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

Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station 314 Rope Ferry Road, Waterford, CT 06385 Dominion Energy.com U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 SEP 2 6 2024 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 LICENSEE EVENT REPORT 2023-006-02 Dominion Energy Serial No.:

24-160B MPS Lie/JP R2 Docket No.: 50-423 License No.: NPF-49 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

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 Page 2 of 2

ATTACHMENT Serial No. 24-1608 Docket No. 50-423 Licensee Event Report 2023-006-02 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 col~on request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons

.**~"""*,,,.~

LICENSEE EVENT REPORT (LER) learned are inCO!]lOfated Into the Hcen~ng process and fed back to industry. Send comments regarding burden t

i estimate to the FOIA, Library, and Information Collections Branch (T-6 A10M), U. S. Nuclear Regulatory

."/*:

(See Page 2 for required number of digits/characters for each block)

Commission, Washington, DC 20555-0001, or by email to lnfocollects.Resource@nrc.gov, and the 0MB reviewe,

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at 0MB Office or Information and Regulatory Affa,s, (3150-0104), Attn: Desk Office, for the Nuclear Regulatory (See NUREG-1O22, R.3 for instruction and guidance for completing this form Commission, 725 17th Street NW, Washington, DC 20503. The NRG may not conduct or sponsor, and a pe,son Is hlltrllwww m~ goy/[eading-[Il]idoc-~2llec!ioosiu[egs/sta!f/s[JO22i[JQ not required to respond to, a collection of lnfOl"mation unless the document requesting or requiing the collection displays a currently vaITd 0MB control 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 Prohibited by Technical Specifications.

5. Event Date

6. LER Number

7. Report Date

8. Other Facilities Involved Month Day Year Year Sequential Revision Month Day Year Faclllty Name Docket Number Number No.

050 10 20 2023 2023 -

006 -

02 09 26 2024 Faclllty Name Docket Number 052

9. Operating Mode 110. Power Level 4

000

11. This Repo 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 abstract. 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 Cause System Component Manufacturer Reportable to IRIS Cause System Component Manufacturer Reportable to IRIS B

AB PSV CROSBY y

14. Supplemental Report Expected Month Day Year 0

15. Expected Submission Date 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

2. DOCKET NUMBER

3. LER NUMBER I VEAR SEQUENTIAL REV NUMBER NO.

423

• I 2023 I-I I -0 006 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, "Train B Pressurizer Steam Space Vent Path and PORV Stroke Time Operability." The PORV was declared inoperable to support both normal operating pressure overpressure 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 pressure 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 material that was not in conforma nce with design. The ball was made of stellite, and upper seat was made of stainless steel. Industry operating experience had previously identified that Stellite internals were susceptible 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 upper 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 stock 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 experienced 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 storage area as blocked stock. The SOV was never refurbished with the lnconel parts and was not entered back into the supply chain inventory tracking system for disposition.

During planning for 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 available to be used in support of PORV rebuild, therefore the block stock SOV (S/N K72047-00-0006) from satellite QA storage was sent to Westinghouse to perform further analysis on leak tightness and pressure tests. This 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

2. DOCKET NUMBER

3. LER NUMBER 18 SEQUENTIAL REV NUMBER NO.

423

- I I -0 006 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 upstream block valve, 3RCS* MV8000B, on August 19, 2022. Following the isolation of the 'B' PORV block valve, temperatures 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 Operability" 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, and 3. The associated action (b) with one PORV inoperable due to causes other than excessive seat leakage is to either restore the PORV to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or close the associated 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 suction relief valve, OR by depressurizing RCS with an RCS vent greater than or equal to two square inches. TS 3.4.9.3 is applicable for Mode 4 (when reactor coolant system less than or 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 Specifications.

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

2. DOCKET NUMBER

3. LER NUMBER I

423 NUMBER NO.

I YEAR SEQUENTIAL REV

~-, 006 1-0 The direct 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.

Additionally, Solenoid operated valves containing Stellite or 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) Chapter 15 was reviewed for the extent to which the PORVs are credited in the safety analysis.

Chapter 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 themselves 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 continue until mitigated by the action of the pressurizer safety valves (PSVs), if necessary. For the Chapter 15 scenarios 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 are modelled to act so that the DNB calculations are performed at a lower 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 Pressurizer safety valves (PSVs) could lead to RCS leakage during an event that does not initially involve loss of primary coolant.

To address those concerns, the analysis 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. For the events most susceptible to a solid pressurizer, Time Critical Operator Actions (TCOAs) have been incorporated into the Emergency Operating 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 credit the availability 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

2. DOCKET NUMBER

3. LER NUMBER I

423 D

NUMBER NO.

I YEAR SEQUENTIAL REV

~-, 006 1-0 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 I

2. DOCKET NUMBER

3. LER NUMBER I YEAR SEQUENTIAL REV NUMBER NO.

423 120231 -I I -0 006 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 Page 6

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