IR 05000440/2021012
| ML21273A047 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 09/30/2021 |
| From: | Karla Stoedter NRC/RGN-III/DRS/EB2 |
| To: | Penfield R Energy Harbor Nuclear Corp |
| References | |
| IR 2021012 | |
| Download: ML21273A047 (24) | |
Text
SUBJECT:
PERRY NUCLEAR POWER PLANT - DESIGN BASIS ASSURANCE INSPECTION (PROGRAMS) INSPECTION REPORT 05000440/2021012
Dear Mr. Penfield:
On August 5, 2021, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Perry Nuclear Power Plant and discussed the results of this inspection with Ms. A. Zelaski, Plant General Manager (representing) and other members of your staff. The results of this inspection are documented in the enclosed report.
Three findings of very low safety significance (Green) are documented in this report. Three of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at Perry Nuclear Power Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; and the NRC Resident Inspector at Perry Nuclear Power Plant.
September 30, 2021 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Karla K. Stoedter, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 05000440 License No. NPF-58
Enclosure:
As stated
Inspection Report
Docket Number:
05000440
License Number:
Report Number:
Enterprise Identifier:
I-2021-012-0019
Licensee:
Energy Harbor Nuclear Corp.
Facility:
Perry Nuclear Power Plant
Location:
Perry, OH
Inspection Dates:
June 28, 2021 to August 05, 2021
Inspectors:
K. Barclay, Reactor Inspector
B. Daley, Senior Reactor Inspector
E. Sanchez Santiago, Senior Reactor Inspector
Approved By:
Karla K. Stoedter, Chief
Engineering Branch 2
Division of Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (programs) inspection at Perry Nuclear Power Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Perform Valve Stroke Time Testing in accordance with American Society of Mechanical Engineers Requirements Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000440/2021012-01 Open/Closed None (NPP)71111.21N.02 A self-revealed Green finding and associated non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, subsection 55a (f) was identified due to the licensee's failure to retest or declare reactor core isolation cooling valve 1E51-F019 inoperable, as required by the American Society of Mechanical Engineers (ASME) Code, following surveillance testing performed on July 24, 2018. Specifically, the licensee used an inappropriate method for calculating the valve stroke time which resulted in them incorrectly concluding the valve met the surveillance test acceptance criteria. The valve subsequently failed to open during next scheduled surveillance test performed on October 22, 2018.
Failure to Perform Surveillance Testing in accordance with the Required Frequency in the ASME OM Code Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000440/2021012-02 Open/Closed
[H.3] -
Change Management 71111.21N.02 The inspectors identified a finding and associated non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, subsection 55a (f) for the failure to perform stroke time testing on three reactor core isolation cooling valves in accordance with the frequency established in the ASME OM Code. Specifically, the licensee updated their probabilistic risk assessment (PRA) and did not update the inservice testing program to reflect changes in valve testing frequency caused by the PRA update. As a result, the licensee did not recognize the testing frequency for three RCIC valves, which also served as containment isolation valves, had changed from every two years to quarterly.
Failure to Establish a Program to Ensure Motor Operated Valves Continue to be Capable of Performing Their Design Basis Safety Functions Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000440/2021012-03 Open/Closed None (NPP)71111.21N.02 The inspectors identified a Green finding and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Subsection 55a(b)(3)(ii), for the licensees failure to establish a program that ensured the design basis safety functions of certain motor-operated valves (MOVs) would be met. Specifically, the licensee did not implement the Joint Owners Group MOV Periodic Verification Program in accordance with MPR-2524-A, Joint Owners Group (JOG) Motor Operated Valve Periodic Verification Program Summary, for reactor core isolation cooling steam supply isolation valves 1E51-F063 and 1E51-F064.
Additional Tracking Items
None.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
Starting on March 20, 2020, in response to the National Emergency declared by the President of the United States on the public health risks of the coronavirus (COVID-19), inspectors were directed to begin telework. In addition, regional baseline inspections were evaluated to determine if all or a portion of the objectives and requirements stated in the IP could be performed remotely. If the inspections could be performed remotely, they were conducted per the applicable IP. In some cases, portions of an IP were completed remotely and on site. The inspections documented below met the objectives and requirements for completion of the IP.
REACTOR SAFETY
71111.21N.02 - Design-Basis Capability of Power-Operated Valves Under 10 CFR 50.55a Requirements POV Review (IP Section 03)
For the valves listed below, the inspectors:
a.
Determined whether the sampled power operated valves (POVs) are being tested and maintained in accordance with NRC regulations along with the licensees commitments and/or licensing bases.
b.
Determined whether the sampled POVs are capable of performing their design-basis functions.
c.
Determined whether testing of the sampled POVs is adequate to demonstrate the capability of the POVs to perform their safety functions under design-basis conditions.
d.
Evaluated maintenance activities including a walkdown of the sampled POVs (if accessible).
