05000423/LER-2023-006-02, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications

Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications

ML24270A029
Person / Time
Site:	Millstone
Issue date:	09/26/2024
From:	O'Connor M Dominion Energy Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
24-160B LER 2023-006-02
Download: ML24270A029 (1)

LER-2023-006, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
4232023006R02 - NRC Website

text

Domini o n Energy Nucl ear Connecticut, Inc. Dom inion Millston e Power Sta tion 314 Ro pe Fe rry Road, W a te rfo rd, CT 06385 En e rgy Dominion Energy.co m

U. S. Nuclear Regulatory Commission Serial No. : 24-160B Attention: Document Control Desk MPS Lie/JP R2 Washington, DC 20555 Docket No.: 50-423 License No. : NPF-49 SEP 2 6 2024

DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 LICENSEE EVENT REPORT 2023-006-02

PRESSURIZER POWER OPERATED RELIEF VALVE FAILED TO OPEN DURING SURVEILLANCE TESTING RES UL TING IN A CONDITION PROHIBITED BY TECHNICAL SPECIFICATIONS

This letter forwards a Licensee Event Report (LER) 2023-006-02, documenting a condition that was discovered at Millstone Power Station Unit 3 (MPS3) on October 20, 2023. This LER is being submitted pursuant to 10 CFR 50. 73 (a)(2)(i)(B) as a condition prohibited by technical specifications.

This is Supplemental Licensee Event Report committed to in LER 2023 - 006-01.

There are no regulatory commitments contained in this letter or its enclosure.

Should you have any questions, please contact Ms. Lori Kelley at (860) 447-1791 x 6520.

Sincerely,

~~

Michael J. O'Connor Site Vice President - Millstone

Enclosure: LER 423/2023 - 006-02 Serial No. 24-1608 Docket No. 50-423 Licensee Event Report 2023-006-02 Page 2 of 2

cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road, Suite 102, King of Prussia, PA 19406-1415.

R. V. Guzman NRG Project Manager Millstone Units 2 and 3 U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738

NRG Senior Resident Inspector Millstone Power Station Serial No. 24-1608 Docket No. 50-423 Licensee Event Report 2023-006-02

ATTACHMENT

LICENSEE EVENT REPORT 2023-006-02

PRESSURIZER POWER OPERATED RELIEF VALVE FAILED TO OPEN DURING SURVEILLANCE TESTING RESULTING IN A CONDITION PROHIBITED BY TECHNICAL SPECIFICATIONS

MILLSTONE POWER STATION UNIT 3 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

NRG FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3160-0104 EXPIRES : 04/30/2027 (04-02-2024) Estimated burden pe, response to comply l"th this mandatory co l~on request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lesson s

.**~"""*,,,.~ LICENSEE EVENT REPORT (LER) learned are inCO!]lOfated Into the Hcen~ng process and fed back to indu stry. Se nd comments regarding burd en t i estimate to the FOIA, Library, a nd Informa tion Co llections Branch (T-6 A10M), U. S. Nuclear Regulatory

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~"...... at 0MB Office or Inform a tion and Reg ulatory Affa,s, (3150-0104), Attn: Desk Office, for the Nu clear Regulatory (See NUREG-1O22, R.3 for instruction and gu idan c e for completing this form Commission, 725 17th Street NW, W ashin gto n, DC 20503. Th e NRG m ay n o t con du c t or spo n so r, and a pe,son Is hlltrllwww m~ goy/[eading-[Il]idoc-~2llec!ioosi u[egs/sta!f/s[JO22i[JQ not requ ired to respon d to, a collection of lnfOl"mation unle ss the docu m ent requ esting or requiing the co llection displays a currently va ITd 0MB con tro l number.

1. Facility Name ~ 050 2. Docket Number 3. Page

Millstone Power Station - Unit 3 052 00423 1 OF 6 I

4. Title Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting In a Condition Pro hibited by Technical Specifications.

