Guidelines for 

Inservice Testing at

Nuclear Power Plants

Inservice Testing of Pumps and Valves and 

Inservice Examination and Testing of Dynamic 

Restraints (Snubbers) at Nuclear Power Plants

Draft Report for Comment

Office of Nuclear Reactor Regulation

NUREG-1482

Rev. 2
Guidelines for 

Inservice Testing at

Nuclear Power Plants

Inservice Testing of Pumps and Valves and 

Inservice Examination and Testing of Dynamic 

Restraints (Snubbers) at Nuclear Power Plants

Draft Report for Comment

Manuscript Completed: July 2011

Date Published: August 2011

Prepared by

Gurjendra S. Bedi

Office of Nuclear Reactor Regulation

NUREG-1482

Rev. 2
COMMENTS ON DRAFT REPORT

Any interested party may submit comments on this report for consideration by the NRC staff. 

Comments may be accompanied by additional relevant information or supporting data. Please 

specify the report number NUREG-1482, Revision 2, draft, in your comments, and send them by 

December 20, 2011 to the following address:

Cindy Bladey, Chief

Rules, Announcements, and Directives Branch

Division of Administrative Services

Office of Administration

Mail Stop: TWB-05-B01M 

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Electronic comments may be submitted to the NRC by e-mail at Gurjendra.Bedi@nrc.gov.

For any questions about the material in this report, please contact:

Gurjendra S. Bedi

OWFN Mail Stop 09-D3

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

Phone: 301-415-1393

E-mail: Gurjendra.Bedi@nrc.gov

Please be aware that any comments that you submit to the NRC will be considered a public 

record and entered into the Agencywide Documents Access and Management System 

(ADAMS). Do not provide information you would not want to be publically available.
iii

ABSTRACT

In this NUREG, the staff of the U.S. Nuclear Regulatory Commission (NRC) discusses the 

applicable regulations for the inservice testing (IST) of pumps and valves and the examination 

and testing of dynamic restraints (snubbers) at commercial nuclear power plants. The 

information in NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plants,” 

Revision 0, issued April 1995, and Revision 1, issued January 2005, has described this topic in 

the past. This NUREG report replaces Revision 0 and Revision 1 to NUREG-1482 and is 

applicable, unless stated otherwise, to editions and addenda (up to the 2004 Edition including 

2005 and 2006 addendas) to the American Society of Mechanical Engineers (ASME) Code for 

Operation and Maintenance of Nuclear Power Plants (OM Code), that are incorporated by 

reference in which Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(b) incorporates 

by reference. In addition, the staff discusses other IST program topics such as the NRC

process for the review of the OM Code, conditions on the use of the OM Code, interpretations of 

the OM Code, and development of IST programs for new reactors. In Appendix A to this 

NUREG, the staff provides guidance included in Revision 1 to NUREG-1482 that has been 

updated to reflect IST lessons learned and operating experience since issuance of that NUREG. 

Appendix B contains guidance related to inservice examination and testing of snubbers.

PAPERWORK REDUCTION ACT STATEMENT

The information collections contained in this final regulatory guidance are covered by the

requirements of 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,”

which the Office of Management and Budget (OMB) has approved under OMB control 

number 3150-0011.

Public Protection Notification

The NRC may neither conduct nor sponsor, and a person is not required to respond to, a 

request for information or an information collection requirement unless the requesting document 

displays a currently valid OMB control number.
iv

CONTENTS

Page

ABSTRACT......................................................................................................................iii

PREFACE.........................................................................................................................vii

ABBREVIATIONS ............................................................................................................ix

1 INTRODUCTION...................................................................................................1-1

2 REGULATORY BASIS .........................................................................................2-1

2.1 Incorporation by Reference........................................................................2-1

2.2 OM Code Cases ........................................................................................2-1

2.3 Inservice Testing of Pumps and Valves Program Scope............................ 2-2

2.4 Inservice Examination and Testing of Snubbers Program Scope............... 2-2

2.5 Mandatory Inservice Testing and Inspection Program Updates.................. 2-2

2.6 Alternative Requests..................................................................................2-3

2.7 Relief Requests .........................................................................................2-3

2.8 Voluntary Use of Later Editions of and Addenda to the ASME OM Code ... 2-3

2.9 Quality Assurance......................................................................................2-3

2.10 Exemptions................................................................................................2-3

3 NRC REVIEW OF THE ASME OM CODE ............................................................3-1

4 CONDITIONS ON USE OF LIMITATIONS AND MODIFICATIONS TO THE 

ASME OM CODE..................................................................................................4-1

4.1 10 CFR 50.55a(b)(3)(i)—Quality Assurance............................................... 4-1

4.2 10 CFR 50.55a(b)(3)(ii)—Motor-Operated Valve Testing ........................... 4-1

4.3 10 CFR 50.55a(b)(3)(iv)(A), (B), (C), and (D)—Appendix II........................ 4-2

4.4 10 CFR 50.55a(b)(3)(v)—Subsection ISTD................................................4-3

4.5 10 CFR 50.55a(b)(3)(vi)—Exercise Interval for Manual Valves .................. 4-3

5 GUIDANCE FOR DEVELOPING AND IMPLEMENTING IST PROGRAMS.......... 5-1

5.1 Alternative Requests..................................................................................5-1

5.2 Relief Requests .........................................................................................5-1

5.3 Voluntary Use of Later Editions of and Addenda to the ASME OM Code ... 5-2

5.4 Identification of Code Noncompliance........................................................5-3

5.5 Revising NRC-Authorized Alternatives.......................................................5-3

5.6 NRC Verbal Authorization of an Alternative Request ................................. 5-4

5.7 Submittal of Previously NRC-Authorized Alternatives ................................ 5-5

5.8 Standard Format for Requests...................................................................5-5

5.9 ASME OM Code Interpretations.................................................................5-5

5.10 Risk-Informed Inservice Testing.................................................................5-6

5.11 Use of Code Cases....................................................................................5-6
v

6 GUIDANCE FOR SPECIFIC PUMP AND VALVE IST ACTIVITIES ..................... 6-1

6.1 Waterleg Pumps ........................................................................................6-1

6.2 Smooth-Running Pumps............................................................................6-1

6.3 Vibration-Measuring Transducers ..............................................................6-1

6.4 Online Check Valve Sample Disassembly and Inspection.......................... 6-2

6.5 Safety/Relief Valves...................................................................................6-2

7 SNUBBER INSERVICE EXAMINATION AND TESTING PROGRAMS ................ 7-1

7.1 Program Controls.......................................................................................7-1

7.2 OM Part 4 Clarification...............................................................................7-1

8 IST PROGRAM GUIDANCE FOR NEW REACTORS........................................... 8-1

9 REFERENCES......................................................................................................9-1

APPENDIX A: Guidelines for Inservice Testing Program for Pumps and Valves at 

Nuclear Power Plants

(Previously NUREG-1482, Revision 1 with updates) A-i

APPENDIX B:Guidelines for Inservice Examination ande Testing Program for 

Dynamic Restraints (Snubbers) at Nuclear Power Plants B-i
vii

PREFACE

NUREG publications consist of reports or brochures on regulatory decisions, results of 

research, results of incident investigations, and other technical and administrative information. 

Some of the information herein is similar in appearance to U.S. Nuclear Regulatory Commission

staff positions given in a regulatory guide because certain recommendations indicate acceptable 

alternatives to Code requirements. However, this information is not equivalent to staff positions 

in a regulatory guide or other generic correspondence because this information is strictly 

intended for voluntary implementation by licensees.
vii

ABBREVIATIONS

ADAMS Agencywide Documents Access and Management System

ANSI American National Standards Institute

ASME American Society of Mechanical Engineers

B&PV boiler and pressure vessel

BWR boiling-water reactor

CFR Code of Federal Regulations

COL combined license

ECCS emergency core cooling system

FR Federal Register

GL generic letter

IM inspection manual

ISI inservice inspection

IST inservice testing or inservice test

LCO limiting condition operation

MOV motor-operated valve

NEI Nuclear Energy Institute

NRC U.S. Nuclear Regulatory Commission

OM Code Code for Operations and Maintenance of Nuclear Power Plants

PdM predictive maintenance

RG regulatory guide

RIS regulatory issue summary

RTNSS regulatory treatment of non-safety systems

SAR safety analysis report

SSC system, structure, and/or component

SR surveillance requirement

S/RV safety/relief valve

TRM technical requirements manual

TS technical specification
1-1

1. INTRODUCTION

The staff of the U.S. Nuclear Regulatory Commission (NRC) is issuing this Revision 2 of the 

NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plants,” to assist the nuclear 

industry in establishing a basic understanding of the regulatory basis for pump and valve 

inservice testing (IST) programs and dynamic restraint (snubbers) examination and testing

programs. This NUREG also provides information regarding the NRC’s involvement in the 

development of the American Society of Mechanical Engineers (ASME) Code for Operation and 

Maintenance of Nuclear Power Plants (OM Code). In this NUREG, the staff discusses 

OM Code inquiries, the inservice examination and testing of snubbers, pump and valve IST, the 

use of ASME code cases, conditions on the use of the OM Code, guidance for OM Code 

noncompliance, requests for alternatives to the OM Code at operating commercial nuclear 

power plants, and the development of IST programs for new reactors. 

The information in NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plants,” 

Revision 0, issued April 1995, and Revision 1, issued January 2005, has described IST 

programs in the past. Revision 2 to NUREG-1482 replaces Revision 0 and Revision 1 to 

NUREG-1482 and is applicable, unless stated otherwise, to editions and addenda (up to the 

2004 Edition including 2005 and 2006 addendas) to the ASME OM Code, that are incorporated 

by reference in which Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(b)

incorporates by reference (Federal Register, Vol. 76, No. 119, page 36232-36279, dated June 

21, 2011).

This NUREG provides guidance for the inservice testing of pumps and valves, and inservice 

testing of dynamic restraints (snubbers) at nuclear power plants based on lessons learned since 

the issuance of Revision 0 and Revision 1 to NUREG-1482. Appendix A to this NUREG 

contains guidance provided in Revision 1 to NUREG-1482 for pumps and valves that has been 

updated for the development of IST programs at nuclear power plants. Appendix B contains 

guidance related to inservice examination and testing of snubbers. The guidance in the main 

text of this NUREG takes precedence over information provided in Appendix A and Appendix B 

to this NUREG.

The guidelines and recommendations provided in this NUREG and its Appendices A and B do 

not supersede the regulatory requirements specified in 10 CFR 50.55a. Further, this NUREG 

does not authorize the use of alternatives to, or grant relief from, the ASME Code requirements 

for inservice testing of pumps and valves, or inservice examination and testing of dynamic 

restraints (snubbers), incorporated by reference in 10 CFR 50.55a.

 
2-1

2. REGULATORY BASIS

The following regulations are applicable to pump and valve IST programs and snubber inservice 

examination and testing programs at operating reactors and are discussed throughout this 

NUREG. As a result of the unique wording in various paragraphs, note that the NRC 

authorizes licensee-proposed alternatives in accordance with 10 CFR 50.55a(a)(3), grants

relief and imposes alternative requirements in accordance with 10 CFR 50.55a(f)(6)(i) and 

10 CFR 50.55a(g)(6)(i), or approves the use of later code editions and addenda in accordance 

with 10 CFR 50.55a(f)(4)(iv) and 10 CFR 50.55a(g)(4)(iv).

2.1 Incorporation by Reference

The regulations at 10 CFR 50.55a, “Codes and Standards,” define the requirements for applying 

industry codes and standards to boiling- or pressurized-water-cooled nuclear power facilities. 

The National Technology Transfer and Advancement Act of 1995 (P.L. 104-113) requires that if 

agencies establish technical standards, they must use technical standards that voluntary 

consensus standards bodies develop or adopt unless the use of such standards is inconsistent 

with applicable law or is otherwise impractical. P.L. 104-113 requires Federal agencies to use 

industry consensus standards to the extent practical; however, it does not require Federal 

agencies to endorse a standard in its entirety. The law does not prohibit an agency from 

generally adopting a voluntary consensus standard while taking exception to specific portions of 

the standard if those provisions are deemed to be “inconsistent with applicable law or otherwise 

impractical.” Furthermore, taking specific exceptions furthers the congressional intent of 

Federal reliance on voluntary consensus standards because it allows the adoption of substantial

portions of consensus standards without the need to reject the standards in their entirety 

because of limited provisions that are not acceptable to the agency. 

The OM Code is a national, voluntary consensus standard, and P.L. 104-113 requires 

Government agencies to use it unless the use of such a standard is inconsistent with applicable 

law or is otherwise impractical. The NRC approves or mandates or both the use of editions and 

addenda to the codes in 10 CFR 50.55a through the rulemaking process of “incorporation by 

reference.” As such, each provision of the codes that 10 CFR 50.55a incorporates by reference 

and mandates constitutes a legally binding NRC requirement imposed by rule. 

The requirements of the OM Code become NRC regulations once they are incorporated by 

reference into 10 CFR 50.55a(b).

2.2 OM Code Cases

The regulation at 10 CFR 50.55a(b)(6) incorporates by reference Regulatory Guide (RG) 1.192, 

Operation and Maintenance Code Case Acceptability, ASME OM Code.” Licensees may 

implement the code cases listed in RG 1.192 without obtaining further NRC review or approval if 

the code cases are used in their entirety with any supplemental conditions specified in the RG. 

RG 1.193, “ASME Code Cases Not Approved for Use,” lists code cases not approved for use. 
2-2

The NRC may authorize the use of a code case that it has not yet approved for use in RG 1.192 

if a licensee requests the use of the code case under 10 CFR 50.55a(a)(3). The NRC may

authorize the use of such a code case until a future revision to RG 1.192 accepts the use of the 

ASME code case. At that time, if the licensee intends to continue implementing the code case, 

it must follow all the provisions of the code case with the conditions specified in RG 1.192, if 

any. The authorization for a specific licensee to use a code case that is not listed in RG 1.192 

does not authorize any other licensee to use the code case without submittal by the subsequent 

licensee of a request to implement an alternative to the ASME OM Code requirements under

10 CFR 50.55a(a)(3).

2.3 Inservice Testing of Pumps and Valves Program Scope

The regulation at 10 CFR 50.55a(f)(4) requires that throughout the service life of a boiling or 

pressurized water-cooled nuclear power facility, pumps and valves which are classified as 

ASME Code Class 1, 2, and 3 must meet the inservice test requirements of the ASME OM 

Code as incorporated by reference in 10 CFR 50.55a(b).

2.4 Inservice Examination and Testing of Snubbers Program Scope

The regulations at 10 CFR 50.55a(g)(4) require that ASME Code Class 1, 2, and 3 components 

(including supports) meet the ISI and IST requirements of Section XI “Rules for Inservice 

Inspections of Nuclear Power Plant Components,” of the ASME Boiler & Pressure Vessel Code

(B&PV Code) or OM Code as incorporated by reference in 10 CFR 50.55a(b). Article 

IWF-5000, “Inservice Inspection Requirements for Snubbers,” of Section XI of the ASME B&PV 

Code in the 2005 Addendum and earlier editions and addenda to Section XI provides the 

requirements for the examination and testing of snubbers in nuclear power plants. Article 

IWF-5000 was deleted in the 2006 Addendum of Section XI of the ASME B&PV Code. The 

inservice examination and testing of snubbers has been a requirement in Article IWF-5000 since 

Article IWF was first issued in the Winter 1978 Addendum to Section XI. Subsection ISTD of 

the OM Code has included provisions for the examination and testing of snubbers since it was 

first issued in 1990. Therefore, 10 CFR 50.55a requires inservice examination and testing of 

snubbers because it incorporates by reference the Section XI requirements in Article IWF-5000

and the OM Code requirements in Subsection ISTD. 

2.5 Mandatory Inservice Testing and Inservice Inspection Program Updates

The regulation at 10 CFR 50.55a(f)(4)(ii) requires licensees to revise their IST programs every 

120 months to reflect the latest edition and addendum to the OM Code incorporated by 

reference into 10 CFR 50.55a(b)(3) that is in effect 12 months before the start of the new 

120-month IST interval.

The regulation at 10 CFR 50.55a(g)(4)(ii) requires licensees to revise their inservice inspection 

(ISI) programs every 120 months to reflect the latest edition and addendum to Section XI, of the 

ASME B&PV Code incorporated by reference into 10 CFR 50.55a(b)(2) that is in effect 

12 months before the start of the new 120-month ISI interval. 
2-3

2.6 Alternative Requests

Licensees can request that the NRC authorize an alternative to an OM or Section XI Code 

requirement in accordance with 10 CFR 50.55a(a)(3). Requests made under 

10 CFR 50.55a(a)(3) are more specifically called “alternatives.” 

2.7 Relief Requests

Licensees can request that the NRC grant relief from an OM or Section XI Code requirement in 

accordance with 10 CFR 50.55a(f)(5)(iii), 10 CFR 50.55a(f)(5)(iv), 10 CFR 50.55a(g)(5)(iii), and 

10 CFR 50.55a(g)(5)(iv). Requests made under 10 CFR 50.55a(f)(5) and 10 CFR 50.55a(g)(5) 

are called “relief.” 

2.8 Voluntary Use of Later Editions and Addenda to the ASME Code

The regulation at 10 CFR 50.55a(f)(4)(iv) states that ISTs for pumps and valves may meet the 

requirements set forth in subsequent editions and addenda to the OM Code that 

10 CFR 50.55a(b)(3) incorporates by reference, subject to NRC approval. The regulation at

10 CFR 50.55a(g)(4)(iv) states that the inservice examination of components (including 

supports) may meet the requirements set forth in subsequent editions and addenda of the OM 

or Section XI Code that 10 CFR 50.55a(b) incorporates by reference, subject to NRC approval.

This includes the examination and testing of snubbers. Licensees may use portions of editions 

or addenda provided that all related requirements of the respective editions or addenda are met. 

When requesting to use editions and addenda to the OM or Section XI Code that have not yet 

been incorporated by reference, licensees must request authorization to use these later editions 

and addenda as an alternative to the regulations under 10 CFR 50.55a(a)(3). The staff issued

Regulatory Issue Summary (RIS) 2004-12, “Clarification on Use of Later Editions and Addenda 

to the ASME OM Code and Section XI,” dated July 28, 2004, in order to clarify this matter. 

Additional discussion of this topic is provided in Section 5.3 of this NUREG.

2.9 Quality Assurance

IST and ISI programs, including implementing procedures, are subject to the requirements of 

Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing 

Plants,” to 10 CFR Part 50. 

2.10 Exemptions

Under 10 CFR 50.12(a), the NRC may, either on its own initiative or upon application by any 

licensee, grant an exemption from the requirements of 10 CFR Part 50 that is authorized by law, 

does not present an undue risk to the public health and safety, is consistent with the common 

defense and security, and is appropriate because of special circumstances. If the NRC 

approves the application, the exemption relieves the licensee from compliance with the 

regulation(s) involved. Exemptions are normally not used for 10 CFR 50.55a requests.
3-1

3. NRC REVIEW OF THE ASME OM CODE

The first edition and addendum to the OM Code that 10 CFR 50.55a incorporated by reference

were the 1995 Edition and the 1996 Addendum. The NRC determines acceptability of new 

provisions in new editions and addenda to the OM Code and the need for conditions on the use 

of the OM Code. Generally, the NRC staff participates with other ASME committee members in 

discussions and technical debates in the development of OM Code revisions. NRC committee 

representatives discuss the codes and technical justifications with other cognizant NRC staff to 

ensure an adequate technical review. Finally, NRC management reviews and approves the 

proposed agency position on the OM Code as part of the rulemaking amending 10 CFR 50.55a 

to incorporate by reference new editions and addenda to the OM Code and conditions on its

use. This process, when considered along with ASME’s own process for developing and 

approving the OM Code, provides reasonable assurance that the NRC approves for use only 

those new and revised OM Code editions and addenda (with conditions as necessary), that 

provide reasonable assurance of adequate protection to public health and safety and that do not 

have significant adverse impacts on the environment.
4-1

4. CONDITIONS ON USE OF LIMITATIONS AND MODIFICATIONS TO 

THE ASME OM CODE

The NRC regulations incorporate by reference specific editions and addenda (up to the 2004 

Edition including 2005 and 2006 addendas) to the ASME OM Code in 10 CFR 50.55a(b)(3),

subject to the five limitations and modifications conditions outlined below. 

4.1 10 CFR 50.55a(b)(3)(i)—Quality Assurance

The OM Code references the use of either the American National Standards Institute 

(ANSI)/ASME NQA-1-1979, “Quality Assurance Program Requirements for Nuclear Facilities,”

issued 1979, or the owner’s Appendix B to 10 CFR Part 50 quality assurance program as part of 

its individual provisions for a quality assurance program. However, ANSI/ASME NQA-1-1979

does not contain some of the quality assurance provisions and administrative controls governing 

operational phase activities that would be required in order to use ANSI/ASME NQA-1-1979 in 

lieu of an owner’s Appendix B quality assurance program description. The NRC originally 

endorsed ANSI/ASME NQA-1-1979 with the knowledge that it was not entirely adequate and 

that other commitments such as the ANSI/ASME standards must supplement it. Hence, 

ANSI/ASME NQA-1-1979 is not acceptable for use without the other quality assurance program 

provisions identified in technical specification (TS) and licensee quality assurance programs. 

4.2 10 CFR 50.55a(b)(3)(ii)—Motor-Operated Valve Testing

This condition requires that licensees establish a program to ensure that motor-operated valves

(MOVs) continue to be capable of performing their design-basis safety functions. The condition 

in 10 CFR 50.55a(b)(3)(ii) supplements the quarterly MOV stroke-time testing requirement in 

Subsection ISTC of the OM Code. Since 1989, the NRC has recognized that quarterly 

stroke-time testing is not sufficient to provide assurance of MOV capability under design-basis 

conditions. For example, in Generic Letter (GL) 89-10, “Safety-Related Motor-Operated Valve 

Testing and Surveillance,” dated June 28, 1989, the NRC stated that stroke-time testing alone is 

not sufficient to provide assurance of MOV operability under design-basis conditions. 