(1)1B21-F028C, Main Steam Line 'C' Outboard Isolation Valve (2)1B21-F041B, Automatic Depressurization System Valve (3)1E22-F004, High Pressure Core Spray Injection Valve (4)1E22-F012, High Pressure Core Spray Pump Minimum Flow Valve (5)1E51-F019, Reactor Core Isolation Cooling Minimum Flow Valve (6)1E51-F063, Reactor Core Isolation Cooling Steam Supply Inboard Isolation Valve (7)1E12-F048A, Residual Heat Removal 'A' Heat Exchanger Bypass Valve (8)1P45-F130B, Essential Service Water Pump 'B' Discharge Valve
INSPECTION RESULTS
Failure to Perform Valve Stroke Time Testing in accordance with American Society of Mechanical Engineers Requirements Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000440/2021012-01 Open/Closed None (NPP)71111.21N.02 A self-revealed Green finding and associated non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, subsection 55a
- (f) was identified due to the licensee's failure to retest or declare reactor core isolation cooling valve 1E51-F019 inoperable, as required by the American Society of Mechanical Engineers (ASME) Code, following surveillance testing performed on July 24, 2018. Specifically, the licensee used an inappropriate method for calculating the valve stroke time which resulted in them incorrectly concluding the valve met the surveillance test acceptance criteria. The valve subsequently failed to open during next scheduled surveillance test performed on October 22, 2018.
Description:
On July 24, 2018, the licensee performed surveillance test SVI-E51-T2001, "RCIC Pump and Valve Operability Test," Revision 42, on Reactor Core Isolation Cooling Pump Minimum Flow Valve 1E51-F0019. During the test, the valve was slow to start opening when the valve's control switch was taken to the open direction. The delayed valve response resulted in the valve's stroke time being unsatisfactory when measured with a stop watch. However, SVI-E51-T2001, Section 2.0, "Precautions and Limitations," Step 11 allowed the licensee to use a computer point to calculate the stroke time. This section specifically stated, "If the stroke time of a valve is unsatisfactory when using a stopwatch, ICS data may be used, if available. If this is done, write "ICS" in the initial power block (along with the initials), and determine the stroke time per PDB-C0010." The inspectors noted the licensee implemented Step 11 of the procedure and obtained a satisfactory valve stroke time. As a result, the licensee took no further action to address the valve's slow stroke time. This licensee documented the valve's delayed opening in corrective action document CR-2018-06590, "RCIC Min Flow Valve slow to open during the performance of SVI-E51-T2001."
Subsequently, on October 22, 2018, when performing the aforementioned procedure, valve 1E51-F019 failed to operate when operations personnel attempted to open the valve using the control switch. Per the surveillance procedure, the valve was declared inoperable, and the operators attempted to operate the valve a second time with the same results. The red pen light did not illuminate, and ICS received no open signal during both attempts to open the valve. Troubleshooting identified mechanical binding for the "open" contactor within the motor control center bucket. The licensee performed an apparent cause evaluation and determined the valve failure was due to a less than adequate preventive maintenance frequency that would lubricate the moving parts of the motor control center bucket. The licensee also determined that early identification of the valve's performance degradation did not occur because the use of ICS data to determine the valve's stroke time masked potential valve performance issues and performance trending for this valve was less than adequate.
During the inspection, the inspectors reviewed the surveillance procedure to determine whether the timing methodology used when the stopwatch stroke time was slow complied with regulatory requirements. The inspectors identified that using the ICS data as discussed in the procedure did not meet the requirements documented in the ASME OM Code.
Specifically ASME OM Code 2001, "Code for Operation and Maintenance of Nuclear Power Plants," Subsection ISTC, "Inservice Testing of Valves in Light Water Reactor Nuclear Power Plants," defined full stroke time as the time interval from the initiation of the actuating signal to the indication of the end of the operating stroke. The inspectors determined the ICS value measured the time the valve actually moved but did not capture the time that elapsed between the initiation signal and the beginning of valve movement. To account for the time ICS is unable to capture, the licensee applied a correction factor. The inspectors reviewed the basis for the correction factor and concluded the correction factor did not represent the actual time from initiation of the actuation signal to the start of valve movement. As a result, the use of ICS data to calculate valve stroke time could not be used to demonstrate compliance with the ASME Code inservice testing requirements.
For instances when the acceptance criteria is not met, ASME OM Code Subsection ISTC-5122, "Stroke Acceptance Criteria," stated, in part, "test results shall be compared to the initial reference values or reference values established in accordance with ISTC-3310 or ISTC-3320. ISTC-5123, "Stroke Test Corrective Action," paragraph (b), stated, in part, "valves with measured strokes times that do not meet the acceptance criteria of ISTC-5122 shall be immediately retested or declared inoperable. During the July 24, 2018, test, the licensee did not perform a retest or declare the valve inoperable in accordance with the aforementioned ASME OM Code requirements. This caused the licensee to fail to recognize a degraded condition that resulted in a subsequent failure of the valve to operate.
The inspectors communicated the failure to meet the ASME OM Code requirements to the licensee. They entered the issue into their corrective action program. The licensee had previously updated their procedure to eliminate the option to use an alternate method for calculating stroke times.
Corrective Actions: The licensee revised procedure SVI-E51-T2001, "RCIC Pump and Valve Operability Test," to eliminate Section 2.0 Step 11, which is the step that described the option to use an alternate method to calculate stoke times. The licensee was also performing a review to determine if additional actions needed to be taken as a result of the violation identified by the inspectors.
