578-2PS1 VT-3 F-A F1.40 CT PUMP 02 SUPPORT 2RF12 NO INDICATIONS TCX-2-2578-H1 VT-3 F-A F1.20 SPRING 2RF12 NO INDICATIONS TCX-2-2578-H12 VT-3 F-A F1.20 STRUT 2RF12 NO INDICATIONS TCX-2-2578-H14 VT-3 F-A F1.20 STRUT 2RF12 NO INDICATIONS CC-2-SB-035-H4 VT-3 F-A F1.30 BOX 2RF12 NO INDICATIONS CC-2-SB-035-H5 VT-3 F-A F1.30 BOX 2RF12 NO INDICATIONS CC-2-SB-052-H3 VT-3 F-A F1.30 STRUT 2RF12 NO INDICATIONS 9

to TXX-20026 Page 1 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

--Hardship without a compensating increase in quality and safety--

1. ASME Code Component(s) Affected Comanche Peak Nuclear Power Plant (CPNPP), Unit 2 check valves and pressure relief valves that are within the scope of the ASME OM Code and are identified in Table 1, Inservice Tests Requested for Deferral to the fall of 2021 (2RF19).
2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 2004 Edition through the 2006 Addenda
3. Applicable Requirement The Code of Federal Regulations, Title 10, Part 50, Section 55a, Codes and standards, requires United States nuclear power plants to follow the requirements of ASME OM Code. This requirement is mandated in paragraph (b), which states in part:

Use and conditions on the use of standards. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code and the ASME OM Code as specified in this paragraph (b).

ISTA-3120, Inservice Test Interval, subparagraph (a), which states:

The frequency for inservice testing shall be in accordance with the requirements of Section IST.

ISTC-5221, Valve Obturator Movement, (c)(3), which states, for check valves:

At least one valve from each group shall be disassembled and examined at each refueling outage; all valves in each group shall be disassembled and examined at least once every 8 years.

Mandatory Appendix I, Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants, paragraph I-1320, Test Frequencies, Class 1 Pressure Relief Valves, subparagraph (a), which states:

5-Year Test Interval. Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested within each interval; however, a minimum of 20% of the valves from each valve group shall be tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 5-year interval, if they exist. The test interval for any individual valve shall not exceed 5 years.

1 to TXX-20026 Page 2 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Mandatory Appendix I paragraph I-1350, Test Frequency, Classes 2 and 3 Pressure Relief Devices, subparagraph (a), which states:

10-year Test Interval. Classes 2 and 3 pressure relief valves, with the exception of PWR main steam safety valves, shall be tested every 10 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested during any single plant operating cycle; however, a minimum of 20% of the valves from each valve group shall be tested within any 48-month interval. This 20% shall consist of valves that have not been tested during the current 10-year test interval, if they exist.

The test interval for any individual valve shall not exceed 10 years. PWR main steam safety valves shall be tested in accordance with I-1320.

4. Reason for Request Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(2), Vistra Operations Company LLC (Vistra OpCo) is requesting authorization for one-time IST Program interval extensions to the next refueling outage in the fall of 2021 (2RF19) for specific CPNPP, Unit 2, check valves and relief valves due to occupational health and safety concerns associated with pandemic related issues pertaining to the Coronavirus Disease 2019 (COVID-19) outbreak. For each of the identified applicable Code paragraphs, alternative testing is proposed as described in Section 5 of this Relief Request (RR) based on the existence of satisfactory valve performance and capability and that Code-compliance would involve activities that would be detrimental to the occupational health and safety of the workforce and result in the potential to spread the virus. The basis of the request is that compliance results in hardship or unusual difficulty without a compensating increase in level of quality or safety during the current pandemic.

At the current time, CPNPP is reducing activities associated with the upcoming spring 2020 outage (2RF18), which is scheduled to begin on Sunday, April 19, 2020, in order to limit the spread of the COVID-19 virus. As discussed in this RR, many of these activities involve close contact with personnel working in close spaces and thereby limits social distancing capabilities. Further review has determined that close contact required to perform the required valve testing could be detrimental to the occupational health and safety of the workforce and result in the potential to spread the virus. Additionally, the station may experience critical shortages of specially trained and qualified personnel due to illness which will greatly affect completion of these required tests during 2RF18. These tests require several valve maintenance technicians, and other support personnel (e.g.,

scaffold builders, radiation protection technicians, and other maintenance personnel).

Accordingly, Vistra OpCo is requesting approval of this relief on an expedited basis.

