Information Notice 2021-01, Lessons Learned from U.S. Nuclear Regulatory Commission Inspections of Design-Basis Capability of Power-Operated Valves at Nuclear Power Plants

Lessons Learned from U.S. Nuclear Regulatory Commission Inspections of Design-Basis Capability of Power-Operated Valves at Nuclear Power Plants

ML23129A014
Person / Time
Issue date:	07/24/2023
From:	Russell Felts NRC/NRR/DRO
To:
References
IN-21-001, Suppl 1
Download: ML23129A014 (10)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, DC 20555-0001

July 24, 2023

INFORMATION NOTICE 2021-01, SUPPLEMENT 1: LESSONS LEARNED FROM U.S.

NUCLEAR REGULATORY

COMMISSION INSPECTIONS OF

DESIGN-BASIS CAPABILITY OF

POWER-OPERATED VALVES AT

NUCLEAR POWER PLANTS

ADDRESSEES

All holders of operating licenses, construction permits, or com bined licenses for nuclear power

reactors, except those that have permanently ceased operations and have certified that fuel has

been permanently removed from the reactor vessel.

PURPOSE

The U.S. Nuclear Regulatory Commission (NRC) is issuing this su pplement to Information

Notice (IN) 2021-01, Lessons Learned from U.S. Nuclear Regulat ory Commission Inspections

of Design-Basis Capability of Pow er-Operated Valves at Nuclear Power Plants, dated May 6,

2021 (Agencywide Documents Access and Management System (ADAMS) Accession No.

ML21061A265) to alert addressees to lessons learned from NRC in spections of the

design-basis capability of power-operated valves (POVs) at nucl ear power plants. The NRC

expects that addressees will review the information for applica bility to their facilities and

consider actions, as appropriate, to identify and address simil ar issues. Suggestions contained

in this IN are not NRC requirements. Therefore, no specific act ion or written response is

required.

DESCRIPTION OF CIRCUMSTANCES

As discussed in IN 2021-01 (ML21061A265), the NRC staff initiat ed an inspection program

described in Attachment 21N.02, Design-Basis Capability of Pow er-Operated Valves Under

10 CFR 50.55a Requirements, to NRC Inspection Procedure (IP) 7 1111, Reactor Safety

Initiating Events, Mitigating Systems, Barrier Integrity. The most recent revision to IP

71111.21N.02 is dated October 9, 2020, and is publicly availabl e at ADAMS Accession No.

ML20220A667. The NRC issued IP 71111.21N.02 to assess the relia bility, functional capability, and design-basis capability of risk-important POVs to determine whether licensees are

maintaining the POV capability to perform as intended under des ign-basis conditions. During

public meetings in late 2019 and early 2020 (for example, see A DAMS Accession Nos.

ML19351E131 and ML20038A207), the NRC staff described the purpo se of the

IP 71111.21N.02 inspections and indicated that lessons learned from those inspections would

be made available to the stakeholders. During a public meeting on December 8, 2020

(ML20338A012), participants requested that the lessons learned from the initial POV

inspections be documented and made available as soon as possibl e. As a result, the NRC

issued IN 2021-01 to provide lessons learned from the POV inspe ctions conducted in 2020.

ML23129A014 IN 2021-01, Supplement 1 During the POV inspection program, the NRC staff presented less ons learned from POV

inspections at several industry meetings. For example, the NRC staff presented lessons learned

from POV inspections at a public meeting with the Boiling Water Reactor Owners Group

(BWROG) on December 1, 2021 (ML21334A168), and at a Motor-Opera ted Valve (MOV) Users

Group meeting on January 24, 2023 (ML23018A081). With the compl etion of the POV

inspection program at the end of 2022, participants at the Janu ary 24, 2023, meeting requested

that the NRC staff provide a complete list of the lessons learn ed from all of the POV inspections

as soon as possible.

DISCUSSION

The NRC staff conducted inspections using IP 71111.21N.02 to as sess the reliability, functional

capability, and design-basis capability of POVs to determine wh ether licensees are maintaining

the POV capability to perform their safety functions as intende d under design-basis conditions.

