PLA-8062, Supplement to Relief Request 4RR-10 Relief from Code Seal Weld Requirement for Valve 252F007B, PLA-8062

Supplement to Relief Request 4RR-10 Relief from Code Seal Weld Requirement for Valve 252F007B, PLA-8062

ML23101A153
Person / Time
Site:	Susquehanna
Issue date:	04/11/2023
From:	Casulli E Susquehanna, Talen Energy
To:	Document Control Desk
References
PLA-8062
Download: ML23101A153 (1)

Text

Edward Casulli Susquehanna Nuclear, LLC Site Vice President 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3795 Fax 570.542.1504 Edward.Casulli@TalenEnergy.com

April 11, 2023

Attn: Document Control Desk 10 CFR 50.55a U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUSQUEHANNA STEAM ELECTRIC STATION SUPPLEMENT TO RELIEF REQUEST 4RR-10 RELIEF FROM CODE SEAL WELD REQUIREMENT FOR VALVE 252F007B PLA-8062 Docket No. 50-388

Reference:

Susquehanna letter to NRC, Relief Request 4RR-10 Relief from Code Seal Weld Requirement for Valve 252F007B (PLA-8061), dated April 10, 2023 (ADAMS Accession No. ML23100A128).

In accordance with 10 CFR 50.55a(z)(2), Susquehanna Nuclear, LLC (Susquehanna), in the referenced letter, requested NRC approval of relief request 4RR-10 associated with the fourth Inservice Inspection interval for the Susquehanna Steam Electric Station, Unit 2. Relief request 4RR-10 requests authorization of alternative requirements for the replacement of the threaded leakoff port plug in Valve 252F007B, Core Spray Injection to Reactor Vessel Valve in accordance with IWA-4000 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI.

On April 11, 2023, Susquehanna met with members of the NRC to discuss the relief request.

During that meeting, supplemental information required for the NRC to complete the review of relief request 4RR-10 was identified. Enclosure 1 to this letter provides the required supplemental information.

Susquehanna requests authorization of the proposed alternative by April 12, 2023.

There are no new or revised regulatory commitments contained in this submittal.

Document Control Desk PLA-8062

Should you have any questions regarding this submittal, please contact Ms. Katie Brown, Acting Manager - Nuclear Regulatory Affairs, at (570) 542-3407.

E. Casulli

Enclosures:

1. Supplement to Relief Request 4RR-10

2. Valve Data Sheet for Valve 252F007B

Copy: NRC Region I Mr. C. Highley, NRC Senior Resident Inspector Ms. A. Klett, NRC Project Manager Mr. M. Shields, PA DEP/BRP

Enclosure 1 to PLA-8062

Supplement to Relief Request 4RR-10

Enclosure 1 to PLA-8062 Page 1 of 3

Supplement to Relief Request 4RR-10

In accordance with 10 CFR 50.55a(z)(2), Susquehanna Nuclear, LLC (Susquehanna), in Reference 1, requested NRC approval of relief request 4RR-10 associated with the fourth Inservice Inspection (ISI) interval for the Susquehanna Steam Electric Station (SSES), Unit 2.

Relief request 4RR-10 requests authorization of alternative requirements for the replacement of the threaded leakoff port plug in Valve 252F007B, Core Spray Injection to Reactor Vessel Valve in accordance with IWA-4000 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI.

On April 11, 2023, Susquehanna met with members of the NRC to discuss the relief request.

During that meeting, additional information required for the NRC to complete the review of relief request 4RR-10 was identified. The required information is provided herein.

Leakage Monitoring

In Reference 1, Enclosure 1, Section 5, Susquehanna stated that drywell leakage is monitored in accordance with Technical Specification (TS) Surveillance Requirement (SR) 3.4.4.1. TS Limiting Condition of Operation (LCO) 3.4.4 places the following limits on Reactor Coolant System (RCS) Leakage:

a. No pressure boundary leakage;

b. 5 gpm unidentified leakage;

c. 25 gpm total leakage averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and

d. 2 gpm increase in unidentified leakage within the previous 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period in Mode 1.

SR 3.4.4.1 requires verification that the RCS leakage limits specified in LCO 3.4.4 are met. The frequency of the SR is controlled under the Surveillance Frequency Control Program, and currently set at once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Should leakage occur from Valve 252F007B, this leakage would register during performance of SR 3.4.4.1. If any of the leakage limits are exceeded, Required Actions A.1 and B.1 require reducing leakage to within the limits within four hours or shutting the unit down in the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. LCO 3.4.4 is required to be met during Modes 1, 2, and 3, and will help ensure Valve 252F007B remains free of unacceptable leakage during the period the proposed alternative is in place.