Therefore, in GL 89-10, the NRC staff requested licensees to verify the design-basis capability 

of their safety-related MOVs and to establish long-term MOV programs. The NRC subsequently 

issued GL 96-05, “Periodic Verification of Design-Basis Capability of Safety-Related 

Power-Operated Valves,” dated September 18, 1996, to provide updated guidance for 

establishing long-term MOV programs. The condition in 10 CFR 50.55a(b)(3)(ii) establishes a 

regulatory requirement for nuclear power plant licensees implementing the applicable editions 

and addenda to the ASME OM Code to establish programs to periodically assess the 

design-basis capability of MOVs within the scope of the IST program at nuclear power plants.

Code Case OMN-1, “Alternative Rules for Preservice and Inservice Testing of Certain Electric 

Motor-Operated Valve Assemblies in Light-Water Reactor Power Plants,” for the ASME OM 

Code allows users to replace quarterly MOV stroke-time testing with a combination of MOV 

exercising at least every refueling outage and MOV diagnostic testing on a longer interval. In 

RG 1.192, the NRC addresses the acceptability of Code Case OMN-1 in lieu of the quarterly 

MOV stroke-time testing requirements in Subsection ISTC of the OM Code. The 

implementation of ASME OM Code Case OMN-1 as accepted in RG 1.192 can be used in 

satisfying the requirement in 10 CFR 50.55a(b)(3)(ii). ASME has incorporated ASME OM Code 
4-2

Cases OMN-1, “Alternative Rules for Preservice and Inservice Testing of Certain Electric 

Motor-Operated Valve Assemblies in LWR Power Plants, “ and OMN-11, “Risk Informed Testing 

of Motor-Operated Valves,” as Appendix III in the 2009 Edition to the ASME OM to replace 

quarterly MOV stroke-time testing with periodic exercising and diagnostic testing. In future the 

NRC staff will review the 2009 Edition of the ASME OM Code (including Appendix III) for 

incorporation by reference in 10 CFR 50.55a with any appropriate modification, limitation 

conditions. (Note: The details related to the ASME OM-2009 for information only)

4.3 10 CFR 50.55a(b)(3)(iv)(A), (B), (C), and (D)—Appendix II

This condition supplements the provisions in Appendix II, “Check Valve Condition Monitoring 

Program,” to the OM Code. Subsection ISTC of the OM Code permits the use of Appendix II as 

an alternative to other testing or examination provisions of Subsection ISTC. If a licensee elects 

to use Appendix II, the provisions of Appendix II become mandatory in accordance with OM 

Code requirements. The conditions in 10 CFR 50.55a(b)(3)(iv) do not apply to the 

2003 Addendum and later editions and addenda to the OM Code because the 2003 Addendum

revised the earlier OM Code provisions on which this regulation was based to address the 

underlying issues that led the NRC to impose the condition.

The condition in 10 CFR 50.55a(b)(3)(iv)(A) applies to the testing or examination of the check 

valve obturator movement to both the open and closed positions to assess its condition and 

confirm acceptable valve performance. The OM main committee approved the bidirectional 

testing of check valves for inclusion in the 1996 Addendum to the OM Code. The NRC agrees 

with the need for a required demonstration of the bidirectional exercising movement of the check 

valve disk. The single direction flow testing of check valves will not always detect degradation of 

the valve. The classic example of this faulty testing strategy is that separation of the disk would 

not be detected during forward flow tests. The separated disk could be lying in the valve bottom 

or another part of the system and could move to block flow or disable another valve. Appendix II 

did not require bidirectional testing of check valves in the 1996 through 2002 Addenda to the OM 

Code. Hence, the condition in 10 CFR 50.55a(b)(3)(iv)(A) was included so that an Appendix II 

condition monitoring program includes bidirectional testing of check valves to assess their 

condition and confirm acceptable valve performance (as is required by the OM Code).

The condition in 10 CFR 50.55a(b)(3)(iv)(B) applies to the length of the check valve test interval. 

Appendix II would permit a licensee to extend check valve test intervals without limit. A policy of 

prudent and safe interval extension dictates that any interval extension must be based on 

sufficient experience to justify the additional time. Condition monitoring and current experience

may qualify some valves for an initial extension, whereas the trending and evaluation of the data 

may dictate reduction in the testing interval for some valves. Extensions of IST intervals must 

consider plant safety and be supported by the trending and evaluation of both generic and 

plant-specific performance data to ensure that the component is capable of performing its 

intended function over the entire IST interval. Thus, the condition in 10 CFR 50.55a(b)(3)(iv)(B) 

limits the time between the initial test or examination and the second test or examination to two 

fuel cycles or 3 years (whichever is longer), with additional extensions limited to one fuel cycle. 

The total interval is limited to a maximum of 10 years. An extension or reduction in the interval 

between tests or examinations would have to be supported by trending and evaluation of 

performance data.
4-3

The condition in 10 CFR 50.55a(b)(3)(iv)(C) applies to a licensee who discontinues a condition 

monitoring program when using the 1995 Edition of the OM Code with the 1996 and 

1997 Addenda. A licensee who discontinues the use of Appendix II is required to implement the 

requirements of Subsections ISTC 4.5.1 through ISTC 4.5.4 of the OM Code. 

The condition in 10 CFR 50.55a(b)(3)(iv)(D) applies to a licensee who discontinues a condition 

monitoring program when using the 1998 Edition through the 2002 Addendum to the OM Code. 

A licensee who discontinues the use of Appendix II is required to implement the applicable 

provisions in Subsection ISTC.

4.4 10 CFR 50.55a(b)(3)(v)—Subsection ISTD

This condition provides requirements for the examination and testing of snubbers. The 

condition in 10 CFR 50.55a(b)(3)(v) allows licensees using editions and addenda up to the 

2005 Addendum of the ASME B&PV Code Section XI, to optionally use Subsection ISTD of the 

OM Code in place of the requirements for snubbers in Section XI. This condition also states 

that snubber preservice and inservice examinations must be performed using the VT-3 visual 

examination method when using Subsection ISTD of the OM Code. The NRC imposed the 

VT-3 visual examination requirement to ensure that licensees use an appropriate visual 

examination method for the inspection of integral and nonintegral snubber attachments, such as 

lugs, bolting, and clamps, when using Subsection ISTD. 

Licensees that use the 2006 Addendum and later editions and addenda to Section XI of the 

ASME B&PV Code must follow the requirements of Subsection ISTD of the OM Code for 

snubbers because snubber inservice examination and testing requirements have been deleted 

from the scope of Section XI in the 2006 Addendum. The condition in 10 CFR 50.55a(b)(3)(v) 

does not invoke the VT-3 visual examination requirement when licensees use the 2006 

Addendum and later editions and addenda to Section XI because ASME revised Figure 

IWF-1300-1 in the 2006 Addendum to Section XI to clarify that integral and nonintegral snubber 

attachments are within the scope of Section XI. Therefore, the visual examination method 

specified in the 2006 Addendum and in later editions and addenda to ASME B&PV Code, 

Section XI applies to the examination of integral and nonintegral snubber attachments. 

4.5 10 CFR 50.55a(b)(3)(vi) —Exercise Interval for Manual Valves

This condition requires that manual valves must be exercised on a 2-year interval rather than

the 5-year interval specified in paragraph ISTC-3540 of the 1999 through 2005 Addenda to the 

OM Code, provided that adverse conditions do not require more frequent testing. The 1998 

Edition and earlier versions of the OM Code specified an exercise interval of 3 months for 

manual valves. The 1999 Addendum to the OM Code revised paragraph ISTC-3540 to extend 

the exercise frequency for manual valves to 5 years; however, the NRC staff did not agree that 

there was sufficient justification to extend the exercise interval for manual valves to 5 years (See 

Federal Register notice 67 FR 60520, 60531-32 (dated September 26, 2002). The condition in 

10 CFR 50.55a(b)(3)(vi) does not apply to the 2006 Addendum to the OM Code because ASME 

revised the exercise interval in paragraph ISTC-3540 of the 2006 Addendum to the OM Code to 

2 years for manually-operated valves.
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5. GUIDANCE FOR DEVELOPING AND IMPLEMENTING IST 

PROGRAMS

5.1 Alternative Requests

The OM Code establishes the requirements for preservice testing and IST and the examination 

of certain components to assess their operational readiness in light-water reactor nuclear power 

plants. These requirements apply to safety-related pumps, valves, pressure relief devices, and 

snubbers. The requirements are constantly being reviewed and improved in order to meet the 

basic function of maintaining the safe and reliable operation and maintenance of nuclear power 

plants.

It is understood that not all plants are designed the same. It is also understood that the general 

requirements developed in the OM Code may not be applicable or that complying with these 

requirements may be difficult. Licensees may use proposed alternatives to the OM Code 

provided that (1) the alternative would provide an acceptable level of quality and safety under 

10 CFR 50.55a(a)(3)(i); or (2) compliance with the specified requirements would result in 

hardship or unusual difficulty without a compensating increase in the level of quality and safety 

under 10 CFR 50.55a(a)(3)(ii). Hardships generally involve reductions in radiation exposure to

as low as reasonably achievable, challenges to operators or plant equipment, components that 

are somewhat unique in design such as jockey pumps (waterleg), or systems where pump flow 

is fixed and cannot be adjusted. 

Licensees must not implement proposed alternatives to the OM Code requirements under 

10 CFR 50.55a(a)(3) until the NRC staff completes its evaluation. For example, if a licensee 

proposes to implement a pump vibration program based on the use of spectral analysis rather 

than the OM-Code-specified method, the licensee must continue to meet the OM Code

requirements until the NRC staff completes its evaluation.

5.2 Relief Requests

The regulation at 10 CFR 50.55a(f)(4) requires licensees to test pumps and valves in the IST 

program to the “extent practical” within the limitations of the design, geometry, and materials of 

construction. The regulations at 10 CFR 5.55a(f)(5)(iii)–(iv) and 10 CFR 50.55a(f)(6)(i) use the 

term “impractical” instead of “extent practical.” The terms “extent practical” and “impractical” 

apply to test requirements in the OM Code that licensees cannot perform due to the design, 

geometry, and materials of construction of the pump or valve. For example, ASME OM Code, 

Subsection ISTC-5131, “Valve Stoke Testing,” requires that the limiting stroke time for power 

operated valves be specified by the licensee and measured within limits based on the full-stroke 

time of the valves. At some plants, the scram discharge volume vent and drain valves are not 

designed to be individually actuated. These valves are required by TSs to close within a 

specified time (45 seconds for some plants) upon receipt of a scram signal. The valves are 

tested quarterly by cycling the valves to ensure operability and performing a valve sequence 

response time test during each refueling outage. This testing performed is essentially a design 

basis test of the valve combination. Requiring these valves to be stroke timed individually is 

impractical and place a burden on the licensee because of the extensive modification that would 

be required to the system to individually stroke the valve. In addition, jumpering the control 

circuit during plant operation to test these valves individually would be impractical because of 
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the potential for a high reactor scram. Some licensees may have difficulty fully implementing 

these ISTC-5131 required tests, and, in certain cases, because of the impracticality of 

implementation, a request for relief under 10 CFR 50.55a(f)(5) would be appropriate.

In accordance with the regulations, when updating a program to a later edition of the OM Code, 

licensees must implement the updated program at the beginning of a 120-month interval. The

regulations state that in cases in which a licensee determines that an OM-Code-specified pump 

or valve test is impractical and is not included in the revised IST program, it must submit a relief 

request demonstrating the basis for its determination to the NRC no later than 12 months after 

the previous 120-month interval ends, or 12 months after the current interval starts. However, 

experience has shown that licensees also identify impractical test provisions throughout the 

interval. In such cases, licensees may request relief as soon as they identify the condition. 

Because the requirements are impractical, the licensee would test the applicable components 

using the method proposed in the relief request in the period of time from the beginning of the 

new interval (or from the time of identification) until the NRC staff completes its evaluation. 

5.3 Voluntary Use of Later Editions of and Addenda to the ASME OM Code

Licensees must conduct ISTs during successive 120-month intervals to verify the operational 

readiness of pumps and valves within the scope of the OM Code in accordance with 

10 CFR 50.55a(f)(4)(ii). In conducting these ISTs, licensees must comply with the provisions of 

the latest edition and addendum to the OM Code, which 10 CFR 50.55a(b) incorporates by 

reference, 12 months before the start of the 120-month interval, subject to the limitations and 

modifications listed in 10 CFR 50.55a(b). 

The regulation at 10 CFR 50.55a(f)(4)(iv) states that ISTs for pumps and valves may meet the 

requirements set forth in subsequent editions and addenda to the OM Code that 

10 CFR 50.55a(b)(3) incorporates by reference, subject to NRC approval. The regulation at

10 CFR 50.55a(g)(4)(iv) states that the inservice examination of components may meet the 

requirements set forth in subsequent editions and addenda of the OM or Section XI Code that 

10 CFR 50.55a(b) incorporates by reference, subject to NRC approval. This includes the 

examination and testing of snubbers. Licensees may use portions of editions or addenda 

provided that all related requirements of the respective editions or addenda are met. When 

planning to use editions and addenda to the OM or Section XI Code that have not been 

incorporated by reference in the regulations, licensees must request authorization to use these 

later editions and addenda as an alternative to the regulations under 10 CFR 50.55a(a)(3).

The amount of written documentation needed for a request to use a later OM Code edition and 

addendum that 10 CFR 50.55a(b) incorporates by reference is significantly less than that 

necessary for other types of requests to use an alternative approach. For example, licensees 

are not required to provide specific justification for requests to use later OM Code editions and 

addenda that 10 CFR 50.55a(b) incorporates by reference because the NRC has reviewed and 

accepted the provisions of those OM Code editions or addenda, with any appropriate 

modifications or limitations conditions, as part of the process for incorporation of the edition and

addenda by reference in the regulations. If a licensee uses portions of a later OM Code edition 

and addendum, it must ensure that all related requirements of the respective editions and 

addenda are met. The licensee should discuss the related requirements in its letter to the NRC. 

The regulations do not specify when the licensee should submit the letter, only that it should

submit the letter before it uses the later OM Code edition and addendum. The staff issued 
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Regulatory Issue Summary (RIS) 2004-12, “Clarification on Use of Later Editions and Addenda 

to the ASME OM Code and Section XI,” dated July 28, 2004, and RIS 2004-16, “Use of Later 

Editions and addenda to the ASME Code section XI for repair/Replacement Activities,” dated 

October 19, 2004, in order to clarify this matter.

5.4 Identification of Code Noncompliance

The attachment to Regulatory Information Summary (RIS) 2005-20, Revision 1, “Revision to 

NRC Inspection Manual Part 9900, Technical Guidance, Operability Determinations & 

Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to 

Quality or Safety,” dated April 16, 2008, includes guidance on resolving degraded and 

nonconforming conditions. Section 6.2 of the Inspection Manual (IM) Part 9900 notes that a 

licensee may discover a noncompliance with a regulation (ASME OM Code requirement). The 

noncompliance should be treated as a degraded or nonconforming condition. The licensee 

should assess the operability or functionality of the affected component and determine if the 

noncompliance raises an immediate safety issue. Immediate action, such as shutting down the 

plant, might not be required unless otherwise specified by NRC requirements, or by the 

associated operability determination. If the Code noncompliance is for an SSC described in 

TSs and the SSC is determined to be inoperable by the operability determination, then the 

component must be declared inoperable and the associated Limiting Condition for Operation 

(LCO) must immediately be declared not met. 

The most common examples of an ASME non-compliance that can result in a TS violation are 

either a missed IST test required by TS Surveillance Requirement (SR) or when components fail 

or are declared inoperable by the OM Code tests. In cases where the components are declared 

inoperable but the ASME Code required action range or limiting values are more conservative 

than the TS or Safety Analysis Report (SAR) limits, the corrective action may be an analysis to 

demonstrate the specific performance degradation does not impair operability and the pump or 

valve will still perform its safety function. These actions would be accomplished in accordance 

with in IM Part 9900 and the applicable edition and addenda of the OM Code. 

The inoperable component could be declared operable once the NRC authorizes the alternative

and the licensee has successfully completed the alternative test (if applicable). An NRC 

authorization of the alternative would not be retroactive because the agency must authorize the 

alternative before it can be implemented.

5.5 Revising NRC-Authorized Alternatives

RG 1.187, “Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments,”

dated November 2000 provides guidance related to use of 10 CFR 50.59 process. This 

10 CFR 50.59 process does not allow the licensee to change NRC approved or authorized relief 

request or alternative. The NRC must authorize any change to an NRC-authorized 

10 CFR 50.55a(a)(3) alternative unless the requirements of the ASME Code can be met. For 

example, many licensees created a technical requirement manual (TRM) to control certain 

provisions relocated from TSs. Licensees relocated snubber examination and testing 

requirements from the TSs to the TRM. The TRM requirements are controlled using the criteria 

in 10 CFR 50.59, “Changes, Tests and Experiments.” The regulations at 10 CFR 50.59 require 

licensees to evaluate proposed changes to their facilities for the effects of these changes on the 

licensing basis of the plant, as described in the final safety analysis report (as updated) and to 
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obtain prior NRC approval for changes that meet specified criteria as having a potential impact 

upon the basis for the issuance of the operating license. In the case of snubber examination 

and testing, the NRC has authorized the use of the TRM snubber examination and testing 

requirements in lieu of the ASME code requirements at numerous operating plants. The NRC 

authorized the use of the requirements contained in the TRM as an alternative to the ASME 

code requirements. The use of an alternative as authorized by the NRC becomes a regulatory 

requirement; therefore, the NRC staff must review and approve changes to these requirements 

under 10 CFR 50.55a(a)(3).

Nuclear Energy Institute (NEI) 96-07, “Guidelines for 10 CFR 50.59 Implementation,” 

Revision 1, dated November 2000, states that licensees’ activities that are controlled by the 

regulations at 10 CFR 50.55a take precedence over 10 CFR 50.59. RG 1.187, “Guidance for 

Implementation of 10 CFR 50.59, Changes, Tests, and Experiments,” issued November 2000, 

endorses NEI 96-07, Revision 1. Similarly, Section D, “Implementation,” of RG 1.187 states that 

10 CFR 50.59 cannot be used in those cases in which a licensee proposes an acceptable 

alternative method for complying with the specified portion of the NRC’s regulation.

5.6 NRC Verbal Authorization of an Alternative Request

On rare occasions, the NRC may grant verbal authorizations as an alternative under 

10 CFR 50.55a(a)(3) when, because of unforeseen circumstances, licensees need NRC 

authorization before the agency is able to issue its written safety evaluation. Temporary verbal 

authorization for an alternative under 10 CFR 50.55a(a)(3) is subject to the following:

• The proposed alternative is in writing, and all the information that the NRC requires to 

complete the safety evaluation has been docketed.

• An identified need for the verbal authorization is recognized given the circumstances of 

the licensee’s request.

• The NRC has completed its review and determined that the proposed alternative is 

technically justified, but the agency has not yet formally documented it in a safety 

evaluation.

• The technical branch and reactor licensing branch chiefs have agreed to the verbal 

authorization.

Verbal authorization is most likely conveyed in a telephone conversation. As such, appropriate 

NRC personnel who are normally involved in authorizing the alternative must be present in the 

telephone conversation. The NRC project manager should promptly (i.e., in 1 or 2 days) 

generate a record of the conversation; this record will meet the definition of an official agency 

record and must be entered into the NRC’s Agencywide Documents Access and Management 

System (ADAMS) and made publicly available. The NRC should issue the final written 

authorization within 150 days after giving verbal authorization.
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5.7 Submittal of Previously NRC-Authorized Alternatives

Licensees occasionally submit alternative requests that are very similar to NRC-authorized 

alternative requests for the previous 10-year IST intervals when updating their IST program in

accordance with 10 CFR 50.55a(f)(4)(ii). This practice is acceptable provided that the licensee

compares the requirements between the old and new OM Codes and evaluates whether 

changes to the alternative request are necessary. For example, the OM Code has new 

provisions added for exercising check valves such as disassembly and condition monitoring 

programs. Addressing the check valve disassembly and condition monitoring programs in the 

alternative request may be appropriate if these provisions were not included in the OM Code 

upon which the original alternative request was based. Furthermore, the addition of 

disassembly and condition monitoring programs to the OM Code may eliminate the need for the 

alternative request.

Licensees should also review new code cases before submitting an alternative request for 

updated IST programs. For example, Code Case OMN-9, “Use of Pump Curve for Testing,” 

provides an alternative method for testing centrifugal and vertical line shaft pumps when the 

licensee is unable to obtain a specific reference value in accordance with Subsection ISTB of 

the OM Code. The NRC conditionally approved Code Case OMN-9 in RG 1.192, Revision 0. 

The Code Case OMN-16, “Use of Pump Curve for Testing,” incorporates all the condition 

specified for approval of Code Case OMN-9 and published in 2006 addendum of the ASME OM 

Code. The use of Code Case OMN-9 or OMN-16 may eliminate the need for an alternative 

request.

5.8 Standard Format for Requests

To improve the effectiveness and efficiency of the request process, NEI developed a white 

paper entitled, “Standard Format for Requests from Commercial Reactor Licensees Pursuant to 

10 CFR 50.55a,” Revision 1, dated June 7, 2004 (ADAMS Accession No. ML070100400). The 

guidance provided to licensees is voluntary. Other details can be obtained from NRC document 

NUREG/CR-6396, “Examples, Clarifications, and Guidance on Preparing Requests for Relief 

from Pump and Valve Inservice Testing Requirements.”

5.9 ASME OM Code Interpretations

Users of the OM Code may submit technical inquires to the OM main committee. Technical 

inquiries include requests for revisions to present code requirements or new or additional code 

requirements, requests for code cases, and requests for code interpretations. Code 

interpretations provide the meaning or the intent of existing requirements in the OM Code, and 

after appropriate committee deliberations and balloting, ASME issues responses as 

clarifications, new or additional code requirements, or a code erratum. Licensees should 

exercise caution when applying interpretations because they are not specifically incorporated by 

reference into 10 CFR 50.55a and have not received NRC review and approval. The NRC 

recognizes that ASME is the official interpreter of the OM Code, but the NRC will not accept 

ASME interpretations that, in the NRC’s opinion, are contrary to the agency’s requirements or 

may adversely impact facility operations. 
5-6

On November 12, 1996, representatives from ASME and the NRC met to discuss continuing 

questions from the industry about ASME interpretations. It was agreed that the NRC should not 

establish a formal method for reviewing interpretations for acceptance. It was agreed that any

concerns that the NRC has on specific ASME code interpretations would be brought to ASME’s 

attention through the NRC staff’s normal interaction with the Code committee. 