5. Event Date 6. LER Number 7. Report Date 8. Other Facilities In v olved Month Day Year Number No. 050

10 20 2023 2023 - 006 - 02 09 26 2024 Faclllty Name 052 Docket Number

9. Operating Mode 110. Power Level 4 000

11. This Rep o rt is Submitted Pursuant to the Requirements of 10 CFR §: (Check all that apply) 10 CFR Part 20 20.22O3(a)(2)(vi) 10 CFR Part 50 50. 73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 73.1200(a) 20.2201(b) 20.22O3(a)(3)(i) 50.36(c)(1)(i)(A) 50. 73(a)(2)(ii)(B) 50. 73(a)(2)(viii)(B) 73.1200(b) 20.2201(d) 20.22O3(a)(3)(ii) 50. 36(c)(1 )(ii)(A) 50. 73(a)(2)(iii) 50. 73(a)(2)(ix)(A) 73.1200(c) 2O.2203(a)(1) 2O.22O3(a)(4) 50.36(c)(2) 50. 73(a)(2)(iv)(A) 50. 73(a)(2)(x) 73.1200(d) 20.2203(a)(2)(i) 10 CFR Part 21 50.46(a)(3)(ii) 50. 73(a)(2)(v)(A) 10 CFR Part 73 73. 1200 (e) 20.2203(a)(2)(il) 21.2(c) 50.69(g) 50.73(a)(2)(v)(B) 73. 77(a)(1) 73.1200(f) 20.2203(a)(2)(iii) 50. 73(a)(2)(i)(A) 50. 73(a)(2)(v)(C) 73. 77(a)(2)(i) 73. 1200(9) 20.2203(a)(2)(iv) [Z] 50. 73(a)(2)(i)(B) 50. 73(a)(2)(v)(D) 73. 77(a)(2)(ii) 73.1200(h) 20.2203(a)(2)(v) 50. 73(a)(2)(i)(C) 50. 73(a)(2)(vii)

OTHER (Specify here, in abst ract. or NRC 366A).

12. Licensee Contact for this LER

Licensee Contact Phone Number (Include area code)

Lori Kelley, Manager Nuclear Station Emergency Preparedness and Licensing 860-447-1791 X 6520

13. Complete One Line for each Comp o nent Failure Described in this Report

Cause System Co mponent Manufacturer Reportable to IRIS Cause System Component Manufacturer Reportable to IRIS B AB PSV CROSBY y

14. Supplemental Report Expected Month Day Year 15. Expected Submission Date 0 No Yes (if yes, complete 15. Expected Submission Dale)

Abstract

On October 20, 2023 at 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />, with Millstone Power Station Unit 3 at O percent reactor power in Mode 4, with RCS pressure of 342 psia and cold leg temperature of 285 deg F, the 'B' pressurizer power operated relief valve (PORV),

3RCS

• PCV456, failed to stroke open upon demand during performance of surveillance testing. The failed PORV was replaced with a rebuilt PORV, and a new pilot solenoid operated valve (SOV). The direct cause of the 'B' PORV failure to stroke was determined to be a failed Stellite pilot solenoid operated valve. The cause analysis identified the SOV contained Stellite material, which was not in conformance with design. Subsequent leakage through the 'B' PORV throughout the operating cycle damaged the SOV top stem and valve ball assembly and prevented it from stroking during surveillance testing. This leakage was present from August 2022 until October 2023; therefore, it is reasonable that the 'B' PORV was inoperable for a period greater than allowed by Technical Specifications. Therefore, this report is being submitted pursuant to 10 CFR 50.73 (a)(2)(i)(B), as an operation or condition that was prohibited by the plant's Technical Specifications.

This licensee event report (LER) supplement provides the results of a completed cause analysis.

I

On October 20, 2023 at 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />, with Millstone Power Station Unit 3 (MPS3) at O percent reactor power in Mode 4, with RCS pressure of 342 psia and cold leg temperature of 285 deg F, the ' B' pressurizer power operated relief valve (PORV), 3RCS

• PCV456, failed to stroke open on demand during performance of surveillance testing per SP 3601B.2, "Tra in B Pressurizer Steam Space Vent Path and PORV Stroke Time Operability." The PORV was declared inoperable to support both normal operating pressure overp ressure protection function and its cold overpressure protection function in Modes 1-3. During plant cooldown, alternative means were utilized to satisfy the associated cold over p ressure protection requirements.