Corrective Action References: CR-2021-05197, "2021 DBAI POV Inspection:E51F0016 Stroke Time Testing Method for SVI-E51-T2001" CR-2018-06590, "RCIC Min Flow Valve slow to open during the performance of SVI-E51-T2001" CR-2018-09241, "E51-F0019 RCIC minimum flow valve failed to operate during SVI"
Performance Assessment:
Performance Deficiency: The licensee's failure to retest or declare reactor core isolation cooling valve 1E51-F019 inoperable after the stroke time exceeded the acceptance criteria was contrary to the requirements of 10 CFR 50.55a and the ASME OM Code. Specifically, the licensee did not use the correct stroke time value when comparing the test results to the acceptance criteria due to incorrect surveillance procedure guidance.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to use the correct stroke time value caused the licensee to fail to take the required actions of either retesting the valve or declaring it inoperable when the acceptance criteria was not met. This slow stroke time represented a degraded condition which was not addressed until the valve failed surveillance testing in October 2018.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2, "Mitigating Systems," the finding screened as having very low safety significance (Green)because the finding did not affect the design or qualification of a mitigating structure, system or component, did not represent a loss of a probabilistic risk assessment (PRA) function of a single train technical specification (TS) system for greater than its TS allowed outage time, did not represent a loss of the PRA function of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, did not represent a loss of a PRA system and/or function as defined in the plant risk information book or the licensee's PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and did not represent a loss of the PRA function of one or more non-TS trains of equipment designated as risk-significant in accordance with the licensee's maintenance rule program for greater than 3 days.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Title 10 CFR 50.55a Section (f), Preservice and Inservice Testing Requirements stated, in part, that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and inservice testing (referred to in this paragraph
ASME OM Code 2001, Code for Operation and Maintenance of Nuclear Power Plants, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, ISTC-2000 Supplemental Definitions, defines full-stroke time as the time interval from the initiation of the actuating signal to the indication of the end of the operating stoke.
ISTC-5122, Stoke Test Acceptance Criteria, states, in part, test results shall be compared to the initial reference values or reference values established in accordance with ISTC-3300, ISTC-3310, or ISTC-3320. ISTC-5122 (b), states, valves with reference stroke times of less than or equal to 10 seconds shall exhibit no more than a plus or minus 25 percent, nor plus or minus a 1 second change in stoke time, whichever is greater, when compared to the reference value.
ISTC-5123, Stroke Test Corrective Action, paragraph (b), states, in part, valves with measured stroke times that do not meet acceptance criteria of ISTC-5122 shall be immediately retested or declared inoperable.
Contrary to the above, on July 24, 2018, the licensee failed to meet the requirements for inservice testing of the ASME OM Code. Specifically, after valve 1E51-F019 failed to meet the acceptance criteria of ISTC-5122, the licensee failed to immediately retest it or declare it inoperable.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Perform Surveillance Testing in accordance with the Required Frequency in the ASME OM Code Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000440/2021012-02 Open/Closed
[H.3] - Change Management 71111.21N.02 The inspectors identified a finding and associated non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, subsection 55a
- (f) for the failure to perform stroke time testing on three reactor core isolation cooling valves in accordance with the frequency established in the ASME OM Code. Specifically, the licensee updated their PRA and did not update the inservice testing program to reflect changes in valve testing frequency caused by the PRA update. As a result, the licensee did not recognize the testing frequency for three RCIC valves, which also served as containment isolation valves, had changed from every two years to quarterly.
Description:
During a review of completed surveillances for valve 1E51-F063, "Reactor Core Isolation Cooling Inboard Steam Supply Isolation Valve," the inspectors identified the licensee was not performing the surveillances at the correct frequency. In accordance with the latest revision of the licensee's PRA program completed on January 15th, 2020, this valve was classified as medium risk significance. Per the licensee's inservice testing program, all motor operated valves (MOVs) classified as medium and high PRA risk significant will be categorized as HSSC (high safety significant) to implement ASME OM Code 2012, Mandatory Appendix III, Subsection III-3720 requirements. Subparagraph III-3620(a) required consideration of more frequent test intervals for MOVs with high safety significance. The licensee's inservice test program implemented the Mandatory Appendix III, Subsection III-3720 requirements by assigning a quarterly surveillance testing frequency interval to HSSC valves. However, valve 1E51-F063 was being tested on a 24-month frequency.
The licensee determined the cause for failing to perform the surveillance testing at the correct frequency was because they failed to update their inservice testing program after the PRA was updated in January 2020. The licensee performed an extent of condition review and identified two additional valves with an incorrect surveillance frequency; 1E51-F064, "Reactor Core Isolation Cooling Steam Supply Outboard Isolation Valve," and 1E51-F068, "Reactor Core Isolation Cooling Turbine Exhaust Isolation Valve." All three valves were being stroke time tested at a 24-month frequency instead of quarterly. Valves 1E51-F063 and 1E51-F064 were last tested on April 1, 2021, and were within the quarterly surveillance test frequency when the inspectors identified this issue. However, the inspectors found previous tests for both valves were performed at greater than a quarterly frequency. The inspectors determined valve 1E51-F068 was last tested on January 21, 2020. The licensee immediately performed a stroke time test of 1E51-F068, verified the complete test for 1E51-F063 and 1E51-F064 had been performed, and determined all three valves were operable. The licensee entered this issue into their corrective action program and corrected the testing interval for the aforementioned valves.