The United States government declared a national emergency associated with the COVID-19 outbreak on March 13, 2020. Additionally, in the state of Texas, where CPNPP is located, a Major Disaster Declaration was declared on March 25, 2020, to take actions necessary to reduce exposure to the virus associated with the COVID-19 outbreak. The Centers for Disease Control (CDC) is recommending social distancing as it applies to COVID-19. The CDC defines social distancing as "remaining out of congregate settings, avoiding mass gatherings, and maintaining distance (approximately 6 feet or 2 meters) from others when possible." In the case of performing the valve testing at a nuclear power plant, which requires work in close spaces, this recommendation cannot be effectively implemented.

2 to TXX-20026 Page 3 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Due to the COVID-19 virus pandemic and in an effort to comply with CDC guidance, Vistra OpCo is requesting this one-time relief associated with performing certain check valve and relief valve testing. Section 5 of this RR demonstrates that there is reasonable assurance that the operational readiness of each identified valve will be maintained through the next refueling outage currently scheduled for the fall of 2021 (2RF19). The technical justification utilizes available detailed data from the most recent valve test, and a review of the maintenance history for each valve. This provides the technical justification necessary to show that the proposed alternative is acceptable, and that deferral of the testing in 2020 will not result in an adverse consequence to safety.

5. Proposed Alternative and Basis for Use Proposed Alternative Vistra OpCo is requesting this one-time relief associated with performing the identified valve examination or testing activities in accordance with 10 CFR 50.55a(z)(2) on the basis that compliance results in hardship or unusual difficulty without a compensating increase in level of quality or safety during the current pandemic due to COVID-19 outbreak. Vistra OpCo proposes this one-time relief from the following ASME Code requirements and proposes alternative testing as follows:
1. The ISTC-5221(c)(3) requirement to disassemble and examine at least one valve from each group at each refueling outage for specific CPNPP, Unit 2, check valves and to alternatively defer the disassembly examination to the next refueling outage currently scheduled for the fall of 2021 (2RF19).
2. The I-1350(a) requirement to test specific Class 2 and 3 pressure relief valves every 10 years and to alternatively extend the interval to 11 years so that the specific relief valve test may be deferred to the next refueling outage currently scheduled for the fall of 2021 (2RF19).
3. The I-1350(a) requirement to test a minimum of 20% of the valves from each relief valve group within any 48-month interval for specific CPNPP, Unit 2, relief valves and to alternatively extend the interval to 54-months for specific CPNPP, Unit 2, relief valves so that the specific relief valve tests may be deferred to the next refueling outage currently scheduled for the fall of 2021 (2RF19).
4. The I-1320(a) requirement to test specific PWR Class 2 Main Steam Safety Valves (MSSVs) every 5 years and to alternatively extend the interval to 6 years so that the specific MSSV tests may be deferred to the next refueling outage currently scheduled for the fall of 2021 (2RF19).
5. The I-1320(a) requirement to test a minimum of 20% of the valves from each valve group within any 24-month interval for specific CPNPP, Unit 2, MSSVs and to alternatively extend the interval to 36-months for specific CPNPP, Unit 2, MSSVs so that the specific valve tests may be deferred to the next refueling outage currently scheduled for the fall of 2021 (2RF19) 3 to TXX-20026 Page 4 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 In the event that CPNPP, Unit 2, has an extended shutdown prior to 2RF19, and the US, State, and/or Local Government has relaxed the pandemic restrictions, then a good faith effort to test the valves that were deferred from 2RF18 during the unplanned shutdown will be made in a similar manner as is expected for cold shutdown exercise testing.

An evaluation was performed for each check valve and pressure relief valve identified in Table 1 of this RR that utilized detailed data from the most recent examination or test, and a review of the maintenance history for each valve. This evaluation concluded that the proposed alternative is acceptable, and that deferral of the specific valve testing in 2020 to the next refueling outage in the fall of 2021 will not result in an adverse consequence to safety. No deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each valves performance capability and exclude it from this interval extension RR. Each valve in the scope of this RR is currently on a standard testing interval with acceptable performance. None of the valve test or examination intervals have been extended on the basis of their Risk Rank. Considering the valves current acceptable performance, there is reasonable assurance that each valve will be operationally ready to perform its safety function for the duration of this RR.

In summary, extending the testing interval for each valve in the scope of this RR to the next refueling outage scheduled for the fall of 2021 (2RF19) would not adversely impact the function of the valve or result in a reduction in plant safety. In the current pandemic environment, performing the required tests would result in an increased risk of virus exposure to plant personnel and a reduction in occupational health and safety without a compensating benefit. Therefore, this one-time request for relief meets the criteria in 10 CFR 50.55a(z)(2) for proposing testing alternatives on the basis that compliance results in hardship or unusual difficulty without a compensating increase in level of quality or safety during the current COVID-19 pandemic.