The MPS3 Pressurizer (PZR) PORV valve is a Crosby pilot solenoid operated (S/N K72047-00-0006) relief valve that relies on a pilot solenoid operated valve lifting to vent main body valve pressure and provide the necessary differential pressure across the main disc to lift. Extensive troubleshooting and offsite testing validated that the failed ' B' PORV was unable to stroke at both high and low pressure conditions. Further investigation identified the solenoid operated valve associated with the 'B' PORV contained Stellite mater ial that was not in conforma n ce with design. The ball was made of stellite, and upper seat was made of stainless steel. Industry ope rating ex perience had previously identified that Stellite internals were s usceptible to leakage when exposed to steam.

In January, 2001, Westinghouse provided a proposal to modify and refurbish several PORV solenoid operated valves (SOVs) to replace the Stellite ball and stainless steel upper seat with lnconel. In June, 2002, an engineering design change was approved and directed four SOVs to be refurbished by the vendor, Westinghouse, and replace the stellite ball and stainless steel upper seat with a ball and uppe r seat made of lnconel.

In 2002, the engineering design change was implemented, however, the material control processes were insufficient to differentiate between the original (Stellite) and modified SOVs. Specifically, material control processes did not require a new material number to be assigned to the modified SOVs. As a result, the modified SOVs did not have a unique material stoc k code number to differentiate them from the original valves. Using the same material stock-code prevented the ability to distinguish between preferred and non-preferred material when ordering parts. Also, the affected drawings and bill of materials (BOM) were not updated to reflect the new design of the MPS3 PORV SOVs.

In April 2005, MPS3 ex perienced an inadvertent safety injection actuation event which caused the pressurizer PORVs to cycle more than 40 times. Subsequent to the event, both the pressurizer PORVs were replaced with refurbished valves that had the lnconel ball and upper seat. Leakage testing of the removed 'B' PORV pilot SOV (S/N K72047-00-0006) determined that the valve was leaking. The SOV was removed and placed in MPS3 Fuel Building in a satellite quality assurance sto rage area as blocked stock. The SOV was never refurbished with the lnconel parts and was not entered back into the supply chain inventory trac king system for disposition.

Dur ing planning fo r PORV refurbishment in 2021, station and vendor personnel discussed SOV material, however, the question was not pursued, and verification or replacement of the ball and upper seat material was not included in the purchase order for SOV refurbishment. As a result, the incorrect material was not identified and corrected. At that time, there was not a viable replacement pilot SOV ava ilable to be used in support of PORV rebuild, therefore the bloc k stock SOV (S/N K72047-00-0006) from satellite QA storage was sent to Westinghou se to perform further analysis on leak tightness and pressure tests. Th is SOV was also sent to a test facility that could perform steam testing to validate leak tightness at plant conditions. The pilot SOV passed its functional test and seat leakage test before installing in 3RCS

• PCV456 during the 3R21 refueling outage in 2022.

I Plant heat-up commenced on May 18, 2022. The 'B' PORV started to show evidence of leakage on August 17, 2022. The leaking PORV was isolated by closing the upst ream block valve, 3RCS

• MV8000B, on August 19, 2022. Following the isolation of the 'B' PORV block valve, tempe ratures downstream of the 'B' PORV continued to rise. The leakage continued throughout the operating cycle and condition reports were generated to document the leakage past the isolation block valve and leaking 'B' PORV. The gross leakage caused steam erosion of the override top stem assembly and valve ball. This prevented the Stellite pilot solenoid operated valve (SOV) from lifting to provide the main body PORV the required pressure differential to stroke. On October 20, 2023, during shutdown for refueling outage 3R22, surveillance SP3601B.2, "Train B Pressurizer Steam Space Vent Path and PORV Stroke Time Ope rability" was performed to stroke the 'B ' PORV. The ' B' PORV failed to stroke open on demand.