An example of an OM Code interpretation that helped to clarify ASME OM Code provisions is

Interpretation 01-18, “ASME OM Code-1995 with OMa ASME Code-1996 Addenda, Appendix 

I,” dated June 26, 2003. Subparagraph I-1.1.3(a), “5-Year Test Interval,” of Appendix I to the 

1995 Edition with the 1996 Addendum to the OM Code requires that Class 1 pressure relief 

valves be tested at least once every 5 years. Interpretation 01-18 clarifies that the 5-year test 

interval starts when the valve is tested. Therefore, licensees must test Class 1 pressure relief 

valves every 5 years (test-to-test) unless the NRC has authorized a different interval in 

accordance with 10 CFR 50.55a(a)(3), or unless a different interval is specified in a code case 

approved for use by the NRC in accordance with 10 CFR 50.55a(b)(6). 

5.10 Risk-Informed Inservice Testing

RG 1.175, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: Inservice Testing,”

issued August 1998, describes an acceptable alternative approach for applying risk insights 

from probabilistic risk assessment, in conjunction with established traditional engineering 

information, to make changes to a nuclear power plant’s IST program. The approach described 

in RG 1.175 addresses the high-level safety principles specified in RG 1.174, “An Approach for 

Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to 

the Licensing Basis,” Revision 2, issued May 2011, and is intended to define an acceptable 

process for developing risk-informed IST programs without being overly prescriptive. The

resultant risk-informed IST programs will have improved effectiveness with regard to the use of 

plant resources while still maintaining acceptable levels of quality and safety. However, 

licensees may propose other approaches for consideration by the NRC staff. The approach 

presented in RG 1.175 should be regarded as an example of acceptable practices and that 

licensees should have some degree of flexibility in satisfying regulatory requirements on the 

basis of their accumulated plant experience and knowledge. As discussed in RG 1.175, 

licensees proposing to implement a risk-informed IST program are required to submit a request 

to implement an alternative to the ASME OM Code in accordance with 10 CFR 50.55a(a)(3). 

5.11 Use of Code Cases

Code cases are typically formatted in terms of an “Inquiry” (as from a licensee) and “Reply” (by 

the applicable ASME code committee). Oftentimes, the “Inquiry” and “Reply” are accompanied 

by an “Applicability” statement that identifies the specific editions and addenda to the OM Code 

to which the “Reply” applies. However, the OM Code committees do not always (1) include an 

“Applicability” statement or (2) identify all of the editions and addenda to which the “Reply”

applies. As a result, several licensees have questioned whether they could use certain code 

cases without additional interaction with the NRC.

If a licensee would like to use a code case with an edition or addendum to the OM Code to 

which it is not applicable, the licensee has the one of the following options:
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• The licensee can request the alternative to use the code case beyond its stated 

applicability and have the NRC authorize the code case under 10 CFR 50.55a(a)(3).

• If the code case is applicable to an edition or addendum of the OM Code that is later 

than the version of the OM Code being used by the licensee, the licensee could update 

to the later version of the OM Code under 10 CFR 50.55a(f)(4)(iv) and then use the code 

case, provided that the code case has been approved for use in RG 1.192. Note that 

the later version of the OM Code must also have been incorporated by reference into 

10 CFR 50.55a, that the licensee must update all related requirements of the respective 

edition or addendum, and that the NRC must specifically approve the use of the update.

Licensees should not use code cases with editions and addenda to the OM Code to which they 

do not apply and that are not specifically incorporated by reference in 10 CFR 50.55a(b).

Note: In the ASME OM Code 2009 Edition, the expiration of Code Cases has been deleted. 

Therefore, in future the NRC staff will review the 2009 Edition of the ASME OM Code for 

incorporation by reference in 10 CFR 50.55a. At that time any published Code Case that has 

not been annulled, and approved in RG-1.192 may continue to use. (Note: The details related 

to the ASME OM-2009 for information only)
6-1

6. GUIDANCE FOR SPECIFIC PUMP AND VALVE IST ACTIVITIES

6.1 Waterleg Pumps

The NRC has received proposed alternatives from licensees of boiling-water reactors (BWRs)

for Group A tests for waterleg pumps. Subsection ISTB-3400 and Table ISTB-3400-1 of the 

OM Code specify that a Group A test must be performed quarterly for Group A pumps. The 

waterleg pumps are low flow pumps that are required to operate whenever their respective 

emergency core cooling system (ECCS) trains are in the operable condition. As such, the 

pumps perform continuous duty on a recirculation line and provide makeup as needed. There is 

typically no flow instrumentation of the recirculation line, and the flow instrumentation on the 

main ECCS header is not sufficiently accurate to measure the low flow of the pumps. When 

requesting an alternative Group A test for a waterleg pump, a licensee should explain how it 

monitors pump discharge pressure, how it verifies that the main ECCS header is full of water, 

what is the pump vibration monitoring frequency, and any other maintenance or testing activitiy 

performed to ensure the pump will continue to meet its’ intended function.

6.2 Smooth-Running Pumps

Pumps that have very low vibration reference values (less than or equal to 0.05 inch per 

second) are called smooth-running pumps. A small increase in smooth-running pump vibration 

during the OM Code-required IST causes the pump to exceed OM Code vibration acceptance 

criteria, which normally results in unnecessary corrective action. The NRC has authorized 

alternative vibration acceptance criteria for smooth-running pumps on a case-by-case basis in 

accordance with 10 CFR 50.55a(a)(3). 

Alternative requests for smooth-running pumps should specify a minimum vibration reference 

value (≤ 0.05 inch per second), and these smooth-running pumps must be included in a 

predictive maintenance (PdM) program. The importance of the PdM program for smoothrunning pumps was demonstrated when a plant using NRC-authorized alternative vibration 

acceptance criteria noted a bearing failure that was not detected by the IST program, but was 

detected through enhanced vibration monitoring as part of the plant’s PdM program. During 

IST, corrective action was not required because the measured vibration was below the alert 

range as specified by the Code. After the pump bearing failed, it became clear that a simple 

minimum vibration reference value alone is not sufficient to identify degradation of a smoothrunning pump. PdM programs normally include bearing temperature trending, oil sampling and 

analysis, thermographic analysis, and enhanced vibration monitoring. The objective of the PdM 

program should be to detect and correct problems involving the mechanical condition of the 

pump before the pump reaches its overall vibration alert limit.

6.3 Vibration-Measuring Transducers

Subsection ISTB of the OM Code requires that the frequency response range of vibrationmeasuring transducers and their readout system be from one-third of the minimum pump shaft 

rotational speed to at least 1,000 hertz (Hz). Licensees have proposed alternatives to this OM 

Code requirement in accordance with 10 CFR 50.55a(a)(3) for pumps with low shaft rotational 

speeds. Similar alternative requests submitted by licensees have been withdrawn following 

discussion with the NRC. The proposed alternatives state that the procurement and calibration 
6-2

of vibration-measuring transducers and their readout systems for the lower end of the OM 

Code-specified range were hardships because of the limited number of vendors supplying such 

equipment, the level of equipment sophistication, and equipment cost. The NRC typically 

authorized these alternative requests in the past. However, vibration-measuring transducers 

and their readout system can now be procured from various suppliers at a reasonably low cost 

due to technology advancement and research work performed in the field of instrumentation. 

Therefore, licensee requests to use this alternative are generally no longer authorized by the 

NRC.

6.4 Online Check Valve Sample Disassembly and Inspection

Licensees have proposed, as an alternative to ISTC-5221(c) and ISTC-5224, to perform sample 

disassembly and inspection of check valves in a group online. Subsection ISTC of OM Code,

Paragraph ISTC-3510, requires that check valves be exercised every 3 months. Paragraph 

ISTC-3522(c) states that if exercising is not practicable during operation at power and cold 

shutdown, it shall be performed during refueling outages. ISTC-5221(c) allows disassembly of 

check valves every refueling outage as an alternative means to verify their operability. Instead 

of disassembly every refueling outage, ISTC-5221(c) provides the option of using a sample 

disassembly and inspection program for groups of identical valves in similar applications. 

Further, ISTC-5221(c)(3) states that at least one valve from each group shall be disassembled 

and examined at each refueling outage and all valves in each group shall be disassembled and 

examined at least once every 8 years. ISTC-5224 requires that check valves in a sample 

disassembly program that are not capable of being full-stroke exercised or have failed or have 

unacceptably degraded valve internals, shall have the cause of failure analyzed and the 

condition corrected. 

ISTC-5224 also states that other check valves in the sample group that may also be affected by 

this failure mechanism be examined or tested during the same refueling outage to determine the 

condition of internal components and their ability to function. A licensee should fully describe 

how it plans to comply with the requirements in ISTC-5224 when submitting alternative requests 

for check valve group sample disassembly and inspection online. The plan description also 

should include information on management of examination and testing of all group valves 

should a scheduled valve inspection be declared inoperable. For example, licensees should 

explain how the disassembly and inspection of the other check valves in a group will be 

completed within the allowed system outage time.

6.5 Safety/Relief Valves (S/RVs)

Many licensees have requested and obtained NRC authorization in accordance with 

10 CFR 50.55a(a)(3) to use the provisions in Code Case OMN-17, “Alternative Rules for Testing 

ASME Class 1 Pressure Relief/Safety Valves,” as an alternative to the 5-year test interval 

specified in the OM Code. Code Case OMN-17 allows an extension of the test frequency for 

safety relief valves (S/RVs) from 60 months to 72 months plus a 6-month grace period. The 

code case imposes a special maintenance requirement to disassemble and inspect each valve 

to verify that parts are free from defects resulting from the time-related degradation or 

maintenance-induced wear before the start of the extended test frequency. Although the OM 

Code does not require that S/RVs be routinely refurbished, refurbishment provides reasonable 

assurance that the S/RVs are operationally ready during the extended test interval. ASME 

published Code Case OMN-17 in the 2009 Edition of the OM Code. The NRC will consider 
6-3

including Code Case OMN-17 in a future revision to RG 1.192. The Code Case OMN-17 will 

not be acceptable for use until it is included in the RG 1.192. Until that time, licensee need to 

submit a request to NRC to use OMN-17.

In recent years, the NRC staff has received numerous requests for relief or TS changes or both

related to the stroke testing requirements for BWR dual-function main steam S/RVs. The 

2003 Addendum and earlier editions and addenda to Mandatory Appendix I to the OM Code 

require the stroke testing of S/RVs after they are reinstalled following maintenance activities. A 

number of licensees have determined that in situ testing of the S/RVs can contribute to 

undesirable seat leakage of the valves during subsequent plant operation and have received 

approval to perform stroke testing at a laboratory facility coupled with in situ tests and other 

verifications of actuation systems as an alternative to the testing required by the OM Code. The 

revised subparagraph I-3410(d) in Mandatory Appendix I to the 2004 Edition of the OM Code 

does not require licensees to stroke test S/RVs at reduced or normal system pressure following 

maintenance. Subparagraph I-3410(d) in the 2004 Edition of the OM Code requires that each 

S/RV that has been removed for maintenance or testing and reinstalled shall have the electrical 

and pneumatic connections verified either through mechanical/electrical inspection or testing

before the resumption of electric power generation. Several licensees have requested and 

obtained NRC approval in accordance with 10 CFR 50.55a(f)(4)(iv) to use Subparagraph 

I-3410(d) of the 2004 Edition of the OM Code in place of Subparagraph I-3410(d) of the 2001 

Edition through the 2003 Addenda to the OM Code. 
7-1

7. SNUBBER INSERVICE EXAMINATION AND TESTING PROGRAMS

7.1 Program Controls

Some licensees have incorrectly interpreted that the examination and testing of snubbers is not 

a 10 CFR 50.55a requirement because (1) 10 CFR 50.55a(g) addresses components (including 

supports) without mentioning snubbers, (2) snubber examination and testing was historically 

covered by TS, and (3) TS allow snubber examination and testing requirements to be relocated 

from the TS to the TRM. Licensees have the option to control the inservice examination and 

testing of snubbers through their TS or other licensee-controlled documents. For plants using 

their TS to govern the inservice examination and testing of snubbers, 10 CFR 50.55a(g)(5)(ii) 

requires that if a revised ISI program for a facility conflicts with the TS, the licensee shall apply 

to the Commission for the amendment of the TS to conform the TS to the revised program. 

Therefore, when performing 120-month program updates in accordance with 

10 CFR 50.55a(g)(4), licensees must submit any required amendments or any alternative

requests to ensure that their TS remain consistent with the new ISI program. The TS, TRM, or 

other licensee-controlled documents governing the snubber inservice examination and testing 

program do not eliminate the 10 CFR 50.55a requirements to update the program at 120-month 

intervals in accordance with 10 CFR 50.55a(g)(4) or to request and receive NRC authorization 

for alternatives to the Code requirements when appropriate. The NRC issued Regulatory Issue 

Summary (RIS) 2010-06, “Inservice Inspection and Testing Requirements of Dynamic 

Restraints (Snubbers)” on June 1, 2010 to remind all the licensees of the NRC’s rules and 

regulations regarding snubber ISI and testing, in accordance with 10 CFR 50.55a, at nuclear 

power facilities.

7.2 OM Part 4 Clarification

The NRC has noted that the relocation of the reference to ASME/ANSI Operation and 

Maintenance of Nuclear Power Plants (OM Part 4) from IWF-5000 of Section XI to Table 

IWF-1600-1 of Section XI has created confusion regarding which edition and addenda of 

OM Part 4 must be used. For clarification, the ASME OM Code and OM Part 4 are two different 

ASME documents. Article IWF-5000 of the 1987 Addendum through the 1992 Edition of 

Section XI requires that the inservice examination and testing of snubbers be accomplished in 

accordance with the 1987 Edition of OM Part 4. The reference to OM Part 4 was deleted from 

IWF-5000 in the 1992 Addendum of Section XI and there is no reference to OM Part 4 in 

IWF-5000 in the 1992 through 2005 Addenda of Section XI. The reference for the applicable 

edition and addenda of OM Part 4 was moved to Table 1600-1 in the 1992 Addendum of 

Section XI. Although IWF-5000 in the 1992 through 2005 Addenda of Section XI no longer 

references OM Part 4, Table 1600-1 of the 1992 through 2005 Addenda of Section XI requires 

that inservice examination and testing of snubbers be performed in accordance with the 1987 

Edition and 1988 Addendum of OM Part 4. 
8-1

8. IST PROGRAM GUIDANCE FOR NEW REACTORS

The nuclear industry has submitted applications for licenses to construct and operate new 

nuclear power plants. The NRC discusses policy and technical issues associated with new 

reactors, including the development of IST programs, in several Commission papers, such as

SECY-90-016, 93-087, 94-084, and 95-132, and their applicable Staff Requirements 

Memoranda (SRMs). In a public memorandum to file dated July 24, 1995, the NRC staff 

consolidated the discussion of the policy and technical issues associated with the regulatory 

treatment of non-safety systems (RTNSS) in new passive plant designs provided in 

SECY-94-084 and SECY-95-132, and their associated SRMs. 

The NRC regulations in 10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear 

Power Plants,” include design certifications for specific new reactor designs, such as the 

Advanced Boiling Water Reactor (ABWR) and AP1000 PWR. In addition, reactor vendors have 

submitted applications for, or updates to, certification of designs for several new reactors. For 

example, the NRC is reviewing design certification applications for the Economic Simplified 

Boiling Water Reactor (ESBWR), Evolutionary Power Reactor (EPR), and U.S Advanced 

Pressurized Water Reactor (US-APWR). The NRC is also reviewing an amendment to the 

AP1000 certified design. The NRC regulations require new reactor vendors to address design 

aspects related to the IST program in their design certification application. In that significant 

flexibility exists during the design phase, new reactor vendors should design their plants to 

minimize the need for requests for relief from the IST provisions in the ASME OM Code. 

Several nuclear power plant organizations have submitted applications under 10 CFR Part 52 

for combined licenses (COLs) to construct and operate new nuclear power plants that reference 

certified reactor designs or designs under certification review. COL applicants are required by 

10 CFR Part 52 to submit a description of the IST program to be developed for the new nuclear 

power plants. Regulatory Guide 1.206 (June 2007), “Combined License Applications for 

Nuclear Power Plants (LWR Edition),” provides guidance for the preparation of COL 

applications for new reactors. In RG 1.206, the NRC staff discusses the information that should 

be provided by COL applicants to fully describe their IST program to avoid the need for 

programmatic inspections, tests, analyses, and acceptance criteria (ITAAC). In addition to 

addressing design aspects related to the IST program, new reactor vendors typically provide a 

description of generic aspects of the IST program to allow the COL applicants to incorporate by 

reference this design certification information in their COL application. The NRC staff reviews 

the description of the IST program in the COL application, with its incorporation by reference of 

IST provisions in the applicable design certification documentation, as part of the safety 

evaluation for the COL application. The NRC staff will conduct inspections of the development 

and implementation of the IST program following COL issuance.

ASME has a program underway to establish improved IST provisions in the ASME OM Code for 

pumps, valves, and dynamic restraints to be used in new reactors. ASME has prepared a 

White Paper that discusses its plans to update the ASME OM Code for new reactors. For 

example, the ASME White Paper identifies lessons learned from operating experience at current 

nuclear power plants, and from research sponsored by the nuclear industry and the NRC, that 

are applicable to IST programs for new reactors. The ASME White Paper also identifies new 

reactor issues that can affect IST programs to be developed for new reactors. 
8-2

Lessons learned from nuclear power plant operating experience and research that should be 

considered in the development of IST programs for new reactors include, for example:

1. Design and qualification of pumps, valves, and snubbers to allow IST activities 

(including sufficient flow testing) to assess the operational readiness of those 

components, including ASME Standard QME-1-2007, “Qualification of Active 

Mechanical Equipment Used in Nuclear Power Plants,” as accepted in Revision 3 to 

RG 1.100, “Seismic Qualification of Electric and Active Mechanical Equipment and 

Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” to 

incorporate lessons learned in the qualification of mechanical equipment for nuclear 

power plants.

2. Performance and testing of Motor-Operated Valves (MOVs) that indicate the need for 

improved MOV activities, such as importance of adequate design and qualification, 

sufficient flow during testing to assess valve performance, consideration of MOV 

performance parameters (including valve disk and stem friction coefficients, reduced 

voltage, elevated temperature, and load sensitive behavior), use of adequate diagnostic

instrumentation to allow proper evaluation and setup, improved maintenance and 

personnel training, monitoring of potential motor magnesium rotor degradation, and 

justification for motor control center testing.

3. Application of MOV lessons learned to other Power-Operated Valves (POVs).

4. Provisions for bi-directional testing of check valves.

5. Implementation of pre-service testing (PST) and comprehensive pump testing (CPT) 

provisions without the need for relief from the ASME OM Code provisions.

6. Consideration of potential adverse flow effects on plant components from flow-induced 

vibration resulting from hydrodynamic loads and acoustic resonance.

New reactor issues that should be considered in the development of IST programs for new 

nuclear power plants include, for example:

1. Description of IST programs by COL applicants in accordance with 10 CFR Part 52 with 

implementation of Design Certification provisions for design, qualification, and IST 

activities.

2. Coordination of PST and ITAAC so that testing is performed once for both purposes. 

For example, implementation of PST requirements and the new Part 52 ITAAC closure 

and maintenance process both need to be accomplished. Under the new Part 52 

process, an applicant is required to meet OM Code requirements after a 52.103(g) 

finding is made although it would be preferable to complete the PST requirements 

earlier.

3. Design, qualification, and IST and inspection activities for pyrotechnic-actuated (squib)

valves that have high safety significance, and represent more significant engineering 

challenges for new reactors than for current operating plants.
8-3

4. Design of plant systems and development of IST programs to minimize the need for 

relief from the ASME OM Code provisions.

5. Design, qualification, PST, and IST activities for RTNSS components that perform 

safety significant functions.

6. Development and implementation of risk-informed IST programs, including 

10 CFR 50.69 programs, for new reactors.

7. Consideration of appropriate Code and standard modifications for design, qualification, 

PST and IST activities in response to application of software-based digital technology in 

mechanical components (e.g., pumps and valves).

Applicants for new nuclear power plants should consider the information in this NUREG and 

other sources, such as the ASME program to update the OM Code, in developing their IST 

programs. 
9-1

9. REFERENCES

9.1 American Society of Mechanical Engineers/American National Standards Institute, 

“Code for Operation and Maintenance of Nuclear Power Plants,” New York, NY.

9.2 American Society of Mechanical Engineers/American National Standards Institute, 

“Operation and Maintenance of Nuclear Power Plants,” New York, NY.

9.3 American Society of Mechanical Engineers/American National Standards Institute, 

NQA-1, “Quality Assurance Program Requirements for Nuclear Facilities,” Washington, 

DC, and New York, NY, 1979.

9.4 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section 

XI, “Rules for Inservice Inspection of Nuclear Power Plant Components,” New York, NY. 

9.5 Nuclear Energy Institute, NEI-96-07, “Guidelines for 10 CFR 50.59 Implementation,”

Revision 1, Dated November 2000, Washington, DC.

9.6 Nuclear Energy Institute White Paper, “Standard Format for Requests from Commercial 

Reactor Licensees Pursuant to 10 CFR 50.55a,” Revision 1, Washington, DC, 

June 7, 2004.

9.7 NRC Inspection Manual Part 9900, “Technical Guidance” 

9.8 The National Technology Transfer and Advancement Act of 1995 (P.L. 104-113)

9.9 U.S. Code of Federal Regulations, “Domestic Licensing of Production and Utilization 

Facilities,” Part 50 and Part 52, Chapter 1, Title 10, “Energy.”

9.10 U.S. Nuclear Regulatory Commission, Generic Letter 89-10, “Safety-Related MotorOperated Valve Testing and Surveillance,” dated June 28, 1989.

9.11 U.S. Nuclear Regulatory Commission, Generic Letter 96-05, “Periodic Verification of 

Design-Basis Capability of Safety-Related Power-Operated Valves,” dated September 

18, 1996.

9.12 U.S. Nuclear Regulatory Commission, NUREG-1482, “Guidelines for Inservice Testing 

at Nuclear Power Plants,” Revision 0, Washington, DC, April 1995.