Corrective Actions: The licensee performed stroke time testing of the valves that were outside of their required frequency. All impacted valves were determined to be operable based on successful completion of the surveillance requirements.
Corrective Action References: CR-2021-05254, "2021 DBAI POV Inspection: Valves 1E51F0063, 1E51F0064, 1E51F0068 Exercise and Stroke Time Requirements did not change from PRA update in 2020"
Performance Assessment:
Performance Deficiency: The licensee failed to update the valve test table in their inservice testing program plan when changes were made to the significance of the valves in the program. This resulted in the licensee failing to meet the testing frequency requirements of the ASME OM Code. Specifically, valves E51-F063, E51-F063 and E51-F068 were not tested on a quarterly basis for four to five consecutive quarters.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee missed surveillance testing of three valves for four to five consecutive quarters. These valves were classified as high safety significance, therefore the failure to perform the surveillance testing at the correct periodicity affected the reasonable assurance the safety significant valves would perform their safety function.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 3, "Barrier Integrity," the finding screened as having very low safety significance (Green)because the finding did not represent an actual open pathway in the physical integrity of reactor containment; did not represent the failure of containment pressure control equipment; did not represent the failure of containment heat removal components; and did not involve the actual reduction in function of hydrogen igniters in the reactor containment.
Cross-Cutting Aspect: H.3 - Change Management: Leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
Specifically, when the licensee made changes to their PRA, they did not used a systematic approach to ensure the inservice testing program was updated as appropriate.
Enforcement:
Violation: Title 10 CFR 50.55a, Section (f), Preservice and Inservice Testing Requirements, states, in part, that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and inservice testing (referred to in this paragraph
ASME OM Code 2012,"Code for Operation and Maintenance of Nuclear Power Plants,"
Mandatory Appendix III, "Preservice and Inservice Testing of Active Electric Motor Operated Valve Assemblies in Light-Water Reactor Power Plants," Section III-3721, "HSSC MOVs,"
stated, HSSC MOVs shall be tested in accordance with paragraph III-3300. Section III-3600, "MOV Exercising Requirements," states, HSSC MOVs that can be operated during plant operation shall be exercised quarterly, unless the potential increase in core damage frequency and large early release associated with a longer exercise interval is small.
Contrary to the above, from January 2020, the licensee failed to meet the requirements for inservice testing of the ASME OM Code. Specifically, for HSSC valves E51-F063 and E51-F064, the licensee performed an exercise test on January 31, 2020, but did not exercise these valves again until April 8, 2021, which exceeded the quarterly testing requirement. For HSSC valve E51-F068, the licensee performed an exercise test on January 31, 2020, but did not exercise this valve again until July 9, 2021, which exceeded the quarterly testing requirement.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Establish a Program to Ensure Motor Operated Valves Continue to be Capable of Performing Their Design Basis Safety Functions Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000440/2021012-03 Open/Closed None (NPP)71111.21N.02 The inspectors identified a Green finding and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Subsection 55a(b)(3)(ii), for the licensees failure to establish a program that ensured the design basis safety functions of certain motor-operated valves (MOVs) would be met. Specifically, the licensee did not implement the Joint Owners Group MOV Periodic Verification Program in accordance with MPR-2524-A, Joint Owners Group (JOG) Motor Operated Valve Periodic Verification Program Summary, for reactor core isolation cooling steam supply isolation valves 1E51-F063 and 1E51-F064.
Description:
During the inspection, the inspectors reviewed Calculation MOVC-0076, Revision 0, which performed an Electric Power Research Institute (EPRI) Performance Prediction Methodology (PPM) Evaluation for valves 1E51-F063 and 1E51-F064 to determine the predicted thrust necessary to close the valve under steam blowdown conditions. The inspectors found that the EPRI PPM software had issued a Type 1 damage warning because the disk seat design radius for both valves did not meet the minimum EPRI PPM default radius.
EPRI TR-103237-R1-T1 (Non-Proprietary Version), "EPRI MOV Performance Prediction Program," states the following about a Type 1 Warning, The conditions at the interface are likely to produce material damage, and the friction coefficient cannot be reliably bounded. Although thrust results are generated by the model to carry the calculation to completion, the valve is considered to be unpredictable at all subsequent stroke positions. Type 1 warnings are only generated for contacts involving a Stellite sharp edge above the appropriate load thresholds.
The licensee adjusted the disk seat design radius inputs for both valves to meet the minimum EPRI PPM default radius and clear the errors. The licensee credited valve 1E51-F063 with the minimum radius because it had previously closed against a reduced differential pressure during performance testing conducted in October 1987. The licensees review of EPRI TR-103255, "July 1994 EPRI Gate Valve Design Effects Testing Results Report," found that stroking an MOV under reduced differential pressure and flow conditions will tend to chamfer the sharp edges. The inspectors review did not identify a similar reduced differential pressure test for valve 1E51-F0064, and Calculation MOVC-0076 provided no basis for changing the disk seat design radius inputs to clear the Type 1 error.
