PLA-8061, Relief Request 4RR-10 Relief from Code Seal Weld Requirement for Valve 25F007B, PLA-8061

Relief Request 4RR-10 Relief from Code Seal Weld Requirement for Valve 25F007B, PLA-8061

ML23100A128
Person / Time
Site:	Susquehanna
Issue date:	04/10/2023
From:	Casulli E Susquehanna, Talen Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
PLA-8061
Download: ML23100A128 (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 10, 2023 Attn: Document Control Desk 10 CFR 50.55a U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION RELIEF REQUEST 4RR-10 RELIEF FROM CODE SEAL WELD REQUIREMENT FOR VALVE 252F007B PLA-8061 Docket No. 50-388 In accordance with 10 CFR 50.55a(z)(2), Susquehanna Nuclear, LLC (Susquehanna), requests NRC approval of the attached relief request associated with the fourth Inservice Inspection (ISI) interval for the Susquehanna Steam Electric Station (SSES), Unit 2. The fourth interval of the SSES Unit 2 ISI Program is based on the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code (Code),Section XI, 2007 Edition through the 2008 Addenda.

The fourth ISI interval at SSES began on June 1, 2014, and is currently scheduled to end May 31, 2024.

The attached relief request, 4RR-10, is associated with repair of the SSES Unit 2 Core Spray Injection to Reactor Vessel Valve 252F007B. Specifically, 4RR-10 requests authorization of alternative requirements for the replacement of the threaded leakoff port plug in accordance with IWA-4000 of the ASME Code,Section XI.

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

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

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

E. Casulli

Document Control Desk PLA-8061

Enclosures:

1. Relief Request 4RR-10

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

to PLA-8061 Relief Request 4RR-10 to PLA-8061 Page 1 of 5

1.

ASME Code Component(s) Affected Susquehanna Steam Electric Station (SSES), Unit 2 Valve 252F007B, Core Spray Injection to Reactor Vessel Valve.

Valve 252F007B is an ASME Code,Section XI, Class 1 (Section III 1971 Edition, through the Winter 1972 Addenda) component, and the valve body is Stainless Steel SA351-CF8M.

The design temperature of the valve is 565°F and the design pressure of the valve is 1240 psig1.

The design pressure of the attached system is 1250 psig. The maximum valve operating temperature is 565°F with a maximum operating pressure of 1260 psig.

2.

Applicable Code Edition and Addenda

The Code of Construction is ASME,Section III, 1971 Edition, through the Winter 1972 Addenda.

For SSES Unit 2, the Inservice Inspection Code of Record and Interval Dates are:

Interval Section XI Edition/Addenda Interval Start Date Interval End Date Fourth 2007 Edition, through 2008 addenda June 1, 2014 May 31, 2024 10 CFR 50.55a(g)(4)(ii) requires plant Inservice Inspection (ISI) Programs to be updated every 120 months to the Edition of ASME Section XI approved for use by reference in 10 CFR 50.55a(b)(2) 18 months prior to the start of the next 120-month inspection interval.

Upon entry into the Fifth Interval on June 1, 2024, Susquehanna Nuclear, LLC (Susquehanna),

will adopt the 2019 Edition of ASME Section XI as approved by 10 CFR 50.55a(b)(2). For the purpose of this document, all references will be to the current Fourth Interval Section XI edition and addenda of record.

3.

Applicable Code Requirement

IWA-4131.1 of ASME Code,Section XI states, in part:

When repair/replacement activities involve the following items, the alternative requirements of IWA-4131.2 may be used.

1 Note that design temperature and pressure co-dependent and are inversely related to each other.

to PLA-8061 Page 2 of 5 (a) Class 1 piping, tubing (except heat exchanger tubing, and sleeves and plugs used for heat exchanger tubing), valves, fittings, and associated supports no larger than the smaller of (1) or (2) below (1) NPS 1 (DN25); or (2) the size and design such that, in the event of postulated failure during normal plant operating conditions, the reactor can be shut down and cooled in an orderly manner, assuming makeup is provided by normal reactor coolant makeup systems operable from on-site emergency power.