9.13 U.S. Nuclear Regulatory Commission, NUREG-1482, “Guidelines for Inservice Testing 

at Nuclear Power Plants,” Revision 1, Washington, DC, January 2005.

9.14 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.147, “Inservice Inspection 

Code Case Acceptability ASME Section XI, Division 1,” Washington, DC, October 2010.

9.15 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.175, “An Approach for 

Plant-Specific, Risk-Informed Decisionmaking: Inservice Testing,” Washington, DC,

August 1998.

.
9-2

9.16 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, “An Approach for Using 

Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to 

the Licensing Basis,” Revision 2, Washington, DC, May 2011.

9.17 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.187, “Guidance for 

Implementation of 10 CFR 50.59, Changes, Tests, and Experiments,” Washington, DC,

November 2000.

9.18 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.192, “Operation and 

Maintenance Code Case Acceptability: ASME OM Code,” Washington, DC, June 2003.

9.19 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.193, “ASME Code Cases Not 

Approved for Use,” Washington, DC, October 2010.

9.20 U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2004-12, “Clarification 

on Use of Later Editions and Addenda to the ASME OM Code and Section XI,” 

Washington, DC, July 28, 2004.

9.21 U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2005-20, “Revision to 

NRC Inspection Manual Part 9900 Technical Guidance, ‘Operability Determinations & 

Functionality Assessments for Resolution of Degraded or Nonconforming Conditions 

Adverse to Quality or Safety,’” Revision 1, Washington, DC, April 16, 2008.

9.22 Regulatory Issue Summary (RIS) 2010-06, “Inservice Inspection and Testing 

Requirements of Dynamic Restraints (Snubbers), dated June 1, 2010. 

9.23 Regulatory Issue Summary (RIS) IS 2004-16, “Use of Later Editions and addenda to the 

ASME Code section XI for repair/Replacement Activities,” dated October 19, 2004. 

9.24 Regulatory Guide 1.206, “Combined Licensee Application for Nuclear Power Plants 

(LWR Edition),” dated June 2007.

9.25 Regulatory Guide 1.100, “seismic Qualification of Electric and Mechanical Equipments 

for Nuclear Power Plants,” dated September 2009.

9.26 QME-1, “Qualification of Active Mechanical Equipments used in Nuclear Power Plants,” 

Edition 2007.

9.27 SECY-90-016, “Evolutionary Light Water Reactor (LWR) Certification Issues and Their 

Relationship to Current Regulatory Requirements.”

9.28 SECY-93-087, “Policy, Technical, and Licensing Issues Pertaining to Evolutionary and 

Advanced Light-Water Reactor (ALWR) Designs.”

9.29 SECY-94-084, “Policy and Technical Issues Associated with the Regulatory Treatment 

of Non-Safety Systems (RTNSS) in Passive Plant Designs.”

9.30 SECY-95-132, “Policy and Technical Issues Associated with the Regulatory Treatment 

of Non-Safety Systems (RTNSS) in Passive Plant Designs (SECY-94-084).”
A-i

Draft NUREG-1482, Rev-2

APPENDIX A

GUIDELINES FOR INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 

AT NUCLEAR POWER PLANTS

(Previously NUREG-1482, Revision 1 with updates)
A-ii

CONTENTS

APPENDIX A

Page

ABBREVIATIONS A-vi

1 INTRODUCTION A 1-1

1.1 Regulatory Basis A 1-1

1.2 Regulatory History A 1-1

1.3 NRC Recommendations and Guidance A 1-3

1.4 Synopsis of Report A 1-3

2 DEVELOPING AND IMPLEMENTING AN INSERVICE TESTING PROGRAM A 2-1

2.1 Compliance Considerations A 2-1

2.1.1 ASME Code Case Applicability A 2-3

2.2 Scope of Inservice Testing Programs A 2-4

2.2.1 Basis for Scope Requirements A 2-4

2.2.2 Examples of Omitted Components A 2-6

2.2.3 Testing of Non-Code Components A 2-7

2.2.4 Commitments to Include Components in IST Programs A 2-7

2.3 Code Class Systems Containing Safety-Related Pumps and Valves A 2-8

2.4 IST Program Document A 2-8

2.4.1 Pumps A 2-9

2.4.2 Valves A 2-10

2.4.3 Piping and Instrument Diagrams A 2-11

2.4.4 Bases Document A 2-12

2.4.5 Deferring Valve Testing to Cold Shutdown or Refueling Outages A 2-12

2.5 Relief Requests and Proposed Alternatives A 2-13

2.5.1 Justifications for Relief A 2-15

2.5.2 Categories of Relief Requests A 2-16

2.5.3 Content and Format of Relief Requests A 2-16

2.6 Program Documents A 2-18

2.7 Developing IST Programs for New Nuclear Power Plants A 2-19

3 GENERAL GUIDANCE ON INSERVICE TESTING A 3-1

3.1 Inservice Test Frequencies and Extensions for Valve Testing A 3-1

3.1.1 Deferring Valve Testing to Each Cold Shutdown or 

Refueling Outage A 3-2

3.1.2 Entry into a Limiting Condition for Operation (LCO) to Perform 

Testing A 3-6

3.1.3 Scheduling of Inservice Tests A 3-7

3.2 Start of the Time Period in Technical Specification Action Statements A 3-8

3.3 120-Month Updates Required by 10 CFR 50.55a(f)(4)(ii) A 3-8

3.3.1 Extension of Interval A 3-9

3.3.2 Concurrent Intervals A 3-10

3.3.3 Implementation of Updated Programs A 3-11

3.3.4 General Comments on Inservice Testing Intervals A 3-12 
A-iii

APPENDIX A

Page

3.4 Skid-Mounted Components and Component Subassemblies A 3-13

3.5 Pre-Conditioning of Pumps and Valves A 3-14

3.5.1 Background A 3-14

3.5.2 NRC Guidance A 3-15

3.5.3 ASME Code Guidance A 3-17

3.5.4 NRC Recommendation A 3-17

3.6 Testing in the As-Found Condition A 3-18

3.7 Testing at Power A 3-18

3.8 Potential Adverse Impact on Plant Components from Flow-Induced 

Vibration A 3-20

3.9 Repair of Pumps and Valves for New Reactors under Construction A 3-20

4 SUPPLEMENTAL GUIDANCE ON INSERVICE TESTING OF VALVES A 4-1

4.1 Check Valves A 4-1

4.1.1 Closure Verification for Series Check Valves without 

Intermediate Test Connections A 4-2

4.1.2 Exercising Check Valves with Flow and Nonintrusive 

Techniques A 4-4

4.1.3 Full Flow Testing of Check Valves A.4-5

4.1.4 Disassembly and Inspection Alternative to Flow Testing A 4-7

4.1.5 Reverse Flow Testing of Check Valves A-4-8

4.1.6 Extension of Test Interval to Refueling Outage for Check 

Valves Verified Closed by Leak Testing A-4-11

4.1.7 Testing and Examination of Check Valves Using Manual 

Mechanical Exercisers A-4-12

4.1.8 Check Valve Bidirectional Testing and Condition Monitoring

Program A 4-13

4.1.9 Instrumentation Requirements A 4-14

4.1.10 Skid-Mounted Valves and Component Subassemblies A 4-15

4.1.11 Check Valves in New Reactors A 4-15

4.2 Power-Operated Valves A 4-16

4.2.1 Stroke-Time Testing for Power-Operated Valves A 4-16

4.2.2 Stroke-Time Measurements for Rapid-Acting Valves A 4-17

4.2.3 Stroke Time for Solenoid Operated Valves A 4-18

4.2.4 Supplement to the POVs Stroke-Time Test Provision of the 

ASME Code A 4-19

4.2.5 Alternatives to POVs Stroke-Time Testing A 4-21

4.2.6 Main Steam Isolation Valves A 4-23

4.2.7 Verification of Remote Position Indication for Valves 

by Methods Other Than Direct Observation A 4-24

4.2.8 Requirements for Verifying Position Indication of Passive Valves A 4-26

4.2.9 Control Valves with a Safety Function A 4-27

4.2.10 Pressurizer Power-Operated Relief Valve Inservice Testing A 4-27
A-iv

APPENDIX A

Page

4.2.11 POVs in New Reactors A 4-28

4.3 Safety and Relief Valves A 4-29

4.3.1 Scope A 4-29

4.3.2 Method of Testing Safety and Relief Valves A 4-30

4.3.3 Jack-and-Lap Process A 4-30

4.3.4 Maintenance and Inspection of Safety and Relief Valves

in Addition to OM Code Requirements A 4-31

4.3.5 Scheduling of Safety and Relief Valve Testing A 4-31

4.4 Miscellaneous Valves A 4-31

4.4.1 Post-Accident Sampling System Valves A 4-32

4.4.2 Post-Maintenance Testing After Stem Packing Adjustments 

and Backseating of Valves to Prevent Packing Leakage A 4-32

4.4.3 Manual Valves A 4-35

4.4.4 Pressure Isolation Valves A 4-3

4.4.5 Containment Isolation Valves That Have Other Leak-Tight

Safety Function(s) A 4-37

4.4.6 Testing Individual Scram Valves for Control Rods in

Boiling-Water Reactors A 4-38

4.4.7 Use of Appendix J, Option B, in Conjunction with ISTC

Exercising Tests A 4-39

4.4.8 Pyrotechnic-Actuated Valves in New Reactors A 4-40

5 SUPPLEMENTAL GUIDANCE ON INSERVICE TESTING OF PUMPS A 5-1

5.1 General Pump Inservice Testing Issues A 5-1

5.1.1 Categories of Pumps for Inservice Testing A 5-1

5.1.2 Testing Requirements and Frequency of Inservice Tests A 5-1

5.2 Use of Variable Reference Values for Flow Rate and Differential Pressure

During Pump Testing A 5-2

5.2.1 Reference Values A 5-2

5.2.2 Reference Curves A 5-3

5.2.3 Effect of Pump Replacement, Repair, and Maintenance 

on Reference Values or Reference Curves A 5-3

5.2.4 Establishment of Additional Sets of Reference Values or

Reference Curves A 5-3

5.3 Allowable Variance from Reference Points and Fixed-Resistance

Systems A 5-4

5.4 Monitoring Pump Vibration in Accordance with ISTB A 5-7

5.5 Pump Flow Rate and Differential Pressure Instruments A 5-8

5.5.1 Range and Accuracy of Analog Instruments A 5-9

5.5.2 Use of Tank Level to Calculate Flow Rate for Positive

Displacement Pumps A 5-10

5.5.3 Use of Tank or Bay Level to Calculate Differential Pressure A 5-11

5.5.4 Accuracy of the Flow Rate Instrument Loop A 5-12

5.6 Operability Limits of Pumps A 5-12
A-v

APPENDIX A

Page

5.7 Duration of Tests A 5-12

5.8 Adjustments for Instrument Inaccuracies A 5-13

5.9 Pump Testing Using Minimum Flow Return Lines With or

Without Flow Measuring Devices A 5-13

5.10 Alternative to ASME OM Code Comprehensive Pump Testing

Requirements A 5-14

5.11 Motor Drivers for Pumps A 5-15

5.12 Pumps in New Reactors A 5-16

6 STANDARD TECHNICAL SPECIFICATIONS A 6-1

7 IDENTIFICATION OF CODE NONCOMPLIANCE A 7-1

7.1 Nonconforming Conditions A 7-1

7.2 Starting Points for Time Periods in Technical Specification

Action Statements A 7-2

8 RISK-INFORMED INSERVICE TESTING A 8-1

8.1 Introduction A 8-1

8.2 Discussion A 8-1

8.3 Online Inservice Testing A 8-1

8.4 ASME Risk-Informed Code Cases A 8-3

9 HISTORICAL REFERENCES A 9-1

Tables

2.1 Typical Systems and Components in an Inservice Testing Program for a

Pressurized-Water Reactor A 2-20

2.2 Typical Systems and Components in an Inservice Testing Program for a

Boiling-Water Reactor A 2-24

2.3 Example Data Table for Pumps A 2-28

2.4 Useful Abbreviations for Valve Data Tables A 2-29

3.1 ASME OM Code Tests and Test Frequencies for Pumps and Valves A 3-1

3.2 ASME OM Code Terms for Inservice Testing Activities A 3-7

Figures

2.1 Flow Chart – Development of Inservice Testing Program for Pumps and Valves A 2-30
A-vi

ABBREVIATIONS

ADS automatic depressurization system

ALARA as low as reasonably achievable

ANSI American National Standards Institute

AOV air-operated valve

ASME American Society of Mechanical Engineers

BEP best efficiency point

BWR boiling-water reactor

BWST borated water storage tank

CFR Code of Federal Regulations

CPT comprehensive pump test

CRD control rod drive

CVCS chemical and volume control system

DBD design-basis document

ECCS emergency core cooling system

FR Federal Register

FSAR final safety analysis report

GDC General Design Criterion

GE General Electric Company

GL generic letter

GSI generic safety issue

HCU hydraulic control unit

HOV hydraulic-operated valve

HPCI high-pressure coolant injection

IEEE Institute of Electrical and Electronics Engineers

IN information notice

IP inspection procedure

ISI inservice inspection

IST inservice testing

JOG Joint Owners Group

LCO limiting condition for operation

LOCA loss-of-coolant accident

LWR light-water reactor

MOV motor-operated valve

MSIV main steam isolation valve

MSSV main steam safety valve

NEI Nuclear Energy Institute (formerly NUMARC)
A-vii

NIC Nuclear Industry Check Valve Group

NOED Notice of Enforcement Discretion

NRC U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation (NRC)

NUMARC Nuclear Management and Resources Council (now NEI)

OM Operations and Maintenance

PASS post-accident sampling system

PIV pressure isolation valve

PORV power-operated relief valve

POV power-operated valve

PRA probabilistic risk assessment

PTC Performance Test Code

PWR pressurized-water reactor

P&ID piping and instrument diagram

RCIC reactor core isolation cooling

RCPB reactor coolant pressure boundary

RCS reactor coolant system

RG regulatory guide

RHR residual heat removal

RIS regulatory issue summary

RWST refueling water storage tank

RWT refueling water tank

SAR safety analysis report

SBLC standby liquid control

SGCV Code Committee Sub-Group on Check Valves (ASME)

SI safety injection

SOV solenoid-operated valve

SR surveillance requirement

SRP Standard Review Plan

SSC system, structure, and/or component

S/RV safety/relief valve

STS Standard Technical Specifications

TS technical specification(s)

UFSAR updated final safety analysis report

WGC Working Group Committee (IEEE)

WGCV Working Group on Check Valves (ASME)
A 1-1

1. INTRODUCTION.

1.1 Regulatory Basis

Title 10, Section 50.55a, of the Code of Federal Regulations (10 CFR 50.55a) defines the 

requirements for applying industry codes and standards to boiling- or pressurized-water-cooled 

nuclear power facilities. Each of these facilities is subject to the conditions in paragraphs (a), 

(f), and (g) of 10 CFR 50.55a, as they relate to inservice inspection (ISI) and inservice testing 

(IST). By rulemaking effective September 8, 1992 (see Federal Register, Vol. 57, No. 3152, 

p. 34666, dated August 6, 1992), the U.S. Nuclear Regulatory Commission (NRC) established 

paragraph (f) of 10 CFR 50.55a to separate the IST requirements from the ISI requirements in 

paragraph (g).

The NRC regulations in 10 CFR 50.55a(b)(3) incorporate by reference the 1998 Edition through 

the 2004 Edition with 2005 and 2006 addendas of the Code for Operation and Maintenance of 

Nuclear Power Plants (OM Code) promulgated by the American Society of Mechanical 

Engineers (ASME), in which Subsection ISTA provides general IST requirements and 

Subsections ISTB, ISTC, and ISTD provide specify the IST requirements for pumps, valves, and 

dynamic restraints, respectively. Based on those requirements, each of the NRC’s nuclear 

power plant licensees must establish IST programs, specify the components included in the 

program as well as the test methods and frequencies for those components, and implement the 

program in accordance with the OM Code.

Where a test requirement of the OM Code is determined to be impractical for a facility, the

NRC’s regulations 10 CFR 50.55a(f)(6)(i) allow the licensee to submit a request for relief from 

the given requirement, along with information to support the determination. Relief requests 

generally detail the reasons for deviating from the Code requirements and propose alternative 

testing methods or frequencies. The Commission is authorized to evaluate licensees’ relief 

requests, and may grant the requested relief or impose alternative requirements, considering 

the burden that the licensee might incur if the Code requirements were enforced for the given 

facility. Pursuant to 10 CFR 50.55a(a)(3)(i) and (ii), the Commission may also authorize the 

licensee to implement an alternative to the Code requirements, provided that the alternative 

ensures an acceptable level of quality and safety or the Code requirement presents a hardship 

without a compensating increase in the level of quality and safety.

The regulations in 10 CFR 50.55a(f)(4)(iv) specify that inservice testing of pumps and valves 

may meet the requirements in editions and addenda of the OM Code that were published more 

recently than those that are incorporated by reference in 10 CFR 50.55a(b), subject to 

Commission approval and the limitations and modifications listed in 10 CFR 50.55a(b).

Requests for approval to use later editions and addenda previously incorporated by reference in 

10 CFR 50.55a may be made via letter to the NRC. See NRC Regulatory Issue Summary (RIS) 

2004-12, “Clarification on Use of Later Editions and Addenda to the ASME OM Code and 

Section XI,” for clarification. 

1.2 Regulatory History

The NRC previously issued guidance for implementing IST requirements. After publishing the

rule that established the IST requirements (see Federal Register, Vol. 41, No. 30, p. 6256, 

dated February 12, 1976), the NRC sent letters to notify operating licensees of the new rule. 
A 1-2

In November 1976, after receiving inquiries from licensees regarding acceptable methods for

complying with the regulation, the NRC issued letters to licensees to transmit NRC Staff 

Guidance for Complying with 10 CFR 50.55a(g), “Inservice Inspection Requirements.”

To eliminate the backlog of IST program reviews for operating nuclear power plants, the NRC 

issued Generic Letter (GL) 89-04, “Guidance on Developing Acceptable Inservice Testing 

Programs,” dated April 3, 1989. That generic letter included 11 technical positions that the staff 

uses in reviewing licensees’ IST program relief requests and described alternatives to the Code 

requirements that the staff considered acceptable. In GL 89-04, the staff also approved 6 of the

11 technical positions (1, 2, 6, 7, 9, and 10), pursuant to 10 CFR 50.55a(g)(6)(i), with the

provision that the licensee must perform the alternative testing delineated in the applicable 

position. The staff approved these alternatives upon recognizing that it might be impractical to 

perform the required testing, and enforcing the requirements might pose an unnecessary 

burden on licensees. 

GL 89-04 stated that licensees must document their uses of Positions 1, 2, 6, 7, 9, and 10 in the

IST program, but did not require the documentation to take the form of a relief request. 

Provided that the licensee followed the provisions of GL 89-04, the generic letter granted 

approval to follow the alternative testing delineated in Positions 1, 2, 6, 7, 9, and 10, pursuant to 

10 CFR 50.55a(g) [now (f)]. For convenience, most licensees documented their uses of these 

positions in relief requests; however, other forms of program documentation were acceptable as 

long the provisions of the referenced positions were clearly documented and discussed in 

adequate detail to substantiate conformance with those provisions. Certain licensees may 

have submitted relief requests to ensure that their conformance was adequately documented in 

their IST program, even though documentation in the program would also have been 

acceptable, as stated in GL 89-04.

The staff held four public meetings to discuss GL 89-04 and stated that the generic letter was a

first step toward resolving various problems associated with developing and implementing IST 

programs at nuclear power plants. The staff had previously identified these problems through 

its reviews of licensees’ IST programs, and by inspecting and auditing IST programs at plant 

sites, participating on the ASME Code committees, and meeting with licensees and industry 

groups. The staff then summarized the questions and answers from the four public meetings in 

a letter entitled “Minutes of the Public Meetings on Generic Letter 89-04,” dated October 25, 

1989. That letter contained useful information about how to apply the guidance in GL 89-04 and

discussed issues of interest to licensees who attended the public meetings. In a subsequent 

letter, dated September 16, 1991, the staff issued “Supplement to Minutes of the Public 

Meetings on Generic Letter 89-04” to address a question on stop-check valve testing.

In the 1988 and 1989 Addenda to Section XI of the OM Code, ASME and the American National 

Standards Institute (ANSI) revised Subsections IWP and IWV to simply reference Parts 6 and 

10 of the 1987 Edition of the ASME OM Code (OM-1987, OM-6 and OM-10, respectively). 

ASME/ANSI also rewrote the OM standards (albeit without making any significant technical 

changes) and, in 1990, the Board on Nuclear Codes and Standards subsequently approved the

revised standards as the “Code for Operation and Maintenance of Nuclear Power Plants, ASME 

OM Code-1990.” The OM-1990 edition of the Code includes pump, valve, and snubber IST 

requirements, as well as requirements for snubber examination.
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Since the NRC issued GL 89-04, the staff has improved its guidance regarding IST by revising 

10 CFR 50.55a and separating the IST and ISI programs in Paragraphs (f) and (g), respectively, 

issuing additional guidance, and coordinating with ASME for periodic symposia on testing 

pumps and valves. The NRC intends to continue to improve its IST-related guidance through 

continued participation in Code and technical organizations, as well as regular updates of the 

agency’s published guidance as future needs arise.

1.3 NRC Recommendations and Guidance

The code of record for NUREG-1482, Revision 2 is the ASME OM Code, 1998 Edition through 

the 2004 Edition with 2005 and 2006 addendas. 

The recommendations herein replace the guidance and technical positions in GL 89-04. Note 

that specific relief is required to implement the guidance in GL 89-04. However, relief 

justification may refer to the positions in the GL with clarifying information to clearly show how it 

would apply to a licensee’s situation. To the extent practical, this document reflects the 

applicable section, subsection, or paragraph of the appropriate documents (subsections of 

10 CFR Part 50, OM Code, regulatory guides, etc.).

The guidance presented herein may be used for requesting relief. However, the guidance 

contained herein is strictly voluntary and licensees may also request relief that is not in 

conformance with the guidance. The NRC may reference a recommendation contained in this 

document in future safety evaluations and may grant relief or authorize the alternative if the 

licensee has addressed all of the aspects included in the applicable section, where applicable.