TR-103244, "EPRI MOV PPM Implementation Guide," Revision 1, which incorporated the Agencys Safety Evaluation comments for the EPRI PPM program, provided instructions for handling a Type 1 damage warning error. Specifically, TR-103244 provided corrective options which included performing a design basis test of the valve or modifying the valve and increasing the chamfer or radii. Crediting a past valve cycling with rounding of the sharp edges was not listed as an acceptable option for correcting a Type 1 error.
The inspectors found the unpredictable EPRI MOV PPM results were carried forward into the JOG MOV program for gate valves 1E51-F063 and 1E51-F064. Specifically, MPR-2524-A, Chapter 7, Implementation of JOG MOV Periodic Verification Approach, Step 5, Coefficient of Friction Threshold Screen, evaluates to determine if the valve is set up using a coefficient of friction (COF) that is susceptible to increase. For each combination of disk-to-seat material and fluid type tested in the JOG Program, a threshold COF is determined. The threshold is the value above which the COF does not increase, based on testing in the JOG MOV PV Program. Valves that are set up (and margin evaluated) using a COF greater than or equal to the applicable threshold COF are not susceptible to degradation and can be classified as a Class A or B, subject to inputs in previous steps. Valves that are set up using a COF less than the applicable threshold COF are susceptible to increases and are classified as Class C.
Perry used the unpredictable results from the EPRI PPM evaluation to calculate a COF and justify a Class A classification for 1E51-F063 and 1E51-F064. The inspectors, in consultation with the Office of Nuclear Reactor Regulation (NRR) mechanical engineers and MOV specialists, concluded the Class A classification was not appropriate because the valves did not have a valid COF. Specifically, the valves had never been tested at the design differential pressure and the EPRI PPM results were unpredictable and not a valid input for the JOG MOV PV Program.
The licensee performed an operability determination and concluded valves 1E51-F063 and 1E51-F064 remained capable of closing during a high energy line break event. Specifically, the licensee used engineering judgment to determine component wear for the period of time the valves have been installed and informed their conclusions by using wear chamfering data gathered during a 2003 valve maintenance evolution on 1E51-F064. The inspectors did not identify any deficiencies associated with the overall conclusions of the operability determination.
Corrective Actions: The licensee entered the issue into their corrective action program, performed an operability determination, and created corrective action assignments to inspect and / or remove sharp edges from 1E51-F063 and 1E51-F064 during their next refueling outage.
Corrective Action References: CR-2021-05435, "2021 DBAI POV Inspection: JOG Classification of RCIC Steam Supply Isolation Valves" CR-2021-05408, "2021 DBAI POV Inspection: RCIC Steam Supply Isolation Valves Questionable Calculation Justification"
Performance Assessment:
Performance Deficiency: The licensees use of unpredictable EPRI PPM results for the RCIC steam supply isolation valves was contrary to the JOG MOV PV Program and was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inspectors concluded that there was reasonable doubt regarding the availability, reliability or capability of the RCIC valves because the licensee was required to use a different approach while performing the operability determination. The inspectors compared the finding with the examples listed in IMC 0612, Power Reactor Inspection Reports, Appendix E, Example of Minor Issues. Example 3.m was found to be similar in that the licensee was not applying justified assumptions when performing a calculation. The example concluded that the reasonable doubt standard was met, when, through the process of completing an operability determination, the licensee used a different approach because the original approach resulted in unfavorable margin. For this finding, Perry needed to use engineering judgement associated with valve wear, instead of the original valve dimensions, until maintenance can be performed during the next refueling outage.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2, "Mitigating Systems," the finding screened as having very low safety significance (Green)because the finding did not affect the design or qualification of a mitigating SSC, did not represent a loss of the PRA function of a single train TS system for greater than its TS allowed outage time, did not represent a loss of the PRA function of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, did not represent a loss of a PRA system and/or function as defined in the PRIB or the licensee's PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and did not represent a loss of the PRA function of one or more non TS trains of equipment designated as risk-significant in accordance with the licensee's maintenance rule program for greater than 3 days.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Title 10 of the Code of Federal Regulations (10 CFR), subsection 50.55a(b)(3)(ii)required, in part, that licensees must establish a program to ensure that MOVs continue to be capable of performing their design basis safety functions.