The ' B' PORV was replaced with a rebuilt main valve and a new lnconel solenoid operated valve (SOV). The ' A' PORV, 3RCS

• PCV455A, was also replaced with a rebuilt main valve and a new lnconel SOV. The spare PORV had its SOV replaced with a new lnconel SOV. Both trains successfully passed all required post maintenance testing to support plant operation.

Technical Specification (TS) 3.4.4 requires that both power operated relief valves and their associated block valves shall be operable for Modes 1, 2, an d 3. The associated action (b) with one PORV inoperable due to causes othe r than excessive seat leakage is to either restore the PORV to OPERABLE status within 1 hou r or close the a ssociated block valve and remove power from the block valve(s); restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Technical Specification (TS) 3.4.9.3 requires that Cold Overpressure Protection shall be OPERABLE with a maximum of one centrifugal charging pump and no safety injection pump capable of injecting into the reactor coolant system (RCS) and with relief valve combinations i.e. both PORV, OR, Two residual heat removal (RHR) suction relief valves OR with one PORV and one RHR suct ion relief valve, OR by depressurizing RCS with an RCS vent greater than or equal to two squa re inches. TS 3.4.9.3 is applicable for Mode 4 (when reactor coolant system less than o r equal to 226 degrees F),

Mode 5, and Mode 6 (when the reactor head is on the reactor vessel)

Based upon failure of the 'B' PORV to stroke open on October 20, 2023, the ' B' PORV was declared inoperable to support its TS 3.4.4 design function in Modes 1 through 3. TS 3.4.9.3 was satisfied as the ' A' PORV, and two residual heat removal suction relief valves (RHR) were available and operable to relieve any pressure for Cold Overpressure Protection purposes.

Although, the exact date of failure could not be determined, with the leakage present from August 2022 until October 2023; it is reasonable that the ' B' PORV was inoperable for a period greater than allowed by Technical Spec ifications.

This report is being submitted pursuant to 10 CFR 50.73 (a)(2)(i)(B), as an operation or condition that was prohibited by the plant 's Technical Specifications.

CAUSE

The d irect cause of the 'B ' PORV failure to stroke was determined to be a failed Stellite pilot solenoid operated valve.

The steam leakage past the 'B' PORV block valve damaged the Stellite pilot SOV top stem and valve ball assembly.

Add itionally, Soleno id operated valves containing Stellite o r lnconel were not differentiated in the material control system, leading to a Stellite SOV being released to the field for installation.

ASSESSMENT OF SAFETY CONSEQUENCES

Final Safety Analysis Report (FSAR) Chapte r 15 was reviewed for the extent to which the PORVs are credited in the safety analysis.

Chapte r 15 Peak RCS Pressure and Core Response

The traditional criteria examined in the Chapter 15 safety analyses have been core Departure of Nucleate Boiling (DNB) response and RCS peak pressure threats. These have typically manifested themse lves close to the time of reactor trips.

The pressure relieving capabilities of the PORVs are not credited in the Chapter 15 safety analysis to limit primary system peak pressure to below event acceptance criteria. Those scenarios examined for peak primary system pressure assume that the PORVs are inoperable and pressure increases cont inue until mitigated by the action of the pressurizer safety valves (PSVs), if necessary. For the Chapter 15 scena rios examined for core DNB response, lower system pressures are more adverse as the lower pressure is adverse for the prediction of DNB. For analysis scenarios to examine the core response, the PORVs a re modelled to act so that the DNB calculations are performed at a lowe r primary system pressure. Having only one PORV act to lower RCS pressure would be benign for DNB response.

Event Escalation

At MPS3, the PORVs have been qualified to operate in liquid relief. In contrast, the PSVs have not been qualified, allowing one to postulate that liquid relief through the Pressurize r safety va lves (PSVs) could lead to RCS leakage du ring an event that does not in itially involve loss of primary coolant.