1.4 Synopsis of Report

This appendix follows the format of a typical IST program plan, including Development and 

Implementation, General Guidance, Valves, Pumps, Technical Specifications, Code 

Noncompliance, and Risk-Informed IST.

Section 2, “Developing and Implementing an IST Program,” describes existing IST 

requirements, discusses the scope of the IST program, and describes guidance for presenting 

information in IST programs, including cold shutdown justifications, refueling outage 

justifications, and relief requests. Section 2 also includes a sample list of plant systems for 

boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) that typically (but not 

necessarily) contain pumps or valves that perform a safety function and are subject to 

requirements of the OM Code.

Section 3, “General Guidance on Inservice Testing,” describes the NRC’s recommendations 

and their bases for several general aspects of IST. Sections 4 and 5 then become more 

specific, describing recommendations on valve-related and pump-related issues, respectively. 

Throughout Sections 3 through 5, this document discusses the IST requirements for which 

licensees have requested relief or proposed alternatives. It also provides guidance concerning 

the types of information that licensees typically should (or in some cases must) include in their

relief requests. Sections 3 through 5 also discuss related Code and regulatory issues and 

provide recommendations and guidance as needed. These discussions do not impose 

additional 
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requirements beyond those imposed by the Code or the regulations and, as such, do not 

represent backfits.

Rather, these discussions are intended to clarify the existing requirements of the Code or the

regulations and, as such, they may provide recommendations to ensure that licensees continue 

to meet the Code and other regulatory requirements.

Sections 6, 7, and 8 then discuss the revised standard technical specifications, the process 

licensees should follow when they identify a Code nonconformance, and the development of a

risk-informed IST program. Section 9 presents a list of related references.

The guidance in this document is similar in appearance and content to NRC staff positions given 

in a regulatory guide because it uses terms such as shall, must, or requires to indicate

mandatory considerations; should to state staff recommendations or reiterate previously 

approved NRC staff positions; and recommendation, acceptable to the staff, or licensee may to 

reflect NRC staff findings from previous plant-specific safety evaluations related to IST relief 

requests. The guidance in this document is also similar to that given in a regulatory guide 

because certain recommendations indicate acceptable alternatives to Code requirements. 

However, this guidance is not equivalent to staff positions in a regulatory guide, because it is 

strictly intended for voluntary implementation by licensees. Licensees must request approval 

through the relief or alternative process described in 10 CFR 50.55a where the ASME OM Code 

will not be met.
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2. DEVELOPING AND IMPLEMENTING

AN INSERVICE TESTING PROGRAM

Licensees may use the following guidance for developing and implementing inservice testing 

(IST) programs. This guidance supplements existing requirements and previously approved 

guidance on IST.

2.1 Compliance Considerations

The NRC regulations in 10 CFR 50.55a specifies requirements for IST of certain safety-related 

pumps and valves that must be tested according to the requirements of the ASME OM Code. 

This testing is intended to assess the operational readiness of the stated components. 

Specifically, the regulations state that the tests conducted during the initial and successive 

120-month intervals must be based on the requirements in the applicable edition and addenda 

of the ASME OM Code, to the extent practical, within the limitations of design, geometry, and 

materials of construction, as described in 10 CFR 50.55a(f)(4).

In addition, Paragraph 50.55a(f)(4)(ii) requires that IST conducted during each 120-month 

interval following the initial interval must be conducted in compliance with the requirements of 

the latest edition and addenda of the Code incorporated by reference in the version of 

10 CFR 50.55a(b) that is in effect 12 months before the start of the interval. The NRC 

regulations in 10 CFR 50.55a(b) incorporate by reference the 1995 Edition through the 2004 

Edition including 2005 and 2006 addendas of the ASME OM Code subject to limitations and 

modifications conditions.

The regulations specify the “upper-tier” requirements for IST, and the requirements of the 

OM Code, as incorporated by reference into the regulations, have the force of law. To augment

the regulations, a plant’s Technical Specifications (TSs) include general and specific 

requirements for IST and other surveillance testing of pumps and valves. Similarly, the plant’s 

safety analysis includes information concerning the design limitations and functional 

requirements for the performance of pumps and valves for the given facility. The plant’s IST 

program, including any relief requests and data analysis methods, describes the licensee’s 

means for implementing the various requirements for the specific plant. 

The implementing procedures include the lowest tier of IST elements. In addition, IST

engineers often use other information (such as bases documents, vendor manuals, trend data, 

and graphs) in developing, maintaining, and implementing the plant’s IST program.

The regulations are the authority governing the implementation of the various IST requirements. 

Therefore, licensees must meet the regulations when they find a conflict between the 

regulations and any of the lower-tier requirements (program or procedures). The staff gives 

guidance on cases where a licensee modifies its plant in a way that affects the basis for relief 

that the NRC has previously granted. Similarly, if a licensee has obtained the NRC’s approval 

of an alternative pursuant to 10 CFR 50.55a(a)(3)(i) or (ii), the licensee need not use that

alternative if it subsequently determines that continued compliance with the Code requirements 

is warranted or necessary for particular circumstances that may preclude implementation of the 

approved alternative. When a licensee revises an implementing procedure, the licensee 
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typically ensures that the IST program continues to reflect the required testing. Similarly, when 

a system, subsystem, or component is modified, or an operating or test procedure or valve 

alignment is changed in accordance with 10 CFR 50.59, the licensee typically reviews the IST 

requirements to determine whether it must change the program for the affected components.

IST of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with 

the OM Code and applicable addenda as required by 10 CFR 50.55a. According to the 

regulation, if a revised IST program conflicts with the facility’s technical specifications, the 

licensee shall apply to the NRC to amend the technical specifications to conform with the 

revised program, or otherwise meet the requirements of the technical specifications and 

10 CFR 50.55a (see 10 CFR 50.55a(f)(5)(ii)). This provision in the rule specifies actions to be 

taken by a licensee when a revised inservice inspection (testing) program for a facility conflicts 

with the technical specifications (see 41 FR 6256, “Statements of Consideration,” dated 

February 12, 1976).

The NRC may authorize alternatives to Code testing requirements submitted as relief requests 

or in a similar format that includes a discussion of the requirements, a description of the

proposed alternative, and the justification for approval of the alternative. 10 CFR 50.55a 

includes the following provisions for accepting alternatives or granting relief:

• 10 CFR 50.55a(a)(3)(i) allows the NRC to authorize alternatives if the proposed 

alternatives would provide an acceptable level of quality and safety. The NRC will 

normally approve an alternative pursuant to this provision only if the licensee proposes a 

method of testing that is equivalent to, or an improvement of, the method specified by 

the code, or if the testing will comply or is consistent with later Code editions approved 

by the NRC in 10 CFR 50.55a(b).

• 10 CFR 50.55a(a)(3)(ii) allows the NRC to authorize an alternative if compliance [with 

the Code requirement] would result in hardship or unusual difficulty without a 

compensating increase in the level of quality and safety. The NRC may approve an 

alternative pursuant to this provision if, although the proposed alternative testing does 

not comply with the Code, the increase in overall plant safety and quality attained by 

complying with the Code requirement is not justified in light of the difficulty associated 

with compliance.

• 10 CFR 50.55a(f)(6)(i) includes the following provision:

The Commission will evaluate determinations that Code requirements are impractical. The 

Commission may grant relief and may impose such alternative requirements as it determines is 

authorized by law, giving due consideration to the burden upon the licensee that could result if 

the requirements were imposed on the facility.

The NRC may grant relief pursuant to this provision or may impose alternatives if the licensee 

demonstrates that the design or access limitations make the Code requirement impractical. 

Thus, the staff’s evaluation considers the burden created by imposing the Code requirements 

on the licensee.
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2.1.1 ASME Code Case Applicability

Code Cases are typically formatted in terms of an Inquiry [as from a licensee] and Reply [by the 

applicable ASME Code Committee]. Oftentimes the Inquiry and Reply are accompanied by an 

Applicability statement that identifies the specific Editions an Addenda of the ASME Code to 

which the Reply applies. However, the ASME Code Committees do not always 1) include an 

Applicability statement or 2) identify all of the editions and addenda to which the Reply applies. 

As a result, several licensees have questioned whether they could use certain ASME Code 

Cases without additional interaction with the regulatory authority having jurisdiction at the plant 

site.

If a licensee would like to use an ASME Code Case with an Edition or Addendum of the ASME 

Code to which it is not applicable, the licensee has the following options:

a. Have the alternative to use the Code Case, beyond its stated applicability, authorized by 

the NRC pursuant to 10 CFR 50.55a(a)(3), or

b. If the Code Case is applicable to an Edition or Addendum of the ASME Code later than 

the version of the Code being used by the licensee, the licensee could update to the 

later version of the Code pursuant to 10 CFR 50.55a(f)(4)(iv) or (g)(4)(iv) and then use 

the Code Case, provided the Code Case has been approved for use in the appropriate 

Regulatory Guide and incorporated by reference into 10 CFR 50.55a. Note that the later 

version of the ASME Code must also have been incorporated by reference into 

10 CFR 50.55a, the licensee must update all related requirements of the respective 

Edition or Addenda, and the update must be specifically approved by the Commission.

Licensees should not use ASME Code Cases with Editions and Addenda of the ASME Code to 

which they do not apply and that are not specifically approved for use by the NRC. More 

specifically, licensees should not reconcile the Applicability of Code Cases without consulting 

with the applicable ASME Code Committee.

In incorporating the OM Code by reference in 10 CFR 50.55a, the NRC staff recognized the

need for a new regulatory guide that would approve OM Code cases. Such a regulatory guide 

would perform a function similar to that of existing Regulatory Guide (RG) 1.147, which 

approves ASME Code cases applicable to Section XI of the ASME Boiler and Pressure Vessel 

Code.

Accordingly, the NRC staff developed RG 1.192, “Operation and Maintenance Code Case 

Acceptability, ASME OM Code,” as well as RG 1.193, “ASME Code Cases Not Approved for 

Use.” 

Code Cases OMN-1, OMN-3, OMN-4, OMN-11, and OMN-12 include risk-informed provisions 

that may be applied in IST programs. Regulatory Guide 1.175, “An Approach for Plant-Specific, 

Risk-Informed Decision-Making: Inservice Testing,” describes an acceptable alternative 

approach for applying risk insights from probabilistic risk assessment (PRA), in conjunction with 

established traditional engineering information, to make changes to a nuclear power plant’s IST 

program. The approach described in RG 1.175 addresses the high-level safety principles 

specified in RG 1.174 and attempts to strike a balance between defining an acceptable process 

for developing risk-informed IST 
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programs without being overly prescriptive. Until such time as a risk-informed regulation is

promulgated and included in the regulations, the alternative approach described in RG 1.175

must be authorized by the NRC pursuant to 10 CFR 50.55a(a)(3)(i) on a plant-specific basis 

before being implemented by a given licensee. However, because 10 CFR 50.55a(a)(3)(i) 

places no restrictions on the scope of alternatives that the NRC may authorize, licensees may

propose risk-informed alternatives to their entire IST program or may propose alternatives that 

are more limited in scope (e.g., for a particular system or group of systems, or for a particular 

group of components). However, with the issuance of RG 1.192, licensees may use

risk-informed IST methods without first obtaining NRC staff review and approval. Section 8 of 

this Appendix A discusses risk-informed IST in greater detail.

If RG 1.193 identifies a Code case as being unacceptable, licensees may not implement the 

specified Code case without first obtaining NRC approval. Licensees may request the NRC’s 

approval to implement a Code case listed in the RG under the provisions of 

10 CFR 50.55a(a)(3), which permits the use of alternatives to the Code requirements 

referenced in 10 CFR 50.55a, provided that the proposed alternative results in an acceptable 

level of quality and safety, by addressing the NRC s concern and submitting a plant-specific 

relief request.

2.2 Scope of Inservice Testing Programs

The intent of General Design Criterion (GDC) 1 (defined in Appendix A to 10 CFR Part 50) and 

Criterion XI (defined in Appendix B to 10 CFR Part 50) is that all components (such as pumps 

and valves) that are necessary for safe operation must be tested to demonstrate that they will 

perform satisfactorily in service. Among other things, GDC 1 requires that components that are

important to safety must be tested to quality standards that are commensurate with the

importance of the safety function(s) to be performed. Appendix B to 10 CFR Part 50 describes 

the requisite quality assurance program, which includes testing, for safety-related components. 

In addition, 10 CFR 50.55a(f) requires that licensees must use the ASME OM Code for inservice 

testing of components that are covered by the Code. Each licensee has the responsibility to 

demonstrate the continued operability or functionality of all components within the scope of their 

IST program. The regulatory guides augment those requirements by providing additional NRC 

guidance regarding scope and classification. In short, the ASME Code defines the scope, 

10 CFR 50.55a endorses the Code with clarifications, and regulatory guides provide additional 

guidance.

An IST program, including implementing procedures, is subject to the requirements of 

10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel 

Reprocessing Plants,” and ASME OM Code Subsection ISTA. Changes to the scope, test 

methods, or acceptance criteria should be reviewed to the requirements of 10 CFR 50.59, 

10 CFR 50.55a and 10 CFR 50.65 as appropriate.

The TSs for some plants may include IST requirements that are more restrictive than the 

regulations. Section 6 of this Appendix A describes how licensees may amend their TS 

requirements for IST to better address the regulations as the governing requirements.

2.2.1 Basis for Scope Requirements

The requirements for the scope of components to be included in an IST program are addressed 

in 10 CFR 50.55a(f). Specifically, 10 CFR 50.55a(f)(4) states, “Throughout the service life of a
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boiling- or pressurized-water-cooled nuclear power facility, pumps and valves which are 

classified as ASME Code Class 1, Class 2, and Class 3 must meet the inservice test 

requirements set forth in the ASME OM Code.”

ASME Code Class 1 components include all components within the reactor coolant pressure 

boundary. RG 1.26, “Quality Group Classifications and Standards for Water-, Steam-, and

Radioactive-Waste-Containing Components of Nuclear Power Plants,” Revision 4, dated March 

2007, provides guidelines for establishing the quality group classification (and ASME Code 

classification) for water-, steam-, and radioactive-waste-containing components of nuclear 

power plants, other than those in the reactor coolant pressure boundary (i.e., ASME Code Class 

2 and 3 components).

The ASME OM Code is incorporated by reference in 10 CFR 50.55a(b)(3). The OM Code 

defines the scope by stating that IST programs shall include pumps and valves that are required 

to perform a specific function in (1) shutting down the reactor to a safe shutdown condition, 

(2) maintaining the safe shutdown condition, or (3) mitigating the consequences of an accident. 

The scope of the OM Code also covers pressure relief devices that are used to protect systems 

(or portions of systems) that perform a required safety-related function. Therefore, the scope of

components to be included in an IST program must encompass ASME Code Class 1, 2, and 3 

components that are covered in Subsection ISTA of the ASME OM Code.

Subsection ISTA-1100 of the OM Code refers to components that are “needed to mitigate the

consequences of an accident.” This statement is intended to provide confidence that the health 

and safety of the public will be protected in the event of certain accidents and anticipated 

transients at a nuclear power plant. The term “accident” is also used throughout the 

Commission’s regulations. For example, Appendix B to 10 CFR Part 50 establishes quality 

assurance requirements for the design, construction, and operation of “structures, systems, and 

components that prevent or mitigate the consequences of postulated accidents that could cause 

undue risk to the health and safety of the public.” Similarly, 10 CFR Part 100 describes 

structures, systems, and components that must be designed to remain functional during and 

following a “safe shutdown earthquake” as those necessary to ensure (1) the integrity of the 

reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in 

a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of 

accidents that could result in potential offsite exposures comparable to the guideline exposures.

In establishing such requirements, the Commission uses the term “accident” to describe a broad

range of possible adverse events at a nuclear power plant. Therefore, although most of the 

accidents of concern to IST are addressed in the accident analyses chapter, licensees should 

be aware that the plant’s final safety analysis report (FSAR) may address other accident

analyses that need to be considered within the context of IST.

Thus, an introductory section of the IST program document submitted to the NRC for each plant 

must state the plant’s safe-shutdown condition (i.e., hot standby, hot shutdown, cold shutdown, 

etc.). If the scope in Subsection ISTA appears to be broader than that specified in 

10 CFR 50.55a, the more narrow scope applies.
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Components within the scope of 10 CFR 50.55a are included in the scope of 10 CFR 50.65, 

Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants (the 

Maintenance Rule). Licensees may elect to consolidate testing for pumps and valves, 

designating any non-Code components as such in the IST program.

The plant’s FSAR (or equivalent) defines the equipment that is necessary to meet specific 

functions. If the FSAR indicates that a system or component is Code Class 1, 2, or 3, that 

system or component is within the scope of 10 CFR 50.55a. By contrast, if the FSAR states 

that a system or component is designed, fabricated, and maintained as Code class at the option 

of the Owner as permitted by Subsection ISTA 1320, the application of the related OM Code

requirements is also optional.

Tables 2.1 and 2.2 (which appear at the end of this chapter) provide examples of systems and

components that licensees typically include in their IST programs. These tables are not

intended to be all-inclusive, but they may form the basis for the initial review of a licensee’s IST 

program scope.

Figure 2.1, Flow Chart (which appears at the end of this chapter) provides a quick reference to 

regulatory requirements for development of the inservice testing program for pumps and valves. 

For complete details, see 10 CFR 50.55a. 

2.2.2 Examples of Omitted Components

During IST program reviews and inspections, the staff has noted that licensees do not always 

include the necessary equipment in the scope of their IST programs. Licensees should review 

their IST programs to ensure adequate scope. Components that are frequently erroneously 

omitted from IST programs include the following examples:

(a) BWR scram system valves

(b) control room chilled-water system pumps and valves

(c) accumulator vent valves or motor-operated isolation valves

(d) auxiliary pressurizer spray system valves

(e) boric acid transfer pumps

(f) valves in the emergency boration flow path

(g) control valves that have a required fail-safe position

(h) valves in mini-flow lines

(i) control rod drive (CRD) system check valves

(j) keep fill systems

Licensees should review the safety significance of these components to ensure that their IST is

adequate to demonstrate their continued operability. Licensees should also recognize that the 

pumps and valves listed above do not apply to every plant and do not satisfy the scope required 

by Subsection ISTA for all plants. For example, items c, d, e, and f do not apply to BWRs. 

Each licensee should review the list and determine which items apply to its facility.
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2.2.3 Testing of Non-Code Components

An IST program is also a reasonable vehicle to periodically demonstrate the operability of

pumps and valves that are not covered by the Code. Thus, if a licensee voluntarily chooses to 

include non-Code components in its ASME Code IST program (or some other licenseedeveloped testing program) and, as a result, is unable to meet certain Code provisions, the 

regulations (10 CFR 50.55a) do not require the licensee to submit a relief request to the NRC. 

Nonetheless, the licensee should maintain documentation that provides assurance of the

continued operability or, as required the continued functionality of the non-Code components 

through the performed tests, and such documentation should be available for staff inspection at 

the plant site.

While 10 CFR 50.55a delineates the testing requirements for ASME Code Class 1, 2, and 3 

pumps and valves, licensees may elect to not limit their inservice testing to only those pumps

and valves that are covered by 10 CFR 50.55a. For example, the emergency diesel generator 

air start system is typically not Code Class 1, 2, or 3 and, therefore, 10 CFR 50.55a does not 

require licensees to test the related components under the provisions of the ASME Code. 

However, emergency diesel generator air start, cooling water, and fuel oil transfer systems are 

considered safety-related and, as such, Appendices A and B to 10 CFR Part 50 require that 

they must be included in the scope of a component testing program and must undergo the 

required testing.

Licensees may implement deviations from the Code for non-Code components without NRC

review and approval, and need not document such deviations as “relief requests.” A notation in 

the licensee’s IST program document would help to identify the deviations and clarify that they 

relate to non-Code components. If it is not clear that the deviations relate to non-Code 

components, the staff erroneously might assume that the licensee is not meeting the 

requirements of 10 CFR 50.55a. Some licensees use the relief request format to document 

such deviations, while other licensees place notes, footnotes, or brief descriptions in their 

program documents.

2.2.4 Commitments to Include Components in IST Programs

If a licensee includes a component in the IST program, the component is considered to be 

within the scope of the program and may only be removed if the licensee meets the applicable 

criteria of 10 CFR 50.55a and 50.59. Similarly, if the TSs require a licensee to test a component 

in accordance with the 10 CFR 50.55a IST program, it is considered to be within the scope.

As a result of the review of design-basis documents (DBDs), many licensees have reassessed 

the scope of their IST programs and considered deleting certain systems from the programs. 

To delete entire systems from the IST program, a licensee would need to perform a review and 

prepare documentation in accordance with 10 CFR 50.59 (if necessary).

Plants licensed under NUREG-0800, the NRC’s “Standard Review Plan for the Review of Safety

Analysis Reports for Nuclear Power Plants” (also known as the SRP), may have classified 

certain systems as Code Class 3, even though those systems would not have been classified as 

such in earlier plants. The SRP recommends (rather than requiring) that licensees should 

classify such systems as Quality Group C (which corresponds to Code Class 3 in RG 1.26). 
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RG 1.26 states that it does not cover systems such as instrument and service air, diesel 

engines, their generators and auxiliary support systems, diesel fuel, emergency and normal 

ventilation, fuel handling, and radioactive waste management systems; however, these systems 

should be designed, fabricated, erected, and tested to quality standards that are commensurate 

with the safety function(s) to be performed.

The licensee is responsible for determining whether a system requires the Code Class 3 

classification, or whether that classification is optional under Subsection ISTA 1320. 

Specifically, Subsection ISTA-1320 states that optional construction of a component within a

system boundary to a classification higher than the minimum class established in the

component design specification shall not affect the overall system classification by which

applicable rules are determined. Thus, if a licensee changes the code classification pursuant to 

10 CFR 50.59, the pumps and valves may remain as “augmented components” (denoted as 

non-Code) in the IST program. (Note that NRC approval of a licensee amendment may be 

necessary, as determined by the evaluation conducted in accordance with 10 CFR 50.59.)