IWA-4131.2 of ASME Code,Section XI states, in part:

For repair/replacement activities involving items identified in IWA-4131.1, the following requirements may be used in lieu of those in IWA-4000.

(a) Items shall be procured in accordance with the requirements of IWA-4142 and the technical requirements of IWA-4200. For Section III items, the requirements of NA-3700 or NCA-3800 need not be met, provided the Owners Quality Assurance Program provides measures to assure that material is furnished in accordance with the material specification and the applicable material requirements of Section III.

IWA-4321 of ASME Code,Section XI states, in part:

(c) Threaded joints in which the threads provide the only seal shall not be used in Class 1 piping systems. If a seal weld is employed as the sealing medium, the stress analysis of the joint shall include the stresses in the weld resulting from the relative deflections of the mated parts.

4.

Reason for Request

In accordance with 10 CFR 50.55a(z)(2), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

4.1 Background

In September 2022, Susquehanna commenced a planned maintenance outage in order to identify and address the source of ongoing drywell unidentified leakage. During drywell walkdowns, Valve 244F103, Reactor Bottom Head Drain Bypass Valve, was found to be leaking from the to PLA-8061 Page 3 of 5 valve packing, and ultimately from the valve packing leakoff port threaded connection. During investigation of the leak, it was determined the leakoff port plug was not replaced with an approved pressure-retaining plug or seal welded to meet ASME Code,Section III requirements during a 1988 valve pack redesign that made the leakoff port plug a pressure-retaining component.

Extent of cause walkdowns performed during the ongoing Unit 2 Refueling and Inspection Outage (RIO) of the leak identified on Valve 244F103 during the September 2022 maintenance outage identified 14 additional valves in Unit 2 with a threaded connection that were not seal welded per the requirements of the ASME Code. All 14 valves were emergently scheduled to have code-compliant repairs installed during the ongoing Unit 2 RIO. The repair/replacement plan for Valve 252F007B entailed removing the existing leakoff port plug and replacing with a new leakoff port plug that can be seal welded into place.

4.2 Valve Details The vendor drawing of the valve is provided in Enclosure 2. The system piping and instrumentation diagram for Valve 252F007B is provided in the Updated Final Safety Analysis Report (FSAR) Figure 6.3-4.2 Valve 252F007B was originally designed with a dual packing set with one set of packing beneath the leakoff port and leakoff port plug. The packing beneath the leakoff port plug made the plug a non-pressure retaining component, and therefore the plug was supplied with commercial steel materials. In 1988, the packing design was changed in accordance with the valve packing program to a single set of packing with no packing beneath the leakoff port plug.

This change inadvertently made the leakoff port plug pressure retaining and therefore part of the ASME Class 1 reactor coolant pressure boundary. At that time, a design change should have been performed to replace the existing leakoff port plug with a new plug that could be seal welded in order to meet the ASME Code requirements.

4.3 Physical Hardship Due to the length of time the non-ASME compliant leakoff port plugs have been installed, Susquehanna has experienced significant difficulty in removing similar leakoff port plugs from other valves during the ongoing Unit 2 RIO. Work to remove the leakoff port plug from Valve 252F007B was planned for the ongoing Unit 2 RIO. Due to the proximity of Valve 252F007B to the reactor pressure vessel and size of the line, it is not possible to isolate Valve 252F007B from the vessel. Therefore, in the event of damage to the packing or repair requiring repacking the valve, the only possible method of preventing water leaking through Valve 252F007B is placing the disc on the backseat of the valve, which may not eliminate all 2 The Referenced FSAR figure is for the Unit 1 Core Spray System. This drawing is representative of the commensurate Unit 2 alignment. The commensurate Unit 1 valve number is 152F007B.

to PLA-8061 Page 4 of 5 leak-by. Thus, performance of this repair activity could require a full core offload and draining the reactor vessel water level below the Core Spray injection lines. This activity would significantly extend the ongoing Unit 2 Refueling Outage, and result in additional, otherwise unnecessary, fuel moves. These additional fuel moves present an increase in the opportunity of damaging fuel during the fuel moves, or inadvertently placing bundles in incorrect locations.