The enclosure to IN 2021-01 contains background information rel ated to the design-basis

capability of POVs in nuclear power plants. The NRC inspections using IP 71111.21N.02 identified numerous lessons learned related to the design-basis capability of POVs installed in

nuclear power plants.

The following summarizes the lessons learned from the POV inspe ctions conducted by the NRC

staff using IP 71111.21N.02:

• Inservice Testing (IST) Program: The NRC regulations in Title 10 of the Code of Federal

Regulations (10 CFR) 50.55a, Codes and standards, require licensees to d evelop an

IST program to provide assurance of the operational readiness o f pumps, valves, and

dynamic restraints in accordance with the applicable edition an d addenda of the

American Society of Mechanical Engineers (ASME) Operation and M aintenance of

Nuclear Power Plants, Division 1, OM Code: Section IST (OM Cod e), as incorporated

by reference in 10 CFR 50.55a. For POVs within the scope of the applicable edition and

addenda of the ASME OM Code, the NRC inspectors found that lice nsees did not

always ensure that valves were properly included and categorize d within the scope of

the IST program, such as POVs with leakage limitation safety fu nctions, remote-operated safety functions, or manual-operated safety fun ctions.

• POV Operating Requirements and Capability: The NRC inspectors found that licensees

did not always properly determine the operating requirements an d actuator capability for

POVs to perform their safety functions. For example, all approp riate parameters (such

as valve friction coefficients or valve factors, maximum differential pressure conditions, motor torque temperature derating factors, stem friction coeffi cients, and butterfly valve

bearing friction coefficients) are expected to be addressed whe n calculating valve

operating requirements or act uator capability. Improper values for various parameters in

POV calculations (such as incorrect stem pitch and lead values, valve, and stem friction

coefficients less than tested values, and incorrect uncertainty assumptions) can lead to

inadequate determinations of POV functionality. The NRC inspect ors found that

licensees did not always justify the use of POV parameters, such as valve friction

coefficients, from outside sources. See IN 2012-14, Motor-Oper ated Valve Inoperable

Due to Stem-Disc Separation, dated July 24, 2012 (ML12150A046) for guidance on

using POV data from outside sources. The NRC inspectors found that licensees did not

always ensure that valve-specific valve factors were used if de termined to be higher than

generic valve factors with an appropriate extent of condition r eview. For globe valves, there is a potential for increased thrust and torque requiremen ts (referred to as side

IN 2021-01, Supplement 1 loading) to operate globe valves under high-flow dynamic condit ions. The unwedging

load required for valves is part of the evaluation of the capab ility of POVs to open to

perform their safety functions. The specific design of each POV , including its valve, is

used in determining appropriate calculation assumptions. The NR C inspectors found that

licensees did not always ensure that all normal operating loads that act simultaneously

with seismic loads were addressed. For MOVs, high ambient tempe rature can impact

MOV motor output, such as described in Limitorque Technical Upd ate 93-03, Reliance

3-Phase Limitorque Corporation Actuator Motors (Starting Torque @ Elevated

Temperature), dated September 1993 , which is available from Flowserve Corporation.

The NRC inspectors found that licen sees did not always ensure t hat sufficient

information and test data were developed to validate the assump tions for rate-of-loading

and load-sensitive behavior for plant-specific MOV applications . Stem lubricant

degradation can impact the performance of all types of MOV stem nuts, including the

ball-screw design. One-time stall torque limits for actuators a re intended to address the

structural capability of the actuator rather than calculating p erformance capability.

• Joint Owners Group (JOG) Program for MOV Periodic Verificatio n: Most licensees

committed to implement the JOG Program on MOV Periodic Verifica tion in response to

Generic Letter (GL) 96-05, Periodic Verification of Design-Bas is Capability of

Safety-Related Motor-Operated Valves, dated September 18, 1996 (ADAMS Legacy

Library Accession No. 9609100488). The NRC staff accepted the J OG topical report on

the JOG Program on MOV Periodic Verification in a safety evalua tion report (SER) dated