6. Duration of Proposed Alternative The proposed alternative, upon approval, will be implemented at CPNPP, Unit 2, starting from 2RF18, which is scheduled to begin on April 19, 2020, through the end of refueling outage 2RF19, which is scheduled to begin in fall of 2021.
7. Precedent Similar relief from the requirements of 10 CFR 50.55a(z)(2) was verbally authorized by the NRC for Exelon Generation Company, LLC, to use at Limerick Generating Station, Unit 1, via Verbal Authorization by the NRC Office of Nuclear Regulation for Relief Request GVRR-9 Associated with Pandemic-Related Issues - Inservice Testing Interval Extension Motor-Operated Valves, Docket No. 50-352 (EPID L-2020-LLR-0046), dated March 31, 2020. (ML20090A652) 4

Enclosure 2 to TXX-20026 Page 5 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Table 1 Inservice Tests Requested for Deferral to 2RF19 Applicable Valve ID Component Inservice ASME OM

& Risk Basis for Relief

  1. Description Test Type Code Rank Requirement 1 2CT-0145 U2 CS TRN Disassembly ISTC- This check valve has both open and close safety function positions. Open tests B HDR IRC Examination 5221(c)(3), of valves in this group with flow are not practical as they would require or risk Low Risk CHK VLV one valve spraying down containment or fill of containment recirculation sumps. Local each outage - Leak Rate Testing is not required since this valve has a water filled loop seal.

RR Section 5, Therefore, the disassembly examination monitors both the open and close safety Item 1 functions of this valve. This valve and entire 4- valve group are not subject to wear inducing service or degradation. During the last two disassembly examinations of each valve in the group, the manual full stroke exercise, verification that valve internals were structurally sound (no loose or corroded parts), and visual inspection of the valve body (for disc contact), disc, and seat for indications of damage or degradation were completed satisfactory (SAT).

These check valves are not subject to wear inducing service or degradation and past acceptable performance supports reasonable assurance that the check valve would continue to be capable of performing its safety function during the extended interval. Furthermore, no deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each valves performance capability and exclude it from this interval extension RR.

Therefore, the requirement to disassemble and examine one valve this outage could be contrary to the health and safety of plant personnel during the current COVID-19 pandemic. Relief pursuant to 10 CFR 50.55a(z)(2) will allow CPNPP to defer the disassembly examination of any members of this group from 2RF18 to 2RF19.

2 2FW-0082 SG 2-01 FW Disassembly ISTC- The safety function position of this check valve is closed. Due to water hammer HDR CHK Examination 5221(c)(3), interlocks and the difficulty of alternate test methods, it is not practical to monitor Low Risk VLV one valve the close function of the check valve in any other manner. The close function is each outage - redundant with the Feedwater Isolation and Isolation Bypass valves, which are exercised tested once each cycle. The non-safety open function is monitored in 5

Enclosure 2 to TXX-20026 Page 6 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Applicable Valve ID Component Inservice ASME OM

& Risk Basis for Relief

  1. Description Test Type Code Rank Requirement RR Section 5, regular use by verification that Main Feedwater flow to the 2-01 Steam Item 1 Generator is not impaired. This valve is in a group of 4 valves whose disassembly examinations are staggered over a 6-year interval. With one exception, all valves and their internals have been installed and in service over the life of the plant (except for outage periods). In 1994, for valve 2FW-0088, the hinge pins were replaced, and hinge pin bores were repaired. During the last two disassembly examinations of each valve in the group, the manual full stroke exercise, verification that valve internals were structurally sound (no loose or corroded parts), and visual inspection of the valve body (for disc contact),

disc, and seat for indications of damage or degradation were completed satisfactory. Furthermore, no deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each valves performance capability and exclude it from this interval extension RR. Based on this history, the requirement to disassemble and examine this valve during 2RF18 could be detrimental to the occupational health and safety of plant personnel. Relief will allow CPNPP to defer the disassembly examination of any members of this group from 2RF18 to 2RF19.