2.3 Code Class Systems Containing Safety-Related Pumps and Valves

The plant safety analysis report (SAR), technical specifications (TSs), and other documents list

the systems and components that are necessary to function to support the safe operation and 

shutdown of the plant. Tables 2.1 and 2.2 (which appear at the end of this chapter) list systems 

and components that are typically included in the IST programs for pressurized-water reactors 

(PWRs) and boiling-water reactors (BWRs). These tables are not intended to apply to all plants; 

the listed systems and components are not considered safety-related at every plant, and are not

necessarily classified as Code Class 1, 2, or 3. (For information on quality group and Code 

classifications, see RG 1.26 and Section 3.9.6 of NUREG-0800.) The licensee’s safety analysis 

generally contains a section describing the Code classification of components. The IST 

program scope should be developed to be consistent with the SAR.

2.4 IST Program Document

Within this discussion of the IST program document, Section 2.4.1 applies to pumps, while

Section 2.4.2 applies to valves. These sections describe the information that licensees 

generally need to prepare the related sections of the IST program document.

The OM Code includes the rules for inservice testing of nuclear power plant components:

• Subsection ISTA includes general requirements for testing components.

• Subsection ISTA-3200 states that the owner shall file IST plans with the regulatory 

authorities having jurisdiction at the plant site. 

• Subsection ISTA-9000 addresses the records and reports that are required for these 

inspection and testing programs.

• Subsection ISTA-9210 states that the owner shall prepare plans for preservice, and 

inservice examinations and tests to meet the requirements of the OM Code.
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• Subsection ISTA-9220 states that licensees shall prepare examination, test, 

replacement, and repair records in accordance with the requirements of the respective 

articles of the OM Code.

• Subsection ISTB-9000 and ISTC-9000 include additional guidance for the information 

that the IST program document must include for pumps and valves that perform safety 

functions.

• Nonmandatory Appendix A, “Preparation of Test Plans,” and Supplement to Appendix A 

to the ASME OM Code gives voluntary guidance for licensees to use in preparing their 

inspection and test plans.

Licensees have found that pump and valve tables are a convenient format for the information. 

These tables typically include sufficient information to allow NRC inspectors to determine 

whether the testing complies with the Code requirements for test method and frequency. The

tables could also note applicable NRC positions or recommendations for each pump or valve.

It is intended that the IST program should reflect design modifications and other activities 

performed under 10 CFR 50.59 that relate to pumps and valves within the scope of the 

program. Thus, the staff recommends that the program plan submitted to the NRC should 

include documentation of the use of positions contained herein, and Code cases.

2.4.1 Pumps

In preparing pump tables, licensees should consider the following information, which includes 

headings and a description of the text that licensees could include under each heading.

Title: List the applicable plant and unit.

Page number: Include the page number and total number of pages in the program document or

the relevant section, such as “Page 15 of 135.”

Program revision or revision date: List the program or page revision number and date (on each 

page). List the revision number for each program change submitted.

System, Code class, and group: List the plant system, Code class, and valve group, and briefly

describe the service of the pump.

Pump identification: List a unique identifier for each pump; this identifier should be used 

consistently in all IST program documentation and design information such as system piping 

and instrument diagrams (P&IDs), test procedures, and relief requests.

Piping and instrument diagram (P&ID) number: List the applicable P&ID or figure that depicts 

the pump in the system.

Drawing coordinates: List the coordinates of the pump on the P&ID.
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Test parameters: List each of the five parameters in Tables ISTB-5121-1, ISTB-5221-1, 

ISTB-5321-1, and ISTB-5321-2 for each pump. A column or a footnote is typically used to list 

factors that affect testing. List a relief request or Code case number where the testing will not 

be performed in accordance with the Code. Notes can be used where Code testing would 

otherwise be required. A relief request is not required if the test requirement is exempted by the 

Code.

Relief request(s): List any applicable relief requests in the pump table. Table 2.3 is an example 

of a data table for pumps.

2.4.2 Valves

In preparing valve tables, licensees should consider the following information, which includes 

headings and a description of the text that licensees could include under each heading. Table

2.4 lists common abbreviations used in valve data tables.

Title: List the applicable plant and unit.

Page number: Include the page number and total number of pages in the program document or

the relevant section, such as “Page 15 of 135.”

Program revision or revision date: List the program or page revision number and date (on each

page). List the revision number for each program change submitted.

System, Code class, and group: List the plant system, Code class, and valve group, and briefly

describe the service of the valve.

Valve identification: List a unique identifier for each pump; this identifier should be used 

consistently in all IST program documentation and design information such as P&IDs, test 

procedures, and relief requests. If valves such as excess flow check valves are grouped 

together in the table, the number of valves and the valve number must be clearly indicated.

P&ID number: List the applicable P&ID or figure that depicts the valve in the system.

Drawing coordinates: List the coordinates of the valve on the P&ID.

Valve type: List the valve type (i.e., gate, globe, check, relief).

Valve size: Specify the valve size in inches, fractions of an inch, or metric units.

Actuator type: List the type of valve actuator (i.e., motor, solenoid, pneumatic, hydraulic, self) 

with the type and function of each valve.

Code category: Specify the Code category (or categories), as defined in Subsection

ISTC-1300. This determines the applicable subsections of the Code. For example, a 

motor-operated gate valve could be in Code Category A or B, while a self-actuated check valve 

could be in Category C or A/C.

Active/Passive: State whether a valve is active or passive, as defined in Subsection ISTC-2000. 

Requirements vary based on the function of the valve. A valve need not be considered active if 
A 2-11

it is only temporarily removed from service or from its safety position, such as manually opening 

a sample valve for a short time to take a sample, while maintaining administrative control over 

the valve. If the plant is in an operating mode that does not require a passive valve to be 

maintained in its “passive” (safety) position, the position of the valve may be changed without 

imposing IST requirements on the valve. By contrast, if a valve is routinely repositioned during 

power operations (or has an active safety function), it is an active valve. If a valve is 

repositioned to create a new alignment (e.g., as a corrective action for a condition of another 

valve in the line), an evaluation (considering the impact on the IST program) may be required to 

ensure operational readiness before positioning the valve in a new position, as determined on a 

case-by-case basis.

Safety position: List the safety function position(s), specifying both positions for valves that 

perform a safety function in both the open and closed positions. Valves must be exercised to

the position(s) required to fulfill their safety function(s). Check valve tests must include both

open and close tests.

Tests performed: Specify which tests are to be performed on each valve.

Test frequency: List the actual frequency for each test to be performed. If it would be 

impractical or burdensome to perform the test at the frequency specified in the Code, reference

cold shutdown or refueling outage justifications or relief requests for the alternative test 

frequency.

Relief requests and cold shutdown/refueling outage justifications: List any applicable relief 

request(s). In addition, when the testing is deferred to cold shutdowns or refueling outages, 

reference the technical justification (cold shutdown justification or refueling outage justification) 

for the test frequency.

Remarks: Include any pertinent information that is not stated elsewhere in the table such as a

brief functional description of the valve. 

2.4.3 Piping and Instrument Diagrams

The staff recommends that licensees’ program submittals should include P&IDs or system 

drawings to assist in locating the pumps and valves that are included in the program and such 

drawings should be the latest revision at the time the program is submitted to the NRC. This 

information will assist the staff in reviewing relief requests or proposed alternatives. Inservice 

inspection boundary system drawings and isometrics, or reduced-size drawings, are suitable for

inclusion in the program document. If the reduced-size drawings are not complete P&IDs, the 

staff may request a set of full-size drawings for use in evaluating relief requests. A partial 

submittal of the program containing relief requests could include applicable drawings to support 

the relief requests or to supersede previous IST program drawings. Licensees need not update 

their program drawings regularly, but if drawings change because of modifications, or if the 

changes affect relief requests, the staff recommends that licensees should revise and resubmit 

the drawings in the next periodic submittal of revisions to the program document. The staff also 

recommends that licensees should include applicable drawings with relief request submittals 

that are very detailed and are submitted to supplement the IST program. Such drawings are 

helpful because the NRC’s technical staff who review relief requests do not maintain a set of 

SARs for each plant and do not receive a copy of the IST program plan (which generally 

contains the applicable drawings). 
A 2-12

Drawings are helpful in reviewing relief requests, regardless of whether they are submitted as 

part of the program document or as an attachment applicable to any relief request or proposed 

alternative.

2.4.4 Bases Document

The staff recommends that each licensee should create a bases document for the IST program. 

A paper discussing the creation and management of a bases document is included in

Supplement 1 to NUREG/CP-0123, “Proceedings of the Second NRC/ASME Symposium on

Pump and Valve Testing,” dated November 1992. Bases documents have typically included a 

description of the methodology used in preparing the IST program, with a list of each pump and 

valve in a system within the boundaries for a Code class, the basis for including (or excluding) 

the pump or valve, and the basis for the testing applied to each component.

Although not required by the NRC, the bases document will help licensees ensure the continuity 

of their IST programs when the responsibilities of personnel or groups change. A bases 

document will also enable the plant staff to clearly understand the reasons that the components 

are either in the program or not, as well as the basis for testing (or not testing) certain functions. 

Although not necessarily a “licensing-basis document,” the bases document is a useful 

reference for reviews performed under 10 CFR 50.59 when changes are made to a facility.

2.4.5 Deferring Valve Testing to Cold Shutdown or Refueling Outages

Exercising valves on a cold shutdown or refueling outage frequency does not constitute a

deviation from the Code. Subsection ISTC 3520 provides guidance for testing valves during

cold shutdown or refueling outages if it is impractical to test during operation. The licensee 

should list the affected valves in the program document and include cold shutdown or refueling 

outage justifications for each affected valve or group of valves. The staff recommends that

licensees should include these cold shutdown and refueling outage justifications in their IST

program submittals to the NRC.

Check valves that can be stroked quarterly, but must be monitored by a nonintrusive technique 

to verify full stroke, may be full-stroke tested during cold shutdown or refueling outages if

another method of verifying full-stroke exists during such plant conditions. The NRC would not 

require a licensee to invest in nonintrusive equipment for the purpose of testing check valves 

quarterly (instead of testing them during cold shutdown or refueling outages), even though the 

use of nonintrusive techniques is recommended where practical. However, the NRC would 

continue to require the quarterly partial-stroke testing as applicable. (See also Section 3.1.1)

A licensee may request relief from quarterly testing where such testing would impose a hardship 

(e.g., entering a limiting condition for operation of 3 to 4 hours in duration, repositioning a 

breaker from “off” to “on,” and necessitating for manual operator actions to restore the system if

an accident were to occur while the test was in progress). For such situations, the risk 

associated with quarterly testing may outweigh the benefits that might otherwise be achieved. 

(Section 3.1.2 gives guidance on these types of situations.) Thus, it is appropriate for licensees 

to weigh the safety impact against the benefits of testing as a basis for deferring testing from a

quarterly frequency to cold shutdown or refueling outages. NUREG/CR-5775, “Quantitative 

Evaluation of Surveillance Test Intervals Including Test-Caused Risks,” dated March 1992, 

describes a method for making this comparison.
A 2-13

In the event of a planned or unplanned maintenance outage, a licensee may decide to test 

valves in a cold shutdown mode, rather than waiting for the refueling outage. In making this 

decision, the licensee should consider the duration of the shutdown and the extent of other 

outage activities. The requirements of Subsection ISTC-3560 for testing valves in systems that 

are out of service may apply for extended outages that last for several months. Guidance on 

minimizing shutdown risk may also apply for extended outages.

Impractical conditions justifying test deferrals may include the following situations that could 

result in an unnecessary plant shutdown, cause unnecessary challenges to safety systems, 

place undue stress on components, cause unnecessary cycling of equipment, or unnecessarily 

reduce the life expectancy of the plant systems and components:

• inaccessibility

• testing would require major plant or hardware modifications

• testing has a high potential to cause a reactor trip

• testing could cause system or component damage

• testing could create excessive plant personnel hazards

• existing technology will not give meaningful results

In the licensing process, the NRC staff weighs the possible safety consequences and benefits of 

performing a required test as part of TSs surveillance, including circumstances in which one

train is out of service. Nonetheless, any related guidance provided by the staff does not 

supersede the TSs requirements. For example, if testing is specified as part of the TSs

surveillance, the cycling of non-redundant valves in a remaining operable train may not be 

deferred to the next cold shutdown when one train is out of service, even though their failure 

would cause a loss of total system function. In this case, a TSs change or enforcement 

discretion would be necessary to defer testing.

Licensees are expected to comply with required IST test frequencies. The Code does not 

require documentation for valves that are not tested during a cold shutdown outage other than 

as required for maintaining the IST schedule. The NRC does not have a position on the efforts 

a licensee expends in performing cold shutdown valve testing during a short outage; however, 

the staff expects licensees to expend a reasonable “good faith” effort.

This issue is further discussed in Sections 3 and 4, which give guidance on deferring testing.

2.5 Relief Requests and Proposed Alternatives

A licensee may submit a request for the NRC to review and approve relief from requirements of

the Code, or to authorize the use of proposed alternatives. The staff recommends that the basis 

for relief should address the following considerations:

(1) Does the proposed alternative provide an acceptable level of quality and safety? 

(10 CFR 55a(a)(3)(i))

(2) Would compliance with the specified requirement result in a hardship or unusual 

difficulty without a compensating increase in the level of quality and safety? 

(10 CFR 55a(a)(3)(ii))
A 2-14

(3) Would it be impractical to comply with the Code requirements? (10 CFR 55a(f)(6)(i))

The justification must include adequate information for the staff to determine if the relief can be

granted or the alternative can be authorized (e.g., as applicable, damage to equipment, hazards 

to personnel, and the possibility of a plant trip). NRC approval is required before a licensee may 

implement proposed alternatives that must be authorized pursuant to 10 CFR 50.55a(a)(3). By 

contrast, a licensee may implement proposed alternative testing while the NRC is reviewing 

requests for relief from Code requirements made pursuant to 10 CFR 50.55a(f)(6)(i), if the 

licensee has determined that the requirements are impractical.

Within 10 CFR 50.55a, the regulation uses the term “extent practical” several times with respect

to ISI and IST requirements for nuclear power plant components. For example, 10 CFR

50.55a(f)(4) states that “pumps and valves which are classified as ASME Code Class 1, Class

2, and Class 3 must meet the requirements... set forth in ASME OM Code ...................... to 

the extent practical within the limitations of design, geometry, and materials of construction of 

the components.” “Extent practical” is a common term used in a general context. However, it is 

helpful to clarify how the NRC staff has used the terms “practical” and “impractical” in 

implementing the requirements of 10 CFR 50.55a.

Webster’s Dictionary defines “practical” as “capable of being used.” With respect to ASME 

Code requirements, the NRC staff has interpreted “practical” requirements to mean those that

are actually capable of being implemented within the limitations of design, geometry, and

materials of construction of the component. Conversely, the staff has interpreted “impractical”

requirements to mean those that are incapable of being implemented within the limitations of 

design, geometry, and materials of construction of the component. Where a licensee 

determines that an ASME Code requirement is impractical, the NRC staff may grant relief from 

the requirement, giving due consideration to the burden that might be imposed on the licensee if

the NRC enforced the requirement. Impractical conditions include the following situations:

• inaccessibility

• testing would require major plant or hardware modifications

• testing has a high potential to cause a reactor trip

• testing could cause system or component damage

• testing could create excessive plant personnel hazards

• existing technology will not give meaningful results

Nuclear power plant licensees may also propose alternatives to ASME Code requirements -

even when such requirements would be “practical” to implement - if the proposed alternatives 

would provide an acceptable level of quality and safety or if compliance with the specified 

requirements would result in hardship or unusual difficulty without a compensating increase in

the level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(a)(3)(ii), 

respectively. The NRC staff has interpreted “hardship” to mean a high degree of difficulty or an

adverse impact on plant operation, as illustrated by the following examples:

• raising ALARA concerns

• replacing equipment or in-line components

• creating significant hazards to plant personnel
A 2-15

The staff performs a detailed review of each relief request, grants relief from the requirements or

authorizes an alternative to those requirements, and may impose alternative requirements. 

When granting relief, the NRC considers the burden that would be imposed upon the licensee if

the agency enforced the specified requirements.

2.5.1 Justifications for Relief

In determining whether to grant relief from the Code requirements or to authorize alternatives, 

the NRC staff considers the merits of the submitted technical information. In requesting relief, 

the licensee would typically identify the specific Code requirement and associated paragraph for

which relief is requested, describe the proposed alternative(s), describe the basis for relief or

authorization of the proposed alternative(s), and clarify the burden that would result if the NRC 

enforced the specified requirements. Situations that warrant granting relief or authorizing 

alternatives (as determined by the staff in previous safety evaluations for plant-specific 

requests) may include the following examples:

(1) In complying with the Code requirements, the licensee would not obtain information that

would be more useful than the information that is currently available. For example, 

installing an analog gauge with a range of three times the reference value (or less) to

comply with Code requirements may not yield more accurate readings than those 

provided by the gauge that is presently installed (see Section 5.5.1).

(2) Compliance with the Code is impractical because of design limitations. Imposition of the

Code requirements would require significant system redesign and modifications. For

example, a flow meter does not meet the accuracy requirements of ISTB-3510 and 

Table ISTB 3510-1 because the present system configuration does not have a straight 

section of pipe of sufficient length in which to measure flow accurately (see Section 5.5).

(3) The required measurements or appropriate observations cannot be made because of 

physical constraints. Examples include a component located in an area that is 

inaccessible during power operation or a pump that is totally immersed in system fluid.

(4) The need to keep personnel radiation exposure as low as reasonably achievable 

(ALARA) may present an adequate justification. The licensee should include information 

about the general area radiation field, local hot spots, plant radiation limits and stay 

times, the amount of exposure personnel would receive in doing the testing, and the 

safety significance of deferring testing or performing an alternative method. ALARA 

relates to controlling exposure during an activity, not specifically to eliminating activities; 

however, it may be a basis for relief or for deferring a test on the basis of hardship when

exposure limits are prohibitive for performing testing (or possibly for accessing a valve 

for repair in the event that it could fail during a test). If the exposure limits are 

prohibitive, the licensee should defer testing to cold shutdown or refueling outages 

during which the exposure limits would no longer be prohibitive. ALARA is part of an 

overall program, including activities such as IST, as required by 10 CFR 20.1101. The 

NRC has not established “predetermined acceptable limits” for deferring an IST activity, 

based on maintaining occupational exposure ALARA.
A 2-16

(5) Testing as required by the Code could cause significant equipment damage. For 

example, shutting off cooling flow to an operating pump by exercising a valve in the 

cooling flow path could damage the pump.

(6) Failure of a component during testing could disable multiple trains of a reactor safety 

system. For example, a motor-operated suction valve common to both trains of 

high-pressure safety injection could not be tested during power operation because a 

failure of the valve would result in both trains being out of service.

Inconvenience or administrative burden is not, in and of themselves, adequate justification for

deviating from the Code requirements. Similarly, entering a TS limiting condition for operation 

(LCO) is not, in and of itself, adequate justification for deviating from the Code-specified 

frequency, except when entering the LCO would be prohibited because the total system function 

would be out of service.

2.5.2 Categories of Relief Requests

The NRC staff categorizes relief requests as follows:

• General: A general relief request is appropriate when the requested relief applies to a 

broad range of similar components in the program, such as all pumps or all containment 

isolation valves.

• Specific: A relief request is specific when the requested relief applies only to a single 

component or a specified group of similar components in the program, such as service 

water pump discharge check valves.

2.5.3 Content and Format of Relief Requests

As a minimum, the staff recommends that each relief request should include the following 

information:

• Title and relief request number: Title each relief request and specify a unique identifier. 

The identifier should remain unique to avoid confusion when later revisions are made. 

Examples include (1) “Relief Request Number 1,” (2) “Safety Injection Pumps Relief 

Request,” or (3) “Check Valves in Series Relief Request.”

• Page number: List the page number and total number of pages in the program 

document or the relevant section, such as “Page 15 of 135.”

• Program revision or page revision date: List the program or page revision number and 

date (on each page). List the revision number for each program change submitted.

• System and Code class: List the plant system and Code class of the system in which 

the component is located.
A 2-17

• Pump/valve category or group: List the ASME category or group for each pump or valve 

(i.e., A, A/C, B, C, or D).

• Component identification: List the identification number for each component in a specific 

relief request. Each individual component need not be listed in a general relief request, 

such as one for all pumps in the IST program. However, the staff recommends that the 

list of program components (pump or valve table) should include the relief request 

number.

• Component function: Briefly describe the functions of the affected components and 

specify the function that is the subject of the relief request.

• ASME Code test requirement(s): List and describe the Code requirement(s) from which 

relief is being requested.

• Basis for relief: Clearly state the legal basis under which relief is requested, and then 

explain the reasons why complying with the Code requirements is impractical, poses a 

hardship, or otherwise should not be enforced. Include justification for each test 

frequency deferral (e.g., quarterly to cold shutdown), and state and justify the proposed 

frequency. Include all information that the NRC staff might need to complete its review. 

For example, most relief requests for check valves list the test direction(s) for which 

relief is requested.

• Proposed alternative testing: Clearly and thoroughly discuss the proposed alternative in 

sufficient detail to clearly demonstrate why it is a reasonable alternative to the Code 

requirement, and provide a technical basis for its acceptability.

• Drawings and/or diagrams: If the relief request or alternative testing is complex, or if 

drawings or diagrams are available for further clarification, include them in the relief 

request, or include them in the IST program document and reference them in the relief 

request.

• References: List references to SAR sections, technical specifications, and other 

pertinent documents. Any document referenced in the relief request should be 

submitted to the NRC on the plant docket. If a document is not docketed but contains 

pertinent information, the relief request should explicitly include the information (if it is 

not readily available to the staff and the public), rather than merely referencing the 

document.

In June 2004, the NEI issued a White Paper, entitled “Standard Format for Requests from 

Commercial Reactor Licensees Pursuant to 10 CFR 50.55a, Revision 1.” This White Paper, 

provides useful guidance for determining the appropriate regulatory requirement under which a 

“relief request” is submitted for NRC approval, as well as the appropriate format and content to 

use in the request. The term “relief request” is used loosely in this instance to denote the 

various types of submittals allowed by 10 CFR 50.55a, including alternatives to the regulation 

[10 CFR 50.55a(a)(3)], impractical relief requests [10 CFR 50.55a(f)(6)(i)], and requests to use 

later Code Editions and Addenda [10 CFR 50.55a(f)(4)(iv)]. The NRC staff has reviewed the 

NEI White Paper and encourages licensees to use the specified format and content.
A 2-18

2.6 Program Documents

The regulations at 10 CFR 50.55a(f)(4) requires that throughout the service life of a boiling or 

pressurized water-cooled nuclear power plant facility, pumps and valves which are classified as 

ASME Code Class 1, 2, and 3 must meet the inservice test requirements of the ASME OM 

Code as incorporated by reference in 10 CFR 50.55a(b). The applicable ASME OM Code, 

Subsection ISTA, General Requirements provides documentation requirements such as the 

following:

• ISTA-3200(a) requires that IST of pumps and valves plans shall be filed with the 

regulatory authorities having jurisdiction at the plant site.

