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
| 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
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.
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.
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 .
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.
(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
- 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.
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.
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.
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