3 2CS-8442 U2 EMER Disassembly ISTC- The safety function position of this valve is open, and its functional capability is BORATE Examination 5221(c)(3), confirmed each cycle during the Boration Flow Verification test. The Low Risk LN CHK one valve disassembly examination monitors the non-safety close capability of the valve to VLV each outage - meet bi-directional testing requirements. This valve is in a disassembly group of RR Section 5, one valve. At the last inspection, plug and bore measurements of this piston Item 1 check valve were recorded indicating that diametrical clearance and bore ovality remained in tolerance (with additional margin for wear) over the last 9 years of service since the valve design was upgraded to bolted ends to facilitate ease of frequent disassembly examinations. Furthermore, during the last two disassembly examinations of the valve, the manual full stroke exercise, verification that valve internals were structurally sound (no loose or corroded 6

Enclosure 2 to TXX-20026 Page 7 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Applicable Valve ID Component Inservice ASME OM

& Risk Basis for Relief

  1. Description Test Type Code Rank Requirement parts), and visual inspection of the valve bore, disc, and seat for indications of damage or degradation were completed satisfactory. Furthermore, no deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each valves performance capability and exclude it from this interval extension RR. The requirement to disassemble and examine this valve during 2RF18 could be detrimental to the occupational health and safety of plant personnel. Relief will allow CPNPP to defer the disassembly examination of this valve from 2RF18 to 2RF19.

4 2DO-0049 DG 2-02 FO Disassembly ISTC- The safety function position of 2DO-0049 is open, and its open function DAY TK 2- Inspection 5221(c)(3), capability is confirmed during each operation of the Train A Emergency Diesel Low Risk 02 XFER one valve Generator and test of the Train A Fuel Oil Transfer Pumps. The disassembly HDR CHK each outage - examination monitors the non-safety close function of the valve to meet bi-VLV RR Section 5, directional testing requirements. This valve is in a disassembly group of two Item 1 valves. At the 2009 examination of 2DO-0049, plug and bore measurements of this piston check valve were recorded indicating that diametrical clearance and bore ovality remained in tolerance (with additional margin for wear) after 16 years of service. At the 2018 examination of sister valve 2DO-0052, plug and bore measurements indicated that diametrical clearance and bore ovality remained in tolerance (with additional margin for wear) after 25 years of service.

Furthermore, during the last two disassembly examinations of both valves, the manual full stroke exercise, verification that valve internals were structurally sound (no loose or corroded parts), and visual inspection of the valve bore, disc, and seat for indications of damage or degradation were completed satisfactory.

Furthermore, no deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each valves performance capability and exclude it from this interval extension RR. The requirement to disassemble and examine this valve during 2RF19 could be detrimental to the occupational 7

Enclosure 2 to TXX-20026 Page 8 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Applicable Valve ID Component Inservice ASME OM

& Risk Basis for Relief

  1. Description Test Type Code Rank Requirement health and safety of plant personnel. Relief will allow CPNPP to defer the disassembly examination of any members of this group from 2RF18 to 2RF19.

5 2MS-0023 Main Steam Relief Valve I-1320(a), 5- Over the last two 5-year testing cycles for this 20-member valve group, no Safety Test Year Interval additional valves have required testing per I-1320(c)(1). Of the 40 tests 6 2MS-0058 Valves & 20% every reviewed, a set pressure adjustment was required to achieve an as left setting of 7 2MS-0061 24-months - +/- 1% of the required set pressure during only 11 of the 40 tests. For each RR Section 5, valve, set pressure adjustment was required only one time over the last two 8 2MS-0062 Items 4 and 5 testing cycles. Of the 11 valves which required adjustment, if set point drift is 9 2MS-0095 extrapolated to a 6-year test interval, their anticipated as-found values fall within acceptance criteria with >/= 0.5% (6 psig) margin to the +/- 3% as-left 10 2MS-0131 acceptance criteria. Furthermore, no deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each 11 2MS-0132 valves performance capability and exclude it from this interval extension RR.

All Low Therefore, a 6-year interval is acceptable, and one-time relief is requested from Risk the requirement to test each MSSVs at least once every 5 years on the basis that the requirement could be detrimental to the occupational health and safety of plant personnel. Additionally, between 2RF18 and 2RF19, one-time relief from the requirement to test a minimum of 20% of the valves from each valve group within any 24-month interval is requested since compliance with this requirement could be detrimental to the occupational health and safety of plant personnel. Relief will allow CPNPP to defer testing of any members of this group of relief valves from 2RF18 to 2RF19.

12 2-8708B RHR PMP Relief Valve I-1350(a), This valve is in a group of only 2 valves. For this 2-valve group, one valve is 2-02 SUCT Test 20% every 48- required to be tested every other refueling outage such that each is required to High Risk RLF VLV months - RR be tested on a 6-year interval due to the limited group size. Group size is limited Section 5, due to the lack of similar components elsewhere in the plant. During the last two Item 3 successive tests, no additional valves have required testing per I-1350(c)(1),

visual inspections and seat tightness have been satisfactory, and no 8

Enclosure 2 to TXX-20026 Page 9 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Applicable Valve ID Component Inservice ASME OM

& Risk Basis for Relief

  1. Description Test Type Code Rank Requirement adjustments have been required to maintain the set pressure of either valve within the acceptance criteria. Since there is assurance that the valves do not degrade over time, the group interval extension is acceptable. Furthermore, no deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each valves performance capability and exclude it from this interval extension RR. Between 2RF18 and 2RF19, one-time relief from the requirement to test a minimum of 20% of the valves from each valve group within any 48-month interval is requested since compliance with this requirement could be detrimental to the occupational health and safety of plant personnel. Relief will allow CPNPP to defer testing of any members of this group of relief valves from 2RF18 to 2RF19.