• ISTA-9000, Records and Reports, provides the requirements for preparation, submittal, 

and retention of records and reports.

• Nonmandatory Appendix-A and Supplement to Nonmandatory Appendix A describes 

voluntary guidance for licensees for development of IST of pumps and valves plans. 

Licensees not meeting ISTA-3200(a) must submit appropriate documents containing pumps and 

valves IST plans and submit a request for relief to the NRC pursuant to 10 CFR 50.55a(a)(3). 

NRC staff will not perform a review of submitted pumps and valves IST programs unless 

requesting alternatives or reliefs to Code requirements. 

IST program documents submitted to the NRC are used to prepare for IST inspections and to

address other licensing actions that may arise. Between a licensee’s 10-year interval program 

submittals, the NRC would like to receive up-to-date program documents when the licensee 

makes significant changes to the IST program to facilitate these regulatory activities.

As long as the IST program is consistent with the regulations, ASME Code relief is not required. 

That is, deletions from or additions to the IST program do not necessarily require NRC approval. 

The burden is on each licensee to verify that its IST program is complete and includes all

components that require IST, and that all such components are tested to the extent practical. If

a licensee deletes a particular component from its IST program, the staff recommends that the

licensee should document the reason in an appropriate place.

The staff expects each licensee to maintain its IST program up-to-date and ensure that it

remains consistent with changes in plant configuration. If a particular relief request is no longer 

required because of changes in hardware, system design, or new technology, the licensee is 

expected to revise its program to withdraw the relief request. Conversely, if a system 

modification results in the addition of a component to the IST program, the licensee should 

ensure that it meets the Code requirements, or that a relief request is submitted for NRC review 

and approval, as appropriate.
A 2-19

2.7 Developing IST Programs for New Nuclear Power Plants

Under 10 CFR Part 52, the development of a plant-specific IST program is the responsibility of 

the applicant for a combined license (COL) to construct and operate a nuclear power plant. The 

Commission’s Staff Requirements Memorandum (SRM), dated September 11, 2002, for 

Commission Paper SECY-02-0067, “Inspections, Tests, Analyses, and Acceptance Criteria

(ITAAC) for Operational Programs (Programmatic ITAAC),” stated that ITAAC for an operational 

program are unnecessary if the program and its implementation are fully described in the COL 

application and found to be acceptable by the NRC. The Commission also stated that the 

burden is on the COL applicant to provide the necessary and sufficient programmatic 

information for approval of the COL without ITAAC. 

In its May 14, 2004, SRM for SECY-04-0032, “Programmatic Information Needed for Approval 

of a Combined License Without Inspections, Tests, Analyses and Acceptance Criteria,” the 

Commission defined “fully described” as meaning that the program is clearly and sufficiently 

described in terms of the scope and level of detail to allow a reasonable assurance finding of 

acceptability. The Commission also noted that required programs should always be described 

at a functional level and at an increasing level of detail where implementation choices could 

materially and negatively affect the program effectiveness and acceptability. SECY-05-0197, 

“Review of Operational Programs in a Combined License Application and Generic Emergency 

Planning Inspections, Tests, Analyses, and Acceptance Criteria,” summarizes the NRC position 

regarding the full description of operational programs to be provided by COL applicants. RG 

1.206, “Combined License Applications for Nuclear Power Plants (LWR Edition),” provides 

guidance for COL applicants with respect to fully describing plant operational programs. 

The NRC staff is reviewing the descriptions of IST programs provided by COL applicants as part 

of their COL applications. The staff will reach a conclusion regarding the acceptability of the 

description of the IST program in its safety evaluation on the COL application. The NRC staff 

will conduct inspections of the development and implementation of IST programs during the 

construction of new nuclear power plants following COL issuance. The guidance in this NUREG 

may be used in developing and implementing the IST program for new nuclear power plants. 

For a COL issued per 10 CFR Part 52, the NRC regulations in 10 CFR 50.55a(f)(4)(i) and 

(g)(4)(i) state that inservice tests and inspections conducted during the initial 120-month interval 

to verify operational readiness of applicable plant components, whose function is required for 

safety, must comply with the requirements in the latest edition and addenda of the ASME Code, 

incorporated by reference in 10 CFR 50.55a(b) on the date 12 months before the date 

scheduled for initial fuel loading (or the optional ASME Code cases listed in RG 1.192) subject 

to the limitations and modifications listed in 10 CFR 50.55a. Licensees may submit a request 

under 10 CFR 50.55a to apply the edition and addenda of the ASME OM Code specified in the 

COL application for the initial 120-month IST interval as an alternative to the latest edition and 

addenda of the ASME OM Code, where the differences between the IST provisions in these 

Code editions and addenda are addressed.
A 2-20

Table 2.1 Typical Systems and Components in an Inservice Testing Program

for a Pressurized-Water Reactor

Typical safety-related, Code-class 

system in pressurized-water reactors

Typical components in an inservice testing program

Reactor coolant system and 

flowpaths for establishing natural 

circulation

Power-operated relief valves and associated block valves

Reactor high point and head vents

Primary system safety and relief valves (pressurizer Code 

safety valves)

Valves in any proposed flowpath used for long-term core 

cooling or safe shutdown

Pressure boundary isolation valves

Valves in lines to pressurizer relief/quench tank

Main steam system Main steam isolation valves (MSIVs)

Main steam non-return valves (if applicable)

Secondary system safety and relief valves

Atmospheric dump valves

Auxiliary feedwater turbine steam supply valves

Steam generator blowdown isolation valves

High-pressure safety injection system High-pressure injection pumps and discharge check valves

Injection valves in injection flowpath

Isolation valves

Valves for the refueling water storage tank (RWST) borated 

water storage tank (BWST), and refueling water tank 

(RWT), including vacuum breakers

Chemical and volume control or 

makeup system

Charging or makeup pumps and suction/discharge check 

valves

Valves in charging/makeup flowpath

Boric acid transfer pumps and suction/discharge check 

valves

Valves in emergency boration flowpaths

Relief valves
A 2-21

Table 2.1 Typical Systems and Components in an Inservice Testing Program

for a Pressurized-Water Reactor (continued)

Typical safety-related, Code-class 

system in pressurized-water reactors

Typical components in an inservice testing program

Low-pressure safety injection system Injection pumps and suction/discharge check valves

Valves associated with safety injection accumulators and 

core flood tanks

Recirculation flowpath valves, including containment sump 

isolation valves

Isolation valves (high-low pressure interface)

Relief valves

Shutdown cooling, residual heat 

removal, or decay heat removal 

systems

Pumps and suction/discharge check valves

Valves in flowpath

Isolation valves (high-low pressure interface)

Relief valves

Containment spray system Containment spray pumps and suction/discharge check 

valves

Valves in flowpaths to spray header

Isolation valves

Valves in spray additive flowpath

Spray additive tank valves, including vacuum breakers

Main feedwater system Main feedwater isolation valves

Auxiliary feedwater system Auxiliary feedwater pumps and suction/discharge check 

valves

Valves in flowpath to steam generators

Valves in suction lines

Valves between normal and ultimate heat sink suction 

sources

Relief valves and isolation valves
A 2-22

Table 2.1 Typical Systems and Components in an Inservice Testing Program

for a Pressurized-Water Reactor (continued)

Typical safety-related, Code-class 

system in pressurized-water reactors

Typical components in an inservice testing program

Primary containment system Containment isolation valves (various systems)

Containment combustible gas venting valves

Containment atmosphere sampling valves (if within the 

scope of 10 CFR 50.55a)

Component cooling water system Component cooling water pumps and discharge check 

valves

Valves in letdown cooling water flowpath

Valves in reactor coolant pump seal injection and cooling 

water flowpath

Valves needed to isolate a rupture of the thermal barrier

Relief valves

Spent fuel pool/pit cooling system Spent fuel cooling pumps and suction/discharge check 

valves

Valves in flowpath from ultimate heat sink source supply

Service water system Service water pumps and suction/discharge check valves

Valves in flowpath to auxiliary feedwater system

Valves in flowpaths to emergency room coolers

Valves in flowpaths to containment emergency coolers

Valves in flowpaths to emergency diesel generator heat 

exchangers

Isolation and cross-tie valves

Valves in ultimate heat sink source flowpaths

Valves in standby or backup service water, if applicable
A 2-23

Table 2.1 Typical Systems and Components in an Inservice Testing Program

for a Pressurized-Water Reactor (continued)

Typical safety-related, Code-class 

system in pressurized-water reactors

Typical components in an inservice testing program

Emergency diesel generator system 

(within scope of 10 CFR 50.55a)

Fuel oil storage and transfer pumps and valves

Diesel generator external cooling (service water)

Engine air start check valves

Air receiver relief valves

Ventilation systems Pumps and valves in control room emergency cooling 

water supply flowpath

Instrument air system (if within the 

scope of 10 CFR 50.55a)

Air supply to containment purge valves

Air supply to power-operated relief valves (PORVs)

Air supply to MSIVs
A 2-24

Table 2.2 Typical Systems and Components in an Inservice Testing Program

for a Boiling-Water Reactor

Typical safety-related, Code-class 

system in boiling-water reactors

Typical components in an inservice testing program

Nuclear boiler and reactor 

recirculation system

Primary system isolation valves

Excess flow check valves

Main steam system MSIVs and actuator valves (pilot valves, accumulator check 

valves)

Main steam safety and relief valves

Main steam safety valve discharge rupture diaphragm valve

MSIV leakage valves

High-pressure core coolant injection 

(HPCI) system

Pump and suction/discharge check valve

Valves in injection flowpath

Isolation valves, including valves in test lines

Excess flow check valves

HPCI pump turbine valves, including turbine exhaust 

vacuum breakers (unless considered skid-mounted)

High-pressure core spray system Pumps and suction/discharge check valves

Valves in injection flowpath

Isolation valves, including valves in test lines 

Reactor core isolation cooling (RCIC) 

system (if safety-related)

Pump and suction/discharge check valve

RCIC pump turbine valves

Excess flow check valves

Isolation valves

Reactor water cleanup system Containment isolation valves
A 2-25

Table 2.2 Typical Systems and Components in an Inservice Testing Program

for a Boiling-Water Reactor (continued)

Typical safety-related, Code-class 

system in boiling-water reactors

Typical components in an inservice testing program

Residual heat removal (RHR) system RHR pumps and suction/discharge check valves

Isolation and cross-tie valves

Pump suction relief valves

RHR heat exchanger thermal relief valves

Valves in injection flowpath

Flow control valves 

Spent fuel pool cooling system Fuel pool pumps and suction/discharge check valves

Ultimate heat sink supply valve

Feedwater coolant injection and isolation 

condenser system (if applicable)

Reactor feedwater pumps and suction/discharge check 

valves

Condensate pumps and suction/discharge check valves

Condensate booster pumps and suction/discharge 

check valves

Emergency condensate transfer pump and 

suction/discharge check and isolation valves

Isolation and bypass valves

Vent valves

Makeup to condenser shell check valves

Standby liquid control (SBLC) system SBLC pumps and suction/discharge check valves

Relief valves

Injection line valves

Explosively-actuated squib valves

Main feedwater system Isolation valves
A 2-26

Table 2.2 Typical Systems and Components in an Inservice Testing Program

for a Boiling-Water Reactor (continued)

Typical safety-related, Code-class 

system in boiling-water reactors

Typical components in an inservice testing program

Primary containment system Containment isolation valves including excess flow 

check valves (various systems)

Containment atmosphere monitoring system valves

Containment atmosphere dilution system valves

Containment pressure suppression and vents

Closed cooling or component cooling 

water system

Pumps and suction/discharge check valves

Valves in flowpaths to safety-related coolers

Service water system Pumps and suction/discharge check valves

Isolation and cross-tie valves

Valves in flowpaths to safety-related coolers

Valves in flowpaths to diesel generator coolers

Valves in standby or backup service water

Valves in flowpath from ultimate heat sink source

Valves in residual heat removal service water flowpath

Control rod drive system (portions within 

the scope of 10 CFR 50.55a)

Scram dump valves

Scram discharge volume vent valves

Scram discharge volume drain valves

Accumulator rupture disks

Hydraulic control unit control valves

Drive water backflow prevention valves

Emergency diesel generator systems (if 

within the scope of 10 CFR 50.55a)

Fuel oil storage and transfer pumps and valves

Diesel generator external cooling (service water)

Engine air start check valves

Air receiver relief valves
A 2-27

Table 2.2 Typical Systems and Components in an Inservice Testing Program

for a Boiling-Water Reactor (continued)

Typical safety-related, Code-class 

system in boiling-water reactors

Typical components in an inservice testing program

Ventilation systems Pumps and valves in control room emergency cooling 

water supply flowpath

Instrument air system (if within the scope 

of 10 CFR 50.55a)

MSIV accumulator check valves

MSIV pilot valves

Automatic depressurization system (ADS) valve 

accumulator check valves

ADS pilot valves

Traversing incore probe system (if within 

the scope of 10 CFR 50.55a)

Containment isolation valves
A 2-28

Table 2.3 Example Data Table for Pumps.;

PLANT NAME/UNIT

PUMP TESTING PLAN

Revision 3

Date: 1-15-03

Page: 1 of 3

Pump List Parameters

SYSTEM PUMP ID P&ID NO. COORD.

PUMP

GROUP S P dp Flow

V

(PR-1)

Residual Heat

Removal

RHR-01

RHR-02

RHR-03

M-402, Sh. 1

M-402, Sh. 2

M-402, Sh. 2

D-4

G-4

F-5

A (1)

(1)

(1)

(2)

(2)

(2)

Q/2Y

Q/2Y

Q/2Y

Q/2Y

Q/2Y

Q/2Y

Q/2Y

Q/2Y

Q/2Y

Auxiliary

Feedwater

AFW-01

AFW-02

AFW-03

M-408, Sh 1

M-408, Sh. 1

M-408, Sh. 1

B-5

B-8

B-11

B (1)

(1)

Q/2Y

(2)

(2)

(2)

Q/2Y

Q/2Y

Q/2Y

Q/2Y

Q/2Y

Q/2Y

2Y

2Y

2Y

Service Water SWS-01

SWS-02

SWS-03

SWs-04

M-335, Sh 1

M-335, Sh 2

M-335, Sh. 3

M-335, Sh. 4

F-9

D-4

E-8

C-4

A (1)

(1)

(1)

(1)

(2)

(2)

(2)

(20

Q/2Y

Q/2Y

Q/2Y

Q/2Y

PR-3

PR-3

PR-3

PR-3

PR-2

PR-2

PR-2

PR-2

Standby

Liquid Control

SLC-01

SLC-02

M-367, Sh. 1

M-367, Sh. 1

D-9

D-4

B (1)

(1)

2Y

2Y

(3)

(3)

Q/2Y

Q/2Y

2Y

2Y

Note (1): Pump is directly coupled to a content speed synchronous or induction type driver

Note (2): Discharge pressure is a required parameter for positive displacement pumps only.

Note (3)” dP is not a required parameter for positive displacement pumps.

Legend:

S

Pi

dP

PR

Q

V

Speed

Pressure

Differential Pressure

Pump Relief Request

Quarterly

Vibration
A 2-29

Table 2.4 Useful Abbreviations for Valve Data

Parameter Abbreviation Description

Valve Type GT

GB

CK

RV

SC

BF

DI

EX

BA

Gate valve

Globe valve

Check valve

Relief valve

Stop check

Butterfly valve

Diaphragm valve

Explosive valve

Ball valve

Actuator Type MO

SO

AO

HO

SA

MA

PA

Motor-operated

Solenoid-operated

Air-operated

Hydraulic-operated

Self-actuated

Manual

Pilot-actuated

Safety Poistion(s) O

C

O/C

T

Open

Closed

Both open and closed

Throttled

Tests(s) Performed FS

PS

LT

LJ

ST

FT

PI

RV

EX

Full-stroke exercise valve to safety position(s)

Part-stroke exercise valve

Leak-rate test valve to Section XI requirements

Leak-rate test valve to Appendix J requirements

Measure the full-stroke times of the valve

Observe the fail-safe operation of the valve

Verify the valve position indication

Safety and relief valve test

Explosive valve test 

Test Frequency Q

CS

RF

2Y

RV

SD

Test performed once every 92 days

Test performed during cold shutdowns, but not more 

frequently than once every 92 days

Test performed each reactor refueling outage

Test performed once every 2 years

Test relief valve at OM schedule

Disassemble, inspect, and manually exercise one valve 

from specified group each reactor refueling outage
*Note: Flow chart provided for guidance only. For complete details see 10 CFR 50.55a

A 2-30

Figure 2.1- Appendix- A

FLOW CHART – Development of Inservice Testing Program for Pumps & Valves*

Develop new Inservice Testing (IST) Program 

for Pumps and Valves

[10 CFR 50.55a(f)(4)(i)]

or

Mandatory update of Pumps and Valves IST 

Program Every 10-Year Interval 

[10 CFR 50.55a(f)(4)(ii)]

Submit Relief Request (RR) to use relief or 

Alternative Requirement in lieu of the 

ASME OM Code requirements for a 10-

year Interval 

[NEI Guidance for Standard Format of RR 

dated June 7, 2004]

Licensee may ask for approval to use later 

Edition and Addenda of the Code per 

10 CFR 50.55a(f)(4)(iv) and 

 

Subsection ISTA, General Requirements

Subsection ISTB, Pump IST Requirements

Subsection ISTC, Valves IST Requirements

IST Program must include conditions in 10 CFR 50.55a(b)(3)

1. Motor-Operated Valves (MOVs): The condition in 

10 CFR 50.55a(b)(3) (ii) supplements the MOVs stroketime testing requirements in ISTC.

2. Check Valves: The Condition in 10 CFR 50.55a(b)(3)(iv) 

(A), (B), (C), and (D) supplements the provision in 

Appendix II, “Check Valve Condition Monitoring 

Program, “

3. Manual Valves: The condition in 10 CFR 50.55a(b)(3)(vi) 

is only applicable for ISTC-3540 of ASME OM Code up to 

 

ISTA-3200, Submittal of the Pump and 

Valve IST Program to Regulatory 

Authority (NRC) 

ISTA-9000, Records and Reports

ISTB-9000, Records and Reports

ISTC-9000, Records and Reports

Note: NRC does not review and 

approve submitted IST program.

NRC authorized alternative in lieu of the ASME 

Code requirements for a 10-Year Interval

10 CFR50.55a(a)(3)(i), or

10 CFR 50.55a(a)(3)(ii), 

NRC granted relief due to impracticability for a 

10-Year Interval [10 CFR 50.55a(f)(6)(i)]

Not meeting 

the OM Code 

requirements

May use Code Cases, 

listed in Regulatory Guide 

(RG) 1.192 as conditioned 

[10 CFR 50.55a(b)(6)]

Yes, meeting 

the OM Code 

requirements

Meet the IST 

requirements in ASME 

OM Code with Edition 

and Addenda as specified 

in 10 CFR 50.55a(b)

IST Program with NRC approvals, 

authorizations, reliefs granted
A 3-1

3. GENERAL GUIDANCE ON INSERVICE TESTING

3.1 Inservice Test Frequencies and Extensions for Valve Testing

The ASME OM Code generally specifies quarterly testing of valves. Subsection ISTC of the 

Code allows licensees to defer valve exercising to cold shutdown or refueling outages if it is not 

practical to exercise the valves during plant operation. The NRC staff may approve relief to 

extend a test interval for extenuating circumstances in which (1) compliance would result in 

hardship or unusual difficulty without a compensating increase in the level of quality and safety, 

or (2) the system design makes compliance impractical. Impractical conditions justifying test 

deferrals are those that could result in an unnecessary plant shutdown, cause unnecessary 

challenges to safety systems, place undue stress on components, cause unnecessary cycling of 

equipment, or unnecessarily reduce the life expectancy of the plant systems and components. 

Any relief request would typically include a technical justification for the deferment. Table 3.1

(below) lists the tests and associated test frequencies required by the Code.

Table 3.1 ASME OM Code Tests and Test Frequencies for Pumps and Valves.

Test Frequency

Measure pump parameters Once every 3 months (Group A, B)

Biennially (Comprehensive Test)

Exceptions:

∙ Pumps in systems that are out-of-service

∙ Group B Pumps lacking required fluid inventory

Exercise Category A and B valves Once every 3 months

Exceptions:

∙ Extension because of impracticality

∙ Valves in regular use

∙ Valves in systems out of service

Measure stroke times of power-operated

Category A and B valves

Once every 3 months

Exceptions:

∙ Extension because of impracticality

∙ Valves in regular use

∙ Valves in systems out of service

Verify remote position indication Once every 2 years

Observe operation of fail-safe actuators for 

applicable valves

Once every 3 months, except for extension 

because of impracticality
A 3-2

Table 3.1 ASME OM Code Tests and Test Frequencies for Pumps and Valves (continued).

Test Frequency

Leak-test Category A and A/C valves Once every 2 years

Test safety and relief valves, primary containment 

vacuum relief valves, and non-reclosing pressure 

relief devices

Test interval specified by OM Appendix I

Exercise check valves Once every 3 months

Exceptions:

∙ Extension because of impracticality

∙ Valves in regular use

∙ Valves in systems out of service

Test explosively actuated valves 20 percent tested once every 2 years.

Charges shall not be older than 10 years.

3.1.1 Deferring Valve Testing to Each Cold Shutdown or Refueling Outage

The OM Code allows licensees to test valves during cold shutdowns if it is impractical to test the

valves quarterly during plant operation. Subsection ISTC-3500 provides guidance and

alternatives. Therefore, exercising valves during cold shutdown outages does not constitute a 

deviation from the Code and does not require a relief request if the licensee determines that 

quarterly testing is impractical. Similarly, the OM Code allows licensees to test valves during 

each refueling outage if it is impractical to test the valves during cold shutdowns. In such 

instances, the licensee must identify the valves for which testing is deferred and the IST 

program document must specify the basis for determining that quarterly and/or cold shutdown 

testing is impractical.