To address those concerns, the analys is of several events was performed to examine the post trip approach to pressurizer fill which could result in a water solid pressurizer and PSV liquid relief. Fo r the events most susceptible to a solid pressurizer, Time Critical Operator Actions (TCOAs) have been incorporated into the Emergency Ope rating Procedures (EOP) network to ensure that at least one PORV is unblocked and available to receive the postulated liquid relief, preventing liquid relief through the PSVs. A single PORV flow path has been demonstrated to be adequate to prevent liquid relief through the PSVs in this post trip period. Specifically, the FSAR 15.2.8 (Feedwater System Pipe Break), 15.5.1 (Inadvertent Safety Injection (IOECCS)), and 15.5.2 (CVCS Malfunction) analyses c red it the ava ilability of a single PORV to preclude event escalation. Therefore, the unaffected 'A' PORV would have been sufficient to prevent event escalation.

Low Temperature Overpressure Protection

With the 'B' PORV unable to stroke, the 'A' PORV and the RHR suction relief valves remained available to provide protection against an over pressurization event during low temperature operation. Analyses have shown that only one PORV or one RHR suction relief valve is sufficient to prevent violation of these limits due to anticipated low temperature mass and heat input transients. RHR was in operation at the time of the event and was the credited decay heat removal mechanism.

Anticipated Transient Without Scram (ATWS)

An assessment of the ATWS was performed in support of the Measurement Uncertainty Recapture (MUR) uprate. The limiting peak pressure ATWS, crediting thick metal masses, resulted in a peak RCS pressure of 3071 psia. The standard analysis assumptions for ATWS are that both PORVs are available for relief. The generic ATWS study, Reference 15.8-2 of the FSAR, included an examination of potential single failures from the 'reference' ATWS. For the Loss of Load ATWS, the failure of one PORV to open resulted in a 166 psia adder to the 'reference' peak pressure case. Adding this differential to the MUR results gives a peak Loss of Load ATWS RCS pressure of 3237 psia. Engineering has reviewed the analysis documented in MPS3 FSAR Reference 15.8-2 and determined that adequate margin exists to accommodate the peak RCS pressure of 3237 psia during the duration of the ATWS event.

Bleed and Feed

Emergency Operating Procedures (EOP) 35 FR-H.1 is the functional restoration guideline used for response to beyond design basis loss of secondary heat sink events. Modeling and simulation software (RELAP) cases were run to gain insights as to possible combinations of safety injection and charging pumps that may be successful. The combinations that produced possible successful recovery are included in the current Revision to EOP 35 FR - H.1. It was demonstrated that the unaffected 'A' PORV in combination with the reactor head vents provide adequate 'bleed' capacity for successful once through cooling. Instructions are included in the current EOP 35 FR-H.1 to unisolate and operate the head vents should insufficient relief be available from operable PORV(s).

Conclusion

Based on the review performed, the pressurizer PORVs are not credited for prevention of core damage in the FSAR Chapter 15 safety analyses. A single, available pressurizer PORV is credited in the Chapter 15 safety analyses to preclude the potential for the Feedwater System Pipe Break, Inadvertent operation of the Emergency Core Cooling System (IOECCS), and Chemical and Volume Control system (CVCS) malfunction events to escalate to a higher classification. Thus, having a single available PORV is sufficient to preclude core damage and event escalation with respect to the Chapter 15 safety analyses.

CORRECTIVE ACTIONS

The 'B' PORV was replaced with a rebuilt main valve with a new lnconel solenoid operated valve and successfully passed all required post maintenance testing to support plant operation. A unique material number will be assigned to solenoid operated valves that contain lnconel. Supply chain procedures will be updated to require new unique material numbers for items that have material changes, and all Stellite solenoid operated valves will be placed into blocked stock.

Additional corrective actions will be taken in accordance with the station's corrective actions program.

PREVIOUS OCCURANCES

There have been no similar events or conditions related to pressurizer PORVs being inoperable for a period longer than the technical specification action statement allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> at Millstone Power Station over the last 3 years.

ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES

AB Reactor coolant system PSV Valve, Solenoid, Pressure