13 2VD-0896 RX CAV Relief Valve I-1350(a), 10- 2VD-0896 is in a group of 3 valves. For this 3-valve group, one valve is tested SMP & Test year Test every other re-fueling outage such that each is tested on a 9-year Low Risk CNTMT Interval and interval. During the last two successive tests of each valve in the group, one SMP 2- 20% every 48- valve failed as-found set pressure testing on one occasion. All other as-found 01/2-02 months - RR visual inspection, seat tightness, and set pressure tests were satisfactory. 2VD-DISCH HDR Section 5, 0896 passed its last two successive tests but was replaced with a spare PRESS RLF Items 2 and 3 following its earlier test due to unsatisfactory seat leakage during as-left VLV testing. Similarly, 2WP-7176 passed its last two successive tests, but was replaced with a spare following its later test due to unsatisfactory seat leakage during as-left testing. The one set pressure failure of 2RC-0036 was attributed to set point drift since no other cause for the change was identified during the subsequent disassembly inspection. Despite the test failure, the valve was capable of performing as designed since it would have provided adequate overpressure protection for the protected components. In summary, all three valves were capable of performing as designed over the last two successive test intervals and 2VD-0896 experienced < 1 psi of set point drift between the last two tests. Therefore, extension of the test interval for 2VD-0896 is acceptable.

9

Enclosure 2 to TXX-20026 Page 10 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Applicable Valve ID Component Inservice ASME OM

& Risk Basis for Relief

  1. Description Test Type Code Rank Requirement Furthermore, no deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each valves performance capability and exclude it from this interval extension RR. The 10-year interval for 2VD-0896 expires on April 5, 2021. The addition of 6 months grace period authorized by RR T-1 (ASME OM Code Case OMN-20, Inservice Test Frequencies) may also be insufficient to prevent a violation of the 10 -year test interval prior to 2RF19. Between 2RF18 and 2RF19, one-time relief from the 10-year test interval requirement and the requirement to test a minimum of 20% of the valves from each valve group within any 48-month interval is requested since compliance with these requirements could be detrimental to the occupational health and safety of plant personnel. Relief will allow CPNPP to defer testing of any members of this group of relief valves from 2RF18 to 2RF19.

14 2-8855C SI ACCUM Relief Valve I-1350, 20% 2-8855C is in a group of 4 valves which provide overpressure protection for their 2-03 RLF Test every 48- respective Safety Injection Accumulator. During the last two successive tests of Low Risk VLV months - RR each valve in the group, one valve failed as-found set pressure testing on both Section 5, occasions. All other as-found tests (visual inspection, seat tightness, and set Item 3 pressure) for the other valves were satisfactory, and set pressure adjustment has been required on only one occasion. The two set pressure failures of 2-8855D were attributed to bonding of the disc to the seat due to a bad O-ring, and the valve has since been replaced with a spare and reworked to include O-ring replacement. Despite the test failures, the valve was capable of performing as designed since it would have provided adequate overpressure protection for the respective accumulator. Since all valves in the group were tested in 2017, the requested group extension will not result in any valve exceeding its 10-year maximum interval. 2-8855C has performed consistently during the last two successive tests without the need for adjustment and set pressure drift of only 1 psi over a 5-year interval. Furthermore, no deficiencies, adverse trends or open maintenance work orders were identified that would impact or degrade each 10

Enclosure 2 to TXX-20026 Page 11 of 11 Comanche Peak Unit 2 RELIEF REQUEST NUMBER V-2, REVISION 0 Applicable Valve ID Component Inservice ASME OM

& Risk Basis for Relief

  1. Description Test Type Code Rank Requirement valves performance capability and exclude it from this interval extension RR.

Therefore, extension of the group test interval is acceptable. Between 2RF18 and 2RF19, one-time relief from the requirement to test a minimum of 20% of the valves from each valve group within any 48-month interval is requested since compliance with this requirement could be detrimental to the occupational health and safety of plant personnel. Relief will allow CPNPP to defer testing of any members of this group of relief valves from 2RF18 to 2RF19.

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