In 1976, the NRC staff issued guidance in the form of letters to licensees, which included 

examples of valves to be specifically excluded from exercising (cycling) tests during plant 

operations. This guidance may not apply in all cases; for example, high-pressure coolant 

injection (HPCI) turbine steam supply and pump discharge valves may be tested quarterly 

during pump testing. The excluded valves include the following examples:

(1) All valves that would cause a loss of system function if they were to fail in a 

nonconservative position during the cycling test. Valves in this category would typically 

include all non-redundant valves in lines such as a single discharge line from the 

refueling water storage tank (RWST) or accumulator discharge lines in PWRs and the 

HPCI turbine steam supply and HPCI pump discharge in BWRs. Other valves may fall 

into this category under certain system configurations or plant operating modes. For 

example, when one train of a redundant system (such as an emergency core cooling 

system (ECCS)) is inoperable, non-redundant valves in the remaining train should not be 

cycled because their failure would cause a loss of total system function.
A 3-3

(2) All valves that would result in a loss of containment integrity if they failed to close during

a cycling test. Valves in this category would typically include all valves in containment 

penetrations where the redundant valve is open and inoperable.

(3) All valves that, when cycled, could subject a system to pressures in excess of their 

design pressures. For the purpose of a cycling test, it is assumed that one or more of

the upstream check valves has failed unless positive methods are available for

determining the pressure or lack thereof on the high-pressure side of the valve to be

cycled. Valves in this category would typically include the isolation valves of the residual 

heat removal/shutdown cooling system and, in some cases, certain ECCS valves.

The guidance herein and in the NRC’s letter issued in 1976 to licensees do not supersede the 

TSs requirements.

Check valves that can be stroked quarterly, but must be monitored by a nonintrusive technique 

to verify full stroke, may be full-stroke tested during cold shutdown or refueling outages if

another method of verifying full-stroke exists during such plant conditions. The NRC would not 

require a licensee to invest in nonintrusive equipment for the purpose of testing check valves 

quarterly (instead of testing them during cold shutdown or refueling outages), even though the 

use of nonintrusive techniques is recommended where practical. However, the NRC would 

continue to require the quarterly partial-stroke testing as applicable. (See also Section 2.4.5)

A licensee may request relief from quarterly testing where such testing would impose a hardship 

(e.g., entering a limiting condition for operation of 3 to 4 hours in duration, repositioning a 

breaker from “off” to “on,” and necessitating for manual operator actions to restore the system if

an accident were to occur while the test was in progress). For such situations, the risk 

associated with quarterly testing may outweigh the benefits that might otherwise be achieved. 

(Section 3.1.2 gives guidance on these types of situations.) Thus, it is appropriate for licensees 

to weigh the safety impact against the benefits of testing as a basis for deferring testing from a

quarterly frequency to cold shutdown or refueling outages. NUREG/CR-5775, “Quantitative 

Evaluation of Surveillance Test Intervals Including Test-Caused Risks,” dated March 1992, 

describes a method for making this comparison.

The following sections discuss issues related to deferring valve testing. These sections do not

apply to testing that is required following maintenance or repair activities.

3.1.1.1 IST Cold Shutdown Testing

Although Subsection ISTC of the ASME OM Code does not include schedules for cold 

shutdown testing, an acceptable method is to ensure that the valves tested in the preceding 

cold shutdown are the last valves tested during the next cold shutdown, with the exception of 

valves that must be tested during each cold shutdown. The following is a sample schedule for

15 cold shutdown tests:
A 3-4

• First cold shutdown: Complete Tests 1, 2, 3, 4, 5, and 6.

• Second cold shutdown: Complete Tests 7, 8, 9, and 10.

• Third cold shutdown: Complete Tests 11, 12, 13, 14, 15, 1, 2, and 3.

• Fourth cold shutdown: Complete Tests 4, 5, 6, and 7.

Subsection ISTC-3520 discusses exercising valves during both plant operation and cold

shutdown as circumstances and situations apply. While the discussion does not specifically 

address testing in hot standby or hot shutdown, valves should be exercised in the appropriate 

mode of operation. For a valve that cannot be tested in operation, testing might be practical 

during hot standby, hot shutdown, cold shutdown, or a refueling outage.

Valves that must be operable during cold shutdown may be tested during plant operation in

accordance with Subsection ISTC-3520, “Exercising Requirements,” or Subsection ISTC-3550, 

“Valves in Regular Use.” By contrast, Subsection ISTC-3550 applies if the component’s “normal 

use” is during cold shutdown.

3.1.1.2 Testing at a Refueling Outage Frequency for Valves Tested During Power 

Ascension

Subsection ISTC-3520 specifies that valves that are tested on a refueling outage frequency 

should be tested before returning the plant to operation at power. Several licensees have 

indicated that they cannot test certain valves until power ascension begins. The NRC staff has 

included this section to provide guidance for such valves and to indicate that the operability TSs 

would control the timing for testing such valves. It is intended that the IST program document 

will identify such valves as being tested on a refueling outage frequency, even though the plant 

may actually return to “operation” at power before the testing is completed. A similar intent 

applies to valves that are tested during power ascension from cold shutdowns (which are not 

refueling outages); however, Subsection ISTC uses different language in discussing valves that 

are tested on a cold shutdown frequency.

Before beginning power ascension from a refueling outage, licensees normally complete the

tests of those valves that are tested during each refueling outage. However, for valves that can 

only be tested during power ascension or at power, licensees may begin increasing the power 

level and changing modes in accordance with TSs requirements and may test the applicable 

valves when plant conditions allow testing. This situation could also apply to valves that are 

tested during power ascension or at power following a cold shutdown outage.

NRC Recommendation

The ASME OM Code Subsection ISTC-3520 requires that all valves testing scheduled for 

performance during a refueling outage shall be completed before returning the plant to 

operation. For valves which can only be tested during power ascension, TSs requirements (for 

the valves or the associated system) determine when the valves are required to be operable. 

The testing for these valves may be scheduled for refueling outages or during cold shutdown 

conditions, but completed during power ascension from the refueling outages. 
A 3-5

The NRC has determined that testing of such valves during plant startup period but prior to 

reaching steady state full power operation is acceptable without a relief, provided that the 

licensee meets all requirements of these provisions.

Basis for Recommendation

The staff has determined that the guidance in this section is consistent with ASME OM Code 

Subsection ISTC-3520 (ISTC-3521 and ISTC-3522) and the TS requirements and, therefore, is

acceptable for meeting these provisions.

3.1.1.3 De-Inerting Containment of Boiling-Water Reactors To Allow Cold Shutdown 

Testing

According to 10 CFR 50.44, “Standards for Combustible Gas Control System in 

Light-Water-Cooled Power Reactors,” each BWR that is equipped with a Mark I or Mark II 

containment must have provisions for an inerted containment atmosphere during power 

operation to protect against a burn or explosion of hydrogen gas generated by the core metalwater reaction following a postulated loss-of-coolant accident (LOCA).

Licensees regularly monitor oxygen content in the containment atmosphere during normal 

power operation, and the plant’s TSs specify the maximum oxygen concentrations allowed. 

Since hydrogen generation is not a concern during cold shutdown or refueling outages, the TSs

allow the containment atmosphere to be de-inerted. However, licensees do not routinely 

de-inert the containment during cold shutdown outages because of impracticality concerns 

associated with the time needed to de-inert and re-inert the containment, and because of the 

amount of nitrogen necessary for inerting.

For certain valves that are located in the inerted containment, Subsection ISTC 3500 [4.2] 

allows licensees to perform testing during cold shutdown outages because it is not practical to

test such valves during power operation. The staff has determined that it is impractical to

de-inert the containment during each cold shutdown outage solely to perform such routine 

testing or repair activities.

NRC Recommendation

The staff considers it impractical to de-inert the containment merely to conduct regularly 

scheduled valve testing, and the ASME Code allows licensees to defer such testing to a

refueling outage when the containment is de-inerted for refueling or other reasons. The staff 

has also determined that few outages require de-inerting, and it is unnecessary to maintain a 

separate schedule for valve testing. Consequently, testing is at the discretion of the licensee in 

the event of an extended cold shutdown that necessitates de-inerting the containment. Factors 

to be considered in the licensee’s decision-making might include the length of the shutdown and 

the extent of other outage activities. In addition, for extended outages that last several months, 

the requirements of Subsection ISTC-3570 may apply for valves in systems that are 

out-of-service.
A 3-6

Basis for Recommendation

Subsection ISTC of the ASME OM Code allows licensees to extend the test interval to defer 

valve testing to refueling outages if such testing is impractical at quarterly intervals (during 

power operations) or during cold shutdown outages. Consequently, it is also acceptable for 

licensees to extend the test interval for valves that cannot be tested unless the containment is 

de-inerted.

Unless the licensee has some other reason to enter the drywell during cold shutdown outages, it 

is impractical to de-inert the drywell during such outages merely to perform valve testing 

because of the time and effort needed to de-inert, re-inert, and replace lost nitrogen gas (which

could delay the plant’s return to power operation). 

3.1.1.4 Stopping Reactor Coolant Pumps for Cold Shutdown Valve Testing

Licensees frequently defer the testing of certain valves in support systems that perform 

functions that are vital to the continued operability of the reactor coolant pumps, such as 

component cooling and the supply and return of seal water during cold shutdown. Exercising 

these valves while the reactor coolant pumps are operating could result in pump damage, and 

stopping the pumps could extend the cold shutdown period.

NRC Recommendation

The staff recommends that licensees should test the affected valves on a refueling outage 

schedule and during plant outages when the reactor coolant pumps are stopped for a sufficient 

period of time, but not more often than once every 92 days.

Basis for Recommendation

Subsection ISTC of the ASME OM Code allows licensees to extend the test interval to defer 

testing to refueling outages when it is not practical to perform the tests during power operation 

or cold shutdown outages. The NRC staff has determined that licensees need not schedule 

valve testing that requires stopping and restarting reactor coolant pumps during each cold

shutdown solely to allow for the testing of such valves. This repetitive cycling would increase 

pump wear and stress, as well as the number of cycles of related plant equipment, and could 

extend the length of cold shutdown outages. Consequently, licensees may consider

establishing a schedule to account for extended cold shutdown outages that would allow for

valve testing when the reactor coolant pumps are stopped for a sufficient period of time. 

However, valves are to be tested at least during each refueling outage.

3.1.2 Entry into a Limiting Condition for Operation (LCO) To Perform Testing

Unless accompanied by some other acceptable rationale, the necessity to enter into an LCO 

ACTIONS to perform IST would not be sufficient justification, because most of the cases ASME 

OM Code Subsections ISTC and ISTB allow deferring testing until a cold shutdown or refueling 

outage. See Section 3.1.3 for guidance on scheduling of inservice testing.
A 3-7

3.1.3 Scheduling of Inservice Tests

TSs use the intervals that the ASME OM Code specifies for IST activities. Such as testing a 

specific pump every biannually, TS would require testing every 184 days. 

NRC Recommendation

To eliminate ambiguity concerning the periods stated in the Code, the staff recommends that

licensees should use the stated test frequency definitions (as shown in Table 3.2, below). For 

example, Subsection ISTC-3510 requires licensees to test Category A and B valves “nominally 

every 3 months.” For quarterly testing, the staff recommends that licensees should schedule 

the pump and valve tests such that a particular test is performed at approximately the same 

time within each quarter. For example, if a test procedure applies to many valves and, thus, 

requires 2 to 3 weeks or more to complete, the licensee would typically begin the procedure at 

approximately the same time in each quarter and include directions to perform tests in a 

specified order to ensure that specific valves are tested “at least once every 92 days.

Table 3.2 ASME OM Code Terms for Inservice Testing Activities.

Term Required frequency for IST activities (at least once every)

Monthly 31 days

Quarterly

(or Every 3 months)

92 days

Yearly (or Annually) 366 days

Refueling refueling outage

2 years 24 months

Licensees must perform each applicable test within the specified time interval, with a maximum 

allowable extension not to exceed 25 percent of the test interval. However, licensees should 

not extend the test intervals for safety and relief valves defined in Appendix I to the OM Code, 

other than to coincide with a refueling outage. Licensees may also apply a 25-percent 

extension to the 2-year interval, unless the TS or relief request stipulates “whichever is more

conservative,” or another statement to that effect.

The Code specifies performing the tests throughout extended shutdown periods for equipment 

that must be returned to operable status. Most equipment must be tested before being returned 

to service after being out-of-service for an extended period in accordance with TS requirements 

(if applicable). The OM Code provisions in Subsections ISTB-3420 and ISTC-3570 specify that 

licensees need not follow the test schedule if the system in which the component was installed 

was declared inoperable or was not required to be operable. However, this applies only if the 

component was not out-of-service for repair or replacement. For repair or replacement, the 

component must be tested within 3 months of the system being returned to service.
A 3-8

Basis for Recommendation

This recommendation is based on the standard technical specifications, which the NRC staff 

have developed, reviewed, and approved as staff technical positions. The specified intervals 

and extensions apply directly to IST of pumps and valves, as applicable. In Interpretation 

XI-78-01, the ASME Code Committee clarified the intent of the “2-year” frequency specified for 

verifying position indication and performing leak rate testing, stating that the Code test and 

examination frequency relates to periods of time, rather than refueling outages. The Code 

references refueling outages to preclude the necessity to shut down the plant solely for IST. 

The OM Code specifies that licensees must perform the valve position indicator test and leak 

rate test at least every 2 years, without regard to the frequency of refueling outages.

The NRC recommendation for testing during extended shutdown periods is consistent with TS 

and Code requirements, whichever are more conservative. Responding to inquiry IN 92-025A, 

the ASME Code Committee stated that Subsection ISTC-3510 intends that testing be conducted 

every 3 months, including during extended shutdown periods, for valves other than those 

declared inoperable in accordance with ISTC-3570.

The OM Committee made a similar clarification in OM Interpretation 93-1, stating that it is 

intended that testing be conducted every 3 months, including during extended periods, for 

valves other than those declared inoperable or not required to be operable. The ASME Code 

committee is in the process of developing a Code Case XX, to address this issue of test interval 

and 25% margin.

3.2 Start of the Time Period in Technical Specification Action Statements

For details see Standard Technical Specification LCO 3.0.2. 

3.3 120-Month Updates Required by 10 CFR 50.55a(f)(4)(ii)

After the initial 120-month interval, in accordance with 10 CFR 50.55a(f)(4)(ii), licensees must

conduct inservice tests during successive 120-month intervals to verify the operational 

readiness of pumps and valves within the scope of the ASME Code. In conducting these 

inservice tests, licensees must comply with the provisions of the latest edition and addenda of

the Code incorporated by reference in 10 CFR 50.55a(b) 12 months before the start of the

120-month interval, subject to the limitations and modifications listed in paragraph (b). In

addition, 10 CFR 50.55a(f)(5)(iv) specifies that where a pump or valve test requirement by the 

Code or addenda is determined to be impractical by the licensees and is not included in the new 

inservice test program interval, such that “the basis for this determination must be demonstrated 

to the satisfaction of the Commission not later than 12 months” from the start of the interval. 

Therefore, the staff recommends that licensees should submit relief requests for new intervals 

approximately 6 months before the start of the interval to allow adequate time for NRC review 

and approval.
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3.3.1 Extension of Interval

The IST interval may be extended in accordance with Subsection ISTA-3120(d):

Each IST interval may be extended or decreased by as much as 1 year. Adjustments 

shall not cause successive intervals to be altered by more than 1 year from the original 

pattern of intervals.

Subsection ISTA-3120(e) further states that, for units that are continuously out-of-service for 6

months or more, licensees may extend the IST interval during which the outage occurred for a 

period equivalent to the outage, and may extend the original pattern of intervals accordingly for 

successive intervals.

NRC Recommendation

In establishing the date for the new interval, licensees must establish the next updated program 

to the latest edition of the Code incorporated in the regulation 12 months before the new date. 

For example, if a licensee has an extension from December 14, 2000, to September 16, 2001, 

in accordance with the Code, the licensee’s program for the new interval must meet the edition 

of the Code incorporated in 10 CFR 50.55a(b) as of September 16, 2000. When extending its

10-year IST interval by as much as 1 year, as allowed by ISTA-3120(d), licensees may continue 

to apply the same Code edition and addenda from its current 120-month interval during this 

extended 1-year period. The staff recommends that licensees should inform the NRC of any 

extension before the date that would have been the end of the current interval. An extension 

beyond 1 year (other than for extended outages, as permitted by ISTA-3120(e)) requires NRC 

approval of an alternative to or exemption from the Code provisions or 10 CFR 50.55a, as 

applicable.

Basis for Recommendation

While it is not mandatory to maintain identical intervals for inservice inspection (ISI) and IST, it is 

often desirable in order to maintain the same edition of the Code for all plant activities related to 

ISI and IST. Even though 10 CFR 50.55a does not discuss extending the intervals, the Code is

incorporated by reference in the regulation and, therefore, has the same effect as the regulation. 

Although NRC approval is not required for 1-year extensions of the interval, licensees would 

avoid any discrepancies in the interval dates by informing the NRC of the extension and 

documenting it in the IST program document. Because the Code does not allow extension 

beyond 1 year (other than for extended outages), such an extension would require NRC 

approval of an alternative to or exemption from the Code provisions in order to comply with the 

regulatory requirements.
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3.3.2 Concurrent Intervals

Several licensees have established concurrent intervals for all units at sites with multiple units, 

so that each unit is updated to a newer edition of the Code at the same starting date. Because

the regulations do not specifically allow for adjustments to accommodate concurrent intervals 

among multiple units, when the interval start dates are to be concurrent, licensees may request 

a one-time alternative to or exemption from 10 CFR 50.55a or the Code, as applicable. If a 

licensee prefers not to request an alternative or exemption, the establishment of concurrent 

intervals would require that the licensee must perform program updates for a particular unit 

more often than once every 120 months. 10 CFR 50.55a(f)(4)(iv) permits IST of pumps and 

valves to meet the provisions in subsequent Code editions and addenda (or portions thereof) 

that are incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and 

modifications listed, and subject to Commission approval. This regulation allows licensees to 

update their programs before the end of a 120-month interval with Commission approval.

NRC Recommendation

If a licensee elects to use the same Code edition for multiple units, the regulation requires that

the licensee must request an alternative to or exemption from the regulation or the Code, as

applicable, to extend a unit’s interval by more than 1 year in order to place multiple units on a

concurrent interval for IST. To establish concurrent intervals without an alternative or

exemption, the licensee must update the referenced edition of the Code more frequently for the 

selected unit(s) to remain in compliance with 10 CFR 50.55a, except in the case where the 

interval dates are within 12 months, whereby the Code allowance for an extension would result 

in concurrent intervals. If the licensee elects to use 10 CFR 50.55a(f)(4)(iv) to update to later 

editions of the Code, this section gives the requisite approval for the IST program, if the licensee 

uses the following guidelines:

(1) Without obtaining an alternative or exemption, the licensee may perform the IST 

program for multiple units using the same edition of the Code at concurrent 

intervals if the initial interval for combining the programs is established such that 

no single unit is tested at an interval of more than 120 months (or no greater than 

the interval extension allowed by the Code). Thus, the licensee must use the 

interval for the first unit that was licensed for commercial operation to establish 

the interval dates and establish the correct Code edition according to the most 

recent required for either unit.

(2) To exceed 120 months, other than as addressed in the Code for an extension, 

the licensee must first obtain approval of an alternative to or exemption from 10

CFR 50.55a; therefore, establishing two units on the same interval requires an 

alternative or exemption unless the licensee intends to repeatedly update both 

units more often than the required 120 months. That is, the licensee will test 

each unit according to the most recent edition of the Code required for either unit.

The IST program document and the request for the alternative or exemption would typically 

describe the method for selecting the interval dates, specifying the dates at which the interval 

will begin and end, and comparing the effect of those dates with that of the dates that would 

otherwise be required.
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Basis for Recommendation

By obtaining an alternative or exemption, a licensee may test multiple units at the same interval. 

However, the licensee may choose to update without requesting an alternative or exemption in

order to periodically maintain an IST program with a more current edition of the Code.

The regulations allow for concurrent intervals among multiple units if the licensee updates the

program each time an interval is due for either unit. The preferred manner for establishing 

concurrent intervals is to consider each unit on the same 120-month interval, with updates 

occurring at the end of the 120 months through an alternative or exemption. While the example 

presented is acceptable for IST programs in which test frequencies for components other than 

safety and relief valves do not exceed 18 to 24 months, licensees might need to adjust this

arrangement for IST intervals that are longer than 5 years, such as might be found in riskinformed and performance-based testing programs.

3.3.3 Implementation of Updated Programs

Updating the IST program to a revised edition and addenda of the Code is an extensive effort 

that involves changes to administrative and implementing procedures. Often, the revised 

requirements will necessitate establishing new reference values, such as by implementing a

vibration program using velocity measurements rather than displacement measurements for

pump testing. Implementing a new comprehensive pump test may be necessary for parameters 

that are not currently measured. New “reference values” for currently monitored parameters 

may not be necessary if previous reference values were acceptable. However, the Code does 

not specifically require licensees to establish new reference values simply because a later 

edition is used.

NRC Recommendation

The NRC staff recommends that, before beginning the first tests during the new interval, 

licensees should revise the implementing procedures according to the appropriate 

requirements. When the testing requires baseline values to be reestablished to meet Code 

changes, this would typically involve establishing the new baseline (reference) values during the 

first quarterly or cold shutdown outage test performed in the new interval, if not before. Before

performing tests during the first refueling outage, licensees would typically revise implementing 

procedures for the tests to be performed during that outage to incorporate any new 

requirements or components.

Before or during startup from the refueling outage, licensees must complete all tests that are

required to be performed during the refueling outage, as required by the Code (ISTC 3500 and 
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Appendix I, Section I-1300). If a licensee determines that timely implementation is not possible, 

the staff recommends that the licensee should submit a schedule to the NRC (1) before the 

beginning of the interval, or (2) before startup from the refueling outage if the interval begins 

while a plant is shut down for refueling.

For 120-month updated programs, the staff recommends that licensees should submit relief

requests before the interval