Further, Susquehanna implements procedure ON-DWLEAK-201, Drywell Leakage, which prescribes a graded approach for responding to drywell leakage up to and including a unit shutdown prior to reaching the TS limits for leakage. Currently, ON-DWLEAK-201 requires, in part, the following actions:

Enclosure 1 to PLA-8062 Page 2 of 3

Drywell Leakage Required Action

Develop Adverse Condition Monitoring Plan 0.3 gpm and increasing and/or Operations Decision Matrix in accordance with applicable procedure to include an action plan with clear action levels.

> 1.5 gpm

> 0.75 gpm AND confirmed reactor coolant Schedule a unit shutdown in accordance with leakage procedure GO-200-004, Plant Shutdown to Minimum Power.

0.5 gpm increase during previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

> 4.0 gpm Initiate a unit shutdown in accordance with procedure GO-200-004.

Once the presence of drywell leakage has been identified, there are multiple methods available to Susquehanna to help determine the location of the leakage. The first is to analyze a sample from the drywell sump to determine if the leakage is reactor coolant leakage. This information helps limit the scope of investigation.

Second, the drywell temperature elements are trended to determine if any temperature changes have occurred. Valve 252F007B is located on elevation 767 of the drywell. Two temperature elements are located on the 799 elevation of the drywell. Since any postulated leakage through the leakoff port will flash to steam, these temperature elements are appropriately located to capture potential temperature changes due to Valve 252F007B leakage. These temperature elements are in the conical section of the drywell so any temperature changes due to a leak are amplified because of the reduced air volume compared to the rest of the drywell.

Third are drywell containment radiation monitors. These monitors will help identify any potential leakage sources from the reactor coolant pressure boundary.

Based on compliance with TS 3.4.4 and ON-DWLEAK-201, Susquehanna is required to take appropriate actions in the event of a leak from Valve 252F007B, up to and including a unit shutdown to address the leak. As discussed above, once the presence of any leakage is identified, Susquehanna personnel have multiple methods, with the primary being the temperature elements on the 799 elevation of the drywell, to help determine the location of the leakage. The approximate leakage location helps inform the necessary plans established per ON-DWLEAK-201. Therefore, consequences associated with any leakage through the leakoff

Enclosure 1 to PLA-8062 Page 3 of 3

port plug on Valve 252F007B will be adequately managed in accordance with existing station processes.

Configuration of Packing on 252F007B

In Reference 1, Enclosure 1, Section 4.2, Susquehanna stated the original design for Valve 252F007B included a dual packing set which was replaced with a single packing set in 1988. Additionally, in Enclosure 2 to Reference 1, Susquehanna provided the vendor drawing for Valve 252F007B. The drawing provided shows the original configuration of Valve 252F007B as procured from the vendor; i.e., that with a dual packing set. Enclosure 2 to this letter provides the valve data sheet for Valve 252F007B, which describes the current packing configuration for the valve; i.e., that with a single packing set.

Duration of Proposed Alternative

In Reference 1, Enclosure 1, Section 6, Susquehanna requested that 4RR-10 be applicable through the duration of Unit 2, Cycle 22. Unit 2, Cycle 22 is projected to end in spring 2025.

However, the fourth ISI interval for SSES, Unit 2, ends on May 31, 2024. The fifth ISI interval for SSES, Unit 2, will begin on June 1, 2024, and is expected to end on May 31, 2034. Because Unit 2, Cycle 22 operation (and by extension, the need for the proposed alternative) will span portions of the fourth and fifth ISI intervals, Susquehanna requests that the proposed alternative cover the remainder of the current fourth ISI interval and the portion of the fifth ISI interval commencing at the start of the interval (i.e., June 1, 2024) and ending with the end of Unit 2, Cycle 22 (projected to occur in spring 2025).

The table in Reference 1, Enclosure 1, Section 2 is reproduced here and updated to include the information for the fifth ISI interval:

Interval Section XI Interval Start Date Interval End Date Edition/Addenda Fourth 2007 Edition, through June 1, 2014 May 31, 2024 2008 addenda Fifth 2019 Edition June 1, 2024 May 31, 2034

References

1. Letter from Susquehanna to NRC, Relief Request 4RR-10 Relief from Code Seal Weld Requirement for Valve 252F007B (PLA-8061), dated April 10, 2023 (ADAMS Accession No. ML23100A128).