Contrary to the above, as of July 16, 2021, the licensee did not establish a program that ensured MOVs were capable of performing their design basis safety functions. Specifically, the licensee did not follow the requirements of the Joint Owners Group MOV Periodic Verification Program to properly classify RCIC Valves 1E51-F063 and 1E51-F064, which have a design basis safety function to close in the event of a high energy line break on the steam supply line to the RCIC turbine. Proper valve classification is necessary to determine the appropriate test type and frequency to ensure the valves are capable of performing all their design basis safety functions. The licensee classified the valve as Class A when they did not meet the criteria for this classification.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On August 5, 2021, the inspectors presented the design basis assurance inspection (programs) inspection results to Ms. A. Zelaski, Plant General Manager (representing)and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2-0030-00000
Environmental Conditions for Drywell Area
K
93C1805.06-01
Evaluation of Seat Ring for Valves 1E12-F0048A and B
for Indicated Overthrust Condition
AOV-8
Determination of Valve Required Thrust, Actuator
Capability and Margin for AOVs: 1B21F0028A, B, C & D
B21-008
B21-025
Inboard/Outboard MSIV Stroke Time Acceptance Criteria
CL-MOV-1E51-1
MOV 1E51-F063/F064 Maximum Differential Pressure
CLMOV-3
Generic Letter 89-10 Program
E51-028
RCIC Pump Minimum Flow Requirements
2/08/1999
EQ-008
Limitorque D.C. Actuator-EQ Life Analysis
EQ-030
Qualified Life at 145 Degrees F Continuous in Drywell
& 1 A-01
EQ-128
MSIV NAMCO Limit Switch EQ Life Analysis
EQ-168
Extend Limitorque Containment Qualification to 180 Days
& 1 A-01
FSPC-0019
Division 2 AC Motor Operated Valve (MOV) Power Fuse
Sizing
FSPC-0019 A-04
Division 2 MOV Power Fuse Determination
MOV1E51-03
1E51-F010, F013, F019, F022, F031, F059, F068, F077 &
F078 Maximum Differential Pressure
2, 2.5 & 2.7
MOVC-0029
DC MOV Torque Capability Using Limitorque and BWR
Owner's Group Methodology
& 5
Addendum 2
MOVC-0044
Required Thrust Calculation for Globe Motor Operated
Valves (MOVs)
2/19/2001
MOVC-0063
Required Torque/Thrust Calculation for Edward
(Rockwell) Globe Valves
MOVC-0076
EPRI PPM Evaluation of Borg-Warner 10" ANSI Class
1500 Flex Wedge Gate Valves 1E51F0063/64
& 0.1
PSTG-0001
PNPP Class 1E Power Distribution System Voltage Study
SQ-0015
Qualification of Valves 1E51-F063, 1E51-F064
1, 1 A-01,
DCC 2 &
DCC 3
71111.21N.02
Calculations
SQ-0049
Motor Operated Valve 1E51-F0019 Qualification
2, 2.1, 2.2 &
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
A-01
TAF-81565 Scan
Pressure Locking and thermal Binding of Gate Valves
2/05/2000
TAF-81565 Scan
Pressure Locking and Thermal Binding of Gate Valves
2/23/1993
CR 2012-01789
Two MSIV Failed Stroke Time Testing
03/13/2021
CR 2017-02721
MSIV Stroke per SVI
03/11/2017
CR 2017-06730
Updated Industry Guidance to Address Part 21 Issue with
Anchor Darling Double Disc Gate Valves
06/20/2017
CR 2018-05939
NRC EQ DBAI: Accounting for Heat Rise in EQ
Component Evaluation
06/28/2018
CR 2018-10269
Reactor Core Isolation Cooling Mitigating Systems
Performance Index Basis Document & Probabilistic Risk
Analysis Model Inaccuracy
11/16/2018
CR 2019-02398
MSIV Stroke Times Unsat
03/16/2019
CR 2020-07503
Level 4 Worker Protection Event CP-4.1.3 Clearance
Development Milestone not Met
09/27/2020
Corrective Action
Documents
CR 2021-02297
SBI-B21-T9000 Acceptance Criteria
03/27/2021
CR 2021-05077
21 DBAI POV Inspection: Grease Spot from Valve
Lubricant on Insulation
06/30/2021
CR 2021-05097
21 DBAI POV Inspection: NRC Walkdown - Proximity
to Handwheel for 1E12F0048A to Scaffold Pole
06/30/2021
CR 2021-05151
21 DBAI POV Inspection: NOP-ER-3204 Misalignment
with ISTP for Perry
07/06/2021
CR 2021-05169
21 DBAI POV Inspection: Corrective Maintenance
Documentation for MSIV Stroke Times from 2017
07/07/2021
CR 2021-05197
21 DBAI POV Inspection: E51F0019 Stroke Time
Testing Method for SVI-E51-T2001
07/07/2021
CR 2021-05254
21 DBAI POV Inspection: Valves 1E51F0063,
1E51F0064, 1E51F0068 Exercise and Stroke Time
Requirements did not Change from PRA Update in 2020
07/09/2021
CR 2021-05375
21 DBAI POV Inspection: Calculation SQ-0049 Normal
Close Thrust Value for Diaphragm Disk
07/14/2021
Corrective Action
Documents
Resulting from
Inspection
CR 2021-05408