Therefore, deferral of the repair/replacement activity for one operating cycle would allow for additional planning and help reduce the risk of human performance errors during an emergent full core offload and vessel draindown.

5.

Proposed Alternative and Basis for Use In lieu of performing a repair during the ongoing Unit 2 RIO to replace the existing leakoff port plug with a new plug that can be seal welded, Susquehanna proposes to accept the current condition as-is for one additional operating cycle. This will allow time to develop an appropriate strategy that minimizes personnel and industrial risk utilizing existing work processes instead of attempting to perform a complicated modification on an expedited basis.

This configuration has existed in the plant since 1988, but was not discovered to be non-compliant with the ASME Code until the extent of cause review performed following the September 2022 maintenance outage. Since the need for an ASME Code-compliant repair was discovered, repairs were planned for 14 Unit 2 valves with this non-compliant configuration. In the time since the re-pack in 1988, there has been no history of leakage at the threaded connection to the leakoff port plug. The packing of Valve 252F007B has continued to be re-torqued per appropriate preventive maintenance activities, the most recent occurring in 2021.

Subsequent walkdowns during outages have not identified any degraded conditions or leakage from Valve 252F007B. Inspections during the ongoing Unit 2 RIO determined there is no indication of leakage. At this time, a repair plan may result in an unusual difficulty without a corresponding benefit in overall safety or quality.

The Repair Replacement section of ASME Code,Section XI, IWA-4131 provides alternative requirements for small items, e.g., items no larger than the smaller of NPS 1 or the size and design such that, in the event of postulated failure during normal plant operating conditions, the reactor can be shut down and cooled in an orderly manner, assuming makeup is provided from normal reactor coolant makeup systems operable for on-site emergency power. The leakoff port plug can be considered a small item in that it is smaller than NPS 1. Small items may be procured to alternative requirements in lieu of those of IWA-4000 provided the Owners Quality Assurance Program provides measures to assure the material is furnished in accordance with the material specification and the applicable material requirements of Section III. Being that the drawings for Valve 252F007B show no more information than that the plug is commercial steel, Susquehanna as the Owner cannot ensure that the plugs meet the material specification and fitting design.

to PLA-8061 Page 5 of 5 In summary, there exists (1) expansive operating history of Valve 252F007B in this configuration with no noted leakage; (2) an allowance within ASME Code,Section XI, Paragraph IWA-4131 to use alternative requirements when procuring parts as small as the leakoff port plug; and (3) a short duration of additional exposure when compared to the time already installed in the plant. Further, system leakage tests have been performed on the affected valve in accordance with Table IWB-2500-1 Examination Category B-P and will be performed during the ongoing Unit 2 RIO. During the operating cycle, Susquehanna monitors the drywell for sources of leakage per Surveillance Requirement 3.4.4.1 with the Technical Specifications requiring a unit shutdown if the leakage reaches unacceptable levels. Thus, Susquehanna concludes that the ASME Code repair does not materially increase the level of safety or quality of SSES, Unit 2. Therefore, the unusual difficulty associated with performing an emergent full core offload and vessel draindown to the Core Spray injection lines in support of an ASME Code-compliant repair/replacement activity during the ongoing Unit 2 Refueling Outage does not provide a compensating increase in the level of quality or safety.

6.

Duration of Proposed Alternative Use of this proposed alternative is applicable to Unit 2, Cycle 22 operation only. During this operating cycle, Susquehanna will develop a plan to permanently address the non-compliant threaded leakoff port plug installed in Valve 252F007B.

7.

Precedent

1. Letter from NRC to Exelon Nuclear, LaSalle Unit 2 - Verbal Authorization of LaSalle Unit 2 Relief Request I4R-12, Revision 1 re: Valve Repairs on Valves 2B33-F060A and 2B33-F060B, dated March 15, 2021 (ADAMS Accession No. ML21089A054).

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