Enclosure 2 of PLA-8062

Valve Data Sheet for Valve 252F007B

Page 1 of2

Valve Packing Datasheet: 252F007B, Rev. 1

Facility: Susquehanna Work Order: Statas: Future-Unverified Unit: 02 Verified Date: Prepared By: Fred Cuny System: 251 Inatalled Date: Prepared Date: 12/10/2007 Make/Model/Size: ANCHOR DARLING (FLOWSERVE), ?, 12.0 Deler. CORE SPRAY lNJBCTION TO REACTOR VSL Location: Area/Elev: Jtn67, Actual Elev: 769, Col/Grid: NIA, Room: Il-516

DESIGNDATA ASJ'OUNDDATA Stem Dia. (+.DOS/-.010)(A): 2.25 (2 1/4) in p. Valve Potitioa:

Stuff. Box Dia. (+/-.010)(B): 3.25 (3 1/4) in !pi Stem Condition:

Stuff. Box Depth(+/-.031)(q: 7.125 (7 1/8) in p, Flat Washers: ()Yes-()No Port Deptll(N): 0in Live Loaded: ()Yes-()No Lantem Ri11g Helght(M): 0 in *. Spring Washers Per Stud:

Gland Stud Diameter(G): 0.75 F Leakoff:

(3/4) in ~

Nut Wrench Size(L): 1.25 (1 1/4) in ~ Gland Take-up: in Useable Gland Length(D): in ~ Gland Bolt / Nut Condition:

Number of Gland Studs(H): 2 ~ Glud Alignment:

Port Active: ()Yes~()No pi Actual Stem Friction: lbf Port Type: ~ Leak Rate:

Lined Gland: () Yes-()No ASLJ:FTDATA

2 Year Retorque: ()Yes* ()No Valve Position:

VALVEDATA Stem Condition:

Actuator Type: Manual Installed Configuration:

\\.* Operating Pressure: 1015 Flat Washers: () Yes-()No l.! Operating Temperature: 522 Live Loaded: () Yes-(}No Criticality: 2 Spring Washers Per Stud:

REPACK DATA (Future)/RETORQUE DATA (lutalled) Leak.oft': Glod Take-up: in Packing Work: Gland Bolt/ Nut Condition:

Packing Type / Material: Gratbil Gland Alignment:

Lower Bashing Height: 4.25 (4 114) in lfi Attual Stem Frlttion: lbf Upper Bulling Height: in ~ Leak Rate:

Packing Set Height: 3 in ffl ADDfflONAL FIELD DATA Spring Wuhen Per Stud: "'

Spring Confignntion: ffi§ Packing Torque As Left: 141.Z ft-lbs Flat Washers Per Stud: P. Retorq. Number of Flats: 1 LP"~-r Paeking Stress Minimum: psi 2 Year lletorque WO #:

Torque Minimum: ft-lbs Lined Gland Condition: ~

Friction Minimum: 5768 lbf COS-22-oA Packing Stress Preferred: 4000 psi.t><1 e, 7 --,!3 ~ ~I Torque Preferred: 95 ft-lbs Friction Preferred: 7210 lbf Signatures

Date:~

Verified Date:

SealPRQJ'M Copydgbt C 2011-2020 A.P. Services. a business unit of Curtiss-Wright Flow Control Company. All ltights Rmserml.

https://ppl.apsealpro.com/Pages/Config/PrintDatasheet.aspx?id=l 103638&d=20203 l 1310... 7/13/2020 Page 2 of2

Valve Packing Datasheet: 252F007B, Rev. 1

Facility: Susquehanna Work Order: Status: Future-Unverified Unit: 02 Verified Date: Prepared By: Fred Curry System: 251 Installed Date: Prepared Date: 12/10/2007 Make/Model/Size: ANCHOR DARLING (FLOWSERVE), ?, 12.0 Descn CORE SPRAY INJECTION TO REACTOR VSL Location: Area/Elev: Jtn67, Actual Elev: 769, Col/Grid: NIA. Room: 11-516

CONFIGURATION MATERIAU Configuration: YGGOYBY T e Cat Id/Part # Q BUSHING 0091217218 SET 0091217062 FLAT 0091219132 COMMENTS

Repack 04/88 woV74164. No live load and no flat washers.

Retorq 03/07 wo787450, Found signs of leakage.

Cleaned, verified nut movement and torqued to 119 ft-lbs. 0 flats, gland is straight. 0.938" Take-up.

Retorque 4/09 WO# 1024351, AsL 142 ft/lbs,no washer, 4 flats, I" Takeup (Updated 5/15/09 mds) l)Inspect pack area fur leakage/corrosion and record observations in action taken.

2)Verify gland nuts move freely by loosening nuts 1/2 turn, returning original position.

3)Retorque in stepts to 100%, 125%, and 150% of suggested value and record flats after each step.

4)Contact packing engineer fur direction if no movement is obtained at 150%. (04/08 file)

SealPRO' Copyright O 2011-2020 A.P. Services. a business unit of Curtiss-Wright Flow Control Company. All Righls Reserved.

https://ppl.apsealpro.com/Pages/Config/PrintDatasheet.aspx?id= 1103638&d=20203 l 1310... 7/13/2020