21 DBAI POV Inspection: RCIC Steam Supply Isolation
07/15/2021
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Valves Questionable Calculation Justification
CR 2021-05410
21 DBAI POV Inspection: PSTG-0001 not Updated
from PERP that Change MOV Motor
07/15/2021
CR 2021-05435
The Basis for Classification of the RCIC Steam Supply
Isolation Valves 1E51F0063 (Inboard) and 1E51F0064
(Outboard) for the Joint Owners Group Program on Motor
Operated Valve Periodic Verification (JOG) will Need to
be Reviewed against the Resolution of the Calculation
Issue Documented by CR 2021-05408 2021 DBAI POV
Inspection: RCIC Steam Supply Isolation Valves
Questionable Calculation Justification
07/16/2021
CR 2021-5067
21 DBAI POV Inspection: Housekeeping Deficiencies in
06/30/2021
2-0004-00000
Environmental Conditions for Auxiliary Building
L
206-0051-00000
Class 1E DC System
DDD
208-0065-00014
High Pressure Core Spray System Pump Injection Shutoff
MOV F004
R
208-0075-00016
Reactor Core Isolation Cooling System Minimum Flow to
Suppression Pool MOV F0019
U
208-0075-00024
Reactor Core Iso Cooling System Stm Sply Line Iso Inbrd
Valve F063
235-0075-00003A
MOV Data Sheet - 1E51F019 Min. Flow to Suppression
Pool
G
235-0075-00003B
MOV Data Sheet 1E51-F0019
235-0075-00009A
MOV Data Sheet - 1E51F063 RCIC Steam Supply
Inboard Isolation Valve
L
2-0631-00000
Reactor Core Isolation Cooling System
HH
2-0632-00000
Reactor Core Isolation Cooling System
40-0426-00003
Motor Operated Valve Assembly Gate 10 inch,
1500 LB-CS
A
4549-40-1047
Sheet 7
2" 900# Globe Valve Assembly
1A
Drawings
4549-40-1047
Sheet 8
Valve Assembly
0A
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
4549-40-1049
Sheet 1
Valve Assembly
A
DCC 01 SQ-0049
Design Control Change to SQ-0049: Motor Operated
Valve 1E51-F0019 Qualification, R 2
07/09/1996
11-0328-000
Replacement of the Carbon Steel 1E51F0019 RCIC
Minimum Flow Line Valve with a Stainless Steel Valve
11-0328-001
Replacement of the Carbon Steel 1E51F0019 RCIC
Minimum Flow Line Valve with a Stainless Steel Valve
DCC 02 SQ-0049
Design Control Change to SQ-0049: Motor Operated
Valve 1E51-F0019 Qualification, Revision 2
10/04/1997
MOV-3 DCC-008
Generic Letter 89-10 Program
04/07/1999
Engineering
Changes
PERP 478
Replacement of 80 ftlbs Start, 3400 rpm, 210 frame AC
Electric Motor for SB-2 Valve Actuator
Engineering
Evaluations
600787197
Motor Operated Valve JOG Program Basis
09/21/2012
MIDAS MOV Software Report 1E51F0063
03/18/2021
MIDAS MOV Software Report 1E51F0019 - Pre-Test
Setup for WO 200644972
03/19/2017
MIDAS MOV Software Report 1E51F0063 - Pre-Test
Setup for WO 200464450
03/20/2021
MIDAS MOV Software Report 1E51F0019 - Thrust and
Torque Calculation
01/15/2021
MIDAS MOV Software Report 1E51F0063 - Thrust and
Torque Calculation
01/15/2021
NRC Safety Evaluation for EPRI Topical Report
TR-103237 (Revision 1)
03/15/1996
MOV Test Report 1E51F0063-10
MOV Test Report 1E51F0019-006
MOV Test Report 1E51F0019-005
MOV Report 1E51F0063-009
MIDAS MOV Software Report 1E51F0019
03/18/2017
10-2917
Vendor Manual 0425-100E Model 952 Contactor
L
1E22F004-008
MOV Report
Miscellaneous
Change Notice
Revision to the Technical Specification Bases for
03/30/2021
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
21-023
SR 3.6.1.3.10
Change Request
03-051
Modification to TS Bases 3.6.1.3.10 when the 100 scfh
and 25 sfch Limits are Applied
05/24/2003
Correspondence
Transmittal of the Non-Proprietary Versions of All NRC
Staff Comments and EPRI Responses on the EPRI MOV
Performance Prediction Methodology
04/05/1996
EPRI TR-103237-
R1-T1
EPRI MOV Performance Prediction Program Topical
Report (Non-Proprietary Version)
EPRI TR-103244-
R1
EPRI MOV Performance Prediction Program
Implementation Guide
EPRI TR-103255
EPRI Performance Prediction Program - Gate Valve
Design Effects Testing Results
07/1994
ISTP R-24
Pump and Valve Inservice Testing Program Plan
NORM-ER-3601
Valve - Motor Operated
& 10
NORM-ER-3602
Valve - Air Operated
OE-2015-1305
IN 15-13, Main Steam Isolation Valve Failure Events
2/02/2016
OE-2017-03
Evaluation of NRC IN 2017-03
07/01/2017
PY-CEI/NRR-
0904 L
Perry Nuclear Power Plant Annual Report of
CFR 50.59 Safety Evaluations for 1987
08/31/1988
RAD 21-00488
Type C Local Leak Rate Test of 1B21 MSL Penetrations
03/29/2021
Revision
Notification
601194675
PM Revision for RCIC Pump Min Flow Valve
11/07/2018
Revision
Notification:
600936099
PM Revision for 125 V Division 1 Reserve Battery