September 25, 2006 (ML061280315), and the associated supplement dated

September 18, 2008 (ML082480638). In November 2006, the JOG iss ued Topical

Report MPR-2524-A, Joint Owners Group (JOG) Motor Operated Va lve Periodic

Verification Program Summary (ML063490194), to reflect the fin al NRC SER and

included the JOG responses to NRC staff requests for additional information and the

final SER. The JOG MOV Program included a limited amount of MOV tests performed

by the participating licensees at their nuclear power plants ov er approximately 5 years to

assess whether there was a potential for degradation of valve f riction coefficients for

various valve types and applications. Because of the limited am ount of MOV test data

and the different methods used by individual licensees to evalu ate the test data, the

valve friction coefficients determined for MOVs as part of the JOG MOV Program do not

represent a database of valve friction coefficients that can be applied in general to

calculate the thrust and torque required to operate various MOV s under design-basis

conditions. Therefore, the MOV test results collected by partic ipants of the JOG MOV

Program are only applicable to the implementation of the JOG MO V Program. The NRC

inspectors found that licensees did not always re-justify the q ualifying basis for MOVs

following extensive maintenance (such as disassembly) to determ ine whether the valves

were susceptible to performance degradation as part of the JOG MOV Program. The

JOG periodic verification test intervals are based on the margi n and risk ranking of each

MOV within the scope of the JOG MOV Program, such that up-to-da te POV risk rankings

are important when implementing the JOG MOV Program.

• ASME OM Code, Appendix III, Preservice and Inservice Testing of Active Electric

Motor-Operated Valve Assemblies in Water-Cooled Reactor Nuclear Power Plants: As

required under 10 CFR 50.55a(b)(3)(ii), licensees implementing the 2009 or later

editions of the ASME OM Code, as incorporated by reference in 1 0 CFR 50.55a, must

meet the MOV requirements in ASME OM Code, Mandatory Appendix I II. For MOVs

within the scope of the JOG MOV Program, a licensee may rely on the dynamic testing

conducted as part of that program to satisfy the requirement in Appendix III for a mix of

IN 2021-01, Supplement 1 static and dynamic testing. The ASME OM Code, Mandatory Append ix III, as

incorporated by reference in 10 CFR 50.55a relies on new MOVs b eing demonstrated to

be capable of performing their safety functions.

• Licensee Commitments: The NRC regulations in 10 CFR 50.55a(b)( 3)(ii) supplement the

testing requirements for MOVs in the ASME OM Code by requiring that licensees

establish a program to ensure that MOVs continue to be capable of performing their

design-basis safety functions. When implementing the JOG MOV Pr ogram, the MOV

diagnostic test frequency is based on the provisions of the JOG MOV Program, such as

when the design-basis capability margin is determined to be low . Licensees committed

to implementing the JOG MOV Program are expected to follow thei r commitment

process to modify the JOG MOV Pr ogram test intervals or notify the NRC in accordance

with that process. For example, the JOG MOV Program does not in clude grace periods

for the specified JOG test intervals. Further, the JOG program schedule is specified in

years rather than refueling outages. In addition, a change in t he risk ranking of an MOV,

or an adjustment to MOV capability margin based on performance data, can result in a

different diagnostic testing interval under the JOG MOV Program .

• MOVs Outside JOG MOV Program Scope: JOG Topical Report MPR-252 4-A indicates

that some MOVs are outside the scope of the JOG MOV Program, wh ich are defined by

JOG as Class D valves. Therefore, licensees committed to implem enting the JOG MOV

Program to satisfy GL 96-05 and that are implementing the JOG M OV Program as part

of their compliance with 10 CFR 50.55a(b)(3)(ii) are required b y the NRC regulations to

establish methods to periodically demonstrate the design-basis capability of their

Class D valves. The NRC staff considers it infeasible to modify the classification of a

JOG Class D valve to a JOG Class A or JOG Class B valve, which the JOG defines as

not susceptible to degradation by direct information or not sus ceptible to degradation by

extension, respectively.