Charger
2/02/2014
SSC-001
Safe Shutdown Capability Report
Energy Harbor Contact Integrating Thermal Monitor
Thermal Environmental Report for Perry Nuclear Power
Plant
Operability
Evaluations
SAP 600595320
IN 2010-03: Failures of Motor-Operated Valves due to
Degraded Stem Lubricant
Procedures
ARI-H13-P601-
21
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
EOP-SPI 6.6
RCIC Injection and Pressure Control
FTI-F0016
Motor Operated Valve Diagnostic Testing
FTI-F0019
Engineering Review of MOV Test Results
IOI-0011
Shutdown from Outside Control Room
MPP
Perry Operations Manual: Motor Operated Valve Program
Plan
NOBP-ER-3601A
Motor Operated Valve Program Torque and Thrust
Requirement and Actuator Capability
NOBP-ER-3601C
Motor Operated Valve Program Periodic Verification
NOBP-ER-3601D
Motor Operated Valve Program Diagnostic Test
Preparation and Evaluation
NOP-ER-2008
Appendix J Program
NOP-ER-3204
Inservice Testing Program
NORM-ER-2001
Preconditioning Structures, Systems and Components
ONI-C61
Evacuation of the Control Room
ONI-P54
Fire
PAP-1101
Inservice Testing of Pumps and Valves
PDB-C0007
Valves Potentially Susceptible to Pressure Locking or
PDB-C0010
ICS Valve Stroke Time Correction Times
PMI-0030
Maintenance of Limitorque Valve Operators
PTI-E51-P0005
Dynamic Diagnostic Testing of RCIC System Valves
SOI-C61
Remote Shutdown System
SOI-E51
Reactor Core Isolation Cooling System
SVI-E51-T1272
RCIC System Low Pressure Operability Test
SVI-E51-T2001
RCIC Pump and Valve Operability Test
& 45
SVI-E51-T9104
Type C Local Leak Rate Test of 1E51 Penetration P104
SVI-E51-T9422
Type C Local Leak Rate Test of 1E51 Penetration P422
TAI-1101-2
Inservice Testing of ASME OM Code Valves
03-003686-000
1E51F0064 RCIC Steam Supply Outboard Isolation Leak
Repair
05/17/2003
200560762
MSIV Full Stroke Operability Test
03/19/2015
Work Orders
200635134
MSIV Full Stroke Operability Test
03/11/2017
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
200637728
MSIV Full Stroke Operability Test
04/08/2015
200643856
Type C Local Leak Rate Test of 1B21 MSL Penetrations
03/13/2017
200643869
SVI-E51T9104 Type C Local Leak Rate Test of 1E51
Penetration P104
03/20/2017
200644390
HPCS Containment Isolation Valve Leak Rate Test
03/06/2017
200644915
Partial PMI-0030 1E51F0063
03/14/2017
200644937
HPCS Waterleg Pump and Associated Valves Cold
Shutdown Operability Test
03/05/2017
200644972
Static MOV Test and Limitorque Maintenance 1E51F0019
03/27/2016
200648070
SVI-E51T9422 Type C Local Leak Rate Test of 1E51
Penetration P422
03/25/2021
200694811
Partial PMI-0030 1E51F0063
03/21/2019
200696003
SVI-E51T9104 Type C Local Leak Rate Test of 1E51
Penetration P104
03/20/2017
200708624
HPCS Waterleg Pump and Associated Valves Cold
Shutdown Operability Test
03/12/2019
200708660
HPCS Containment Isolation Valve Leak Rate Test
03/12/2019
200709140
MSIV Full Stroke Operability Test
03/16/2019
200709311
Type C Local Leak Rate Test of 1B21 MSL Penetrations
03/17/2019
200710412
Partial PMT SVI-E51-T2001
03/27/2017
200712704
Safety Relief Valve Set Pressure Testing
03/27/2019
200712769
HPCS Waterleg Pump and Associated Valves Cold
Shutdown Operability Test
03/25/2021
200718796
SVI-E51T2001 RCIC Pump and Valve Operability Test
07/25/2018
200728988
SVI-E51T2001 RCIC Pump and Valve Operability Test
10/26/2019
200728990
SVI-E51T2001 RCIC Pump and Valve Operability Test
04/21/2020
200731306
SVI-E12T2001 RHR A Pump and Valve Operability
09/24/2019
200738377
SVI-E51T2001 RCIC Pump and Valve Operability Test
07/21/2020
200740485
SVI-E51T2001 RCIC Pump and Valve Operability Test
01/31/2020
200750875
SVI-E51T2001 RCIC Pump and Valve Operability Test
10/19/2020
200769783
T/S Repair RCIC Min Flow Valve
10/24/2018
200779935
HPCS Waterleg Pump and Associated Valves Cold
Shutdown Operability Test
03/11/2021
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
200783359
MSIV Full Stroke Operability Test
03/16/2021
200783960
HPCS Containment Isolation Valve Leak Rate Test
03/15/2021
200784128
Type C Local Leak Rate Test of 1B21 MSL Penetrations
04/01/2021
200792383
SVI-E51T2001 RCIC Pump and Valve Operability Test
01/19/2021
200847374
Replace 2" SRV Air Supply Hose; Hose Damaged and
Leaking
03/25/2021
200847948
SVI-E51T1272 RCIC System Low Pressure Operability
Test
04/02/2021
200858798
RCIC Pump and Valve Operability Test
07/09/2021
MOV1E22-01
1E22-F001, F004, F010, F011, F012, F015 and F023
delta p
2