• Electric Power Research Institute (EPRI) MOV Performance Predi ction Methodology

(PPM): The NRC inspectors found that licensees evaluating MOVs using the EPRI

MOV PPM did not always address all of the applicable provisions when determining

valve operating requirements under the EPRI MOV PPM Program. JOG Topical

Report MPR-2524-A, and the EPRI M OV PPM Topical Report TR-10323 7, as

accepted in the applicable NRC safety evaluations 1 specify the conditions for

implementing these programs. As part of the EPRI MOV PPM Method ology, EPRI

assumed that each valve is maintained in good condition for the EPRI MOV PPM to

remain valid for that valve. Therefore, MOVs classified as JOG Class A or JOG

Class B need to be maintained in good internal condition to sat isfy the EPRI MOV

PPM. Further, this method includes EPRI Type 1 warnings, which indicate potential

valve damage, when implementing the EPRI MOV PPM. Where the EPRI MOV PPM

is used as the best available information, industry data should be monitored for those

valves to identify any information that might challenge that as sumption. When

implementing the EPRI MOV PPM for butterfly valves, the calcula ted maximum

transmitted torque is applied when evaluating the acceptability of the valve weak link

and actuator ratings. When applying the EPRI MOV PPM for globe valves, the globe

valve model in the EPRI methodology specifies the provisions to be implemented, such as using the outside seat diameter to calculate the requir ed operating thrust.

1 The EPRI MOV PPM safety evaluation report is available at ML15142A761 with later updates based on topical

report supplements.

IN 2021-01, Supplement 1 Separate EPRI guidance for evaluating MOV diagnostic test data obtained under

static conditions (i.e., without differential pressure or flow) cannot be applied beyond

the capability of that testing to predict MOV performance under dynamic conditions

(i.e., differential pressure and flow). Additional guidance on the EPRI methodology is

provided in NUREG-1482, Guidelines for Inservice Testing at Nu clear Power

Plants, Revision 3, issued July 2020 (ML20202A473).

• Limitorque Actuator Structural Capability: The NRC inspectors found that licensees

evaluating Limitorque motor actuators for their structural capa bility did not always justify

increasing the thrust ratings beyond their original limits. Lim itorque Technical Update

92-01, Thrust Rating Increase SMB-000, SMB-00, SMB-0 & SMB-1 A ctuators (undated

technical guidance available from Limitorque) evaluated Kalsi E ngineering Document

1. 1707C (a proprietary report by Kalsi Engineering) and approved its use to increase the

maximum allowable thrust for Limitorque actuator models SMB-000 , SMB-00, SMB-0,

and SMB-1 up to 140 percent of the original ratings, with certa in conditions.2 Limitorque

has indicated that licensees that participated in the Kalsi stu dy or that possess a copy of

proprietary Kalsi Engineering Document #1707C may apply the 162 percent maximum

thrust rating described in the Kalsi report, where the specific conditions are implemented

as provided in that document. The individual POV subparts are e xpected to be able to

withstand the maximum thrust and torque that the POV actuator c an produce

(sometimes referred to as a weak link evaluation). The structu ral limits specified in the

ASME Boiler and Pressure Vessel Code are not applicable to POV internal parts that

involve the operating motion of the valve and actuator. Proper bolt material and length

are part of weak link calculations for POVs.

• POV Testing: For POV diagnostic testing, the NRC inspectors fo und that licensees did

not always ensure that (1) POV tests were properly conducted, ( 2) acceptance criteria

for the POV testing applied the correct assumptions (such as ac tuator thrust limits), (3)

proper evaluations of test data were completed to demonstrate t hat the POVs can

perform their safety functions, and (4) records of evaluations were maintained in

accordance with plant procedures. Computer software relies on a ppropriate values for

applicable parameters to be input when conducting diagnostic te sting to determine

accurate thrust and torque values (such as proper stem material properties). POV test

acceptance criteria are expected to be properly translated from POV design calculations

into test procedures. Diagnostic equipment are expected to be i nstalled and operating

properly as part of the POV testing and evaluation of results. Operating requirements for

valves apply throughout the full valve stroke. Fully complete P OV test data evaluations

will ensure that the required parameters (such as valve frictio n coefficient or valve factor, stem factor, and rate of loading) are properly calculated and w ithin the acceptable range.

The JOG MOV Program specifies that valve friction values from t esting are compared to

the JOG threshold values for valve friction to verify that the valve is operating in a

manner consistent with the results of the JOG program assumptio ns. Variation in valve

performance can occur when relying on a single test to establis h POV operating

requirements.

• POV Leakage Limitations: Some POVs have specific limitations r elated to leakage past

the valve disk when closed. MOVs can be set to fully close and meet their leakage

2 NRC IN 92-83, Thrust Limits for Limitorque Actuators and Potential Overstressing of Motor-Operated

Valves, dated December 17, 1992, discussed Limitorque Technical Update 92-01 and the applicable study

by Kalsi Engineering.

IN 2021-01, Supplement 1 limitations when controlled by the torque switch. MOVs that hav e a safety function to

close and be leaktight have more challenges when controlled by the limit switch instead

of the torque switch. For example, the NRC inspectors found tha t licensees did not

always have a valid test or analysis demonstrating that the lim it switch control setting of

the MOV under static conditions would achieve the required leak tight performance when

the MOV is closed under dynamic conditions. The leak rate requi rements are also to be

addressed for MOVs with long closing torque switch bypass setti ngs. The ASME OM

Code as incorporated by reference in 10 CFR 50.55a requires a d ocumented program

for leak-testing power-operated relief valves. With respect to previous POV capability

issues, GL 79-46, Containment Purging and Venting During Norma l Operation

Guidelines for Valve Operability, dated September 27, 1979 (ML 031320191), provides

recommendations to demonstrate that containment purge valves ca n close and seal

under design-basis conditions, including seismic loads.

• POV Qualification: The NRC inspectors found that licensees di d not always justify the

qualification of POVs to perform their design-basis safety func tions, including functional, environmental, and seismic capab ility. With respect to environm ental qualification, preventive maintenance activities include replacing all valve s ubcomponents within their

specific qualified lifetime. Environmental effects can affect t he performance of POVs

(including squib valves) that must remain functional for long p eriods of time following a

loss-of-coolant accident or other adverse conditions. NRC inspe ctions identified that

some licensees lacked adequate justification to extend the qual ified life of POVs

installed in their nuclear power plants. Limitorque qualified i ts safety-related MOV

actuators for 40 years or 2,000 cycles, whichever comes first. Licensees may extend the

qualified life of their Limitorque actuators if they have adequ ate justification. The

justification for the extension of the qualified life of the ac tuator, including attention to

radiation levels and ambient temperature conditions where MOVs are located, includes

assurance that the environmental qualification requirements are not exceeded and that

appropriate replacement frequencies for MOVs or their individua l parts are established.

EPRI has developed guidance for extending the qualified life of Limitorque actuators

beyond their original qualified life. The presence of radiation hot spots and ambient

temperature conditions can impact the service life for the envi ronmental qualification of a

valve actuator.

• MOV Stem-Disk Connections: The NRC staff discussed operating e xperience with

MOV stem-disk connections in IN 2017-03, Anchor/Darling Double Disc Gate Valve

Wedge Pin and Stem-Disc Separation Failures, dated June 15, 20 17 (ML17153A053). The BWROG prepared guidance to address the issue of potential

failure of the stem-disk connection in Anchor/Darling double-di sk gate valves. The

BWROG guidance (such as evaluating the weak link of the wedge p in under motor

stall conditions) includes specific provisions in assessing the susceptibility for

separation of the stem-disk connection in Anchor/Darling double -disk gate valves.

• Valve Position Verification: Paragraph ISTC-3700, Position Ve rification Testing, in

Subsection ISTC, Inservice Testing of Valves in Water-Cooled R eactor Nuclear

Power Plants, of the ASME OM Code requires that valves with re mote position

indicators be observed locally at least once every 2 years to v erify that valve

operation is accurately indicated. The NRC regulations in 10 CF R 50.55a(b)(3)(xi)

specify supplemental position indication (SPI) requirements whe n implementing

ASME OM Code, 2012 Edition (or later editions), paragraph ISTC- 3700, for

licensees to verify that valve operation is accurately indicate d by supplementing

IN 2021-01, Supplement 1 valve position indicating lights with other indications, such a s flow meters or other

suitable instrumentation, to provide assurance of proper obtura tor position for valves

with remote position indication within the scope of Subsection ISTC including its

mandatory appendices and their verification methods and frequen cies. Licensees

proposing additional time to implement the 2012 or later editio ns of the ASME OM

Code (including 10 CFR 50.55a(b)(3)(xi)) may submit a request f or an alternative in

accordance with 10 CFR 50.55a(z) for NRC staff review. Addition al information on

this topic is found in two monthly Reactor Oversight Process me eting summaries

(ML21041A409 and ML21047A290). The NRC regulations in 10 CFR

50.55a(b)(3)(xi) require verification of valve position indicat ion, including specifying

actions to meet SPI requirements such as leakage testing, flow measurement, or

diagnostic trace analysis.

• Valve Packing and Backseating: Valve packing replacements or a djustments can cause

anomalous behavior that might adv ersely impact valve performance. A bent or damaged

stem can cause packing loads to become more severe with valve o peration. On

occasion, some licensees backseat the stem of a valve to limit packing leaks. The NRC

inspectors found that licensees did not always conduct a detail ed evaluation (including

appropriate examination) of the effects of backseating on the v alve bonnet and stem to

verify structural integrity. NUREG-1482 provides additional guidance for controlling the

backseating process for a valve stem.

• Use of POV Computer Software: The NRC inspectors found that li censees did not

always perform a complete verification and validation of POV co mputer software prior to

implementation. These calculation methodologies need verificati on and validation for

appropriate assumptions and data points. Further, stroke time m ight be calculated

improperly when computer data are used to measure the MOV strok e time. The ASME

OM Code specifies that the stroke time for a valve begins with the initiating signal and

ends with completion of the valve stroke. However, some compute r data output does not

include the initial portion of the stroke signal for calculatin g the stroke time. It is important

to update POV programs to address new computer software used in POV calculations.

• MOV Thermal Overload Devices: Thermal overload devices are ins talled in the control

circuitry for some MOVs to protect the motor from damage in the event of an overload

event. The performance of thermal overload devices can impact t he safety function of

MOVs if not evaluated periodically. NRC Regulatory Guide 1.106 (Revision 2), Thermal

Overload Protection for Electric Motors on Motor-Operated Valve s, dated

February 2012 (ML112580358) provides guidance for the use of th ermal overloads that

reflects lessons learned from MOV programs.

• MOV Throttling Operation: Motors used to operate MOVs have lim itations regarding their

operating time. Limitorque specifies cooldown times for the fre quent operation of MOV

motors. The NRC inspectors found that licensees did not always evaluate the impact of

motor heat-up on the capability of MOVs with design-basis safet y functions to throttle

system flow.

• Actuator Handwheel Operation: Some licensees rely on the actua tor handwheel to

manually operate MOVs to perform important functions at their n uclear power plants. For

such MOVs, the NRC inspectors found that licensees did not alwa ys evaluate the

handwheel for proper sizing and good working condition in demon strating that the MOV

IN 2021-01, Supplement 1 could perform its safety function. Improperly operating a valve by its manual handwheel

can result in excessive handwheel torque that can damage the ac tuator and the valve.

• Preventive Maintenance and Modifications: The NRC inspectors found that licensees

did not always determine a proper lubrication interval for each MOV stem to address

potential lubrication grease degradation which can adversely af fect MOV operation.

MOVs installed in non-normal positions can cause MOV maintenanc e issues. For

example, grease leakage into t he limit switch compartment might interfere with the

electrical operation of actuator wiring. Further, an MOV orient ed with the disk in the

horizontal plane can lead to abnormal performance of a gate val ve as a result of

increased disk and guide wear over time. In addressing potentia l pressure locking of a

valve, modifications that prevent a valve from pressure locking , such as drilling a hole in

the valve disk, can have long-term consequences (such as a perm anent one-way valve).

The NRC regulations in 10 CFR 50.59, Changes, tests and experiments, are applicable

to pressure-locking modifications for MOVs. Potential degradati on of magnesium rotors

in motors can adversely impact MOV performance. Missing or dam aged external and

internal parts of motors and actuators can impact operational r eadiness or qualification

of a POV.

• Corrective Action: The NRC inspectors found that licensees did not always ensure that

appropriate corrective actions in accordance with plant procedu res were implemented

when (1) POV test results fell outside of the specified accepta nce criteria, (2) POV

performance anomalies were observed, such as abnormal diagnosti c traces or valve

friction degradation, or (3) a mechanical problem with the POV was identified, such as a

manual declutch lever malfunction. The ASME OM Code as incorpor ated by reference in

10 CFR 50.55a includes corrective action requirements for POV l eak testing. Overload

events when testing or operating POVs are expected to be addres sed in accordance

with the licensees corrective action program and the manufactu rer recommendations.

• POV Records: The NRC inspectors found that licensees did not a lways follow their

procedures for maintaining records associated with POV qualific ation, testing, operation, maintenance, and corrective action, in accordance with the qual ity assurance

requirements in 10 CFR Part 50, Domestic Licensing of Producti on and Utilization

Facilities, Appendix B, Quality Assurance Criteria for Nuclea r Power Plants and Fuel

Reprocessing Plants. As part of the QA program, POV performanc e is monitored and

appropriate reports prepared in accordance with plant procedure s to identify any

adverse indications.

• IST Programs and Technical Specifications: Nuclear power plant licensees are required

to meet the NRC regulations in both 10 CFR 50.36, Technical sp ecifications, and

10 CFR 50.55a for IST programs. Following the criteria in 10 CF R 50.59(c)(1), licensees

must prepare a license amendment to revise its technical specif ications when making

changes to POV parameters (such as main steam isolation valve a ccumulator pressure)

as part of its IST program.

• IST Programs and 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage

Testing for Water-Cooled Power Reactors: The ASME OM Code, as incorporated by

reference in 10 CFR 50.55a, allows licensees to follow leak tes ting intervals for valves in

accordance with 10 CFR Part 50, Appendix J, in certain instance s. Licensees might

perform POV static testing to meet the containment leakage test ing requirements in

10 CFR Part 50, Appendix J. In addition, the NRC regulations i n 10 CFR 50.55a(b)(3)(ii)

IN 2021-01, Supplement 1 require that MOV design-basis capability be justified periodica lly. POV leakage

requirements might be specified in final safety analysis as par t of the IST program

description, in addition to the 10 CFR Part 50, Appendix J, req uirements.

The NRC staff discussed the above issues in detail with the app licable licensees during the

POV inspections. The licensees took action to address any immed iate concerns related to these

issues identified by the NRC inspectors. In many instances, the issues were determined to be

minor because of the capability margin available for the specif ic POVs being evaluated at the

applicable nuclear power plant. The issues might have been more significant where less

capability margin was available for POVs at other nuclear power plants. Some licensees

initiated long-term activities as appropriate to address specif ic issues as part of their corrective

action programs. The NRC staff suggests that licensees review t his information for applicability

to their facilities and consider actions, as appropriate, to id entify and address similar issues.

CONTACT

S

This IN requires no specific action or written response. Please direct any questions about this

matter to the technical contacts listed below or to the appropr iate Office of Nuclear Reactor

Regulation (NRR) project manager.

/RA/

Russell Felts, Director

Division of Reactor Oversight

Office of Nuclear Reactor Regulation

Technical Contacts:

Douglas Bollock, NRR Kenneth Kolaczyk, NRR Thomas Scarbrough, N RR

301-415-6609 585-773-8917 301-415-2794 Douglas.Bollock@nrc.gov Kenneth.Kolaczyk@nrc.gov Thomas.Scarbrough@nrc.gov

Note: NRC generic communications may be found on the NRC public website, http://www.nrc.gov, under Electronic Reading Room/Document Collections.

IN 2021-01, Supplement 1 NRC INFORMATION NOTICE 2021-01, SUPPLEMENT 1, LESSONS LEARNED FROM NRC

INSPECTIONS OF DESIGN-BASIS CAPABILITY OF POWER-OPERATED VALVES AT

NUCLEAR POWER PLANTS, DATED: July 24, 2023

AD AMS Accession No.: ML23129A014 EPIDS No.

OFFICE Author QTE NRR/DEX/EMIB/BC OE NRR/DRO/IOEB/PM

NAME TScarbrough Jay Dougherty SBailey JPeralta PClark

DATE 5/22/23 5/15/2023 5/18/23 5/19/23 5/22/23

OFFICE NRR/DRO/LA NRR/DRO/ NRR/DRO/IOE NRR/DRO/I

IOEB/PM B/PM OEB/BC NRR/DRO/D

NAME IBetts BBenny PClark LRegner RFelts

DATE 7/13/2023 5/22/23 5/22/23 7/20/23 7/24/23

OFFICIAL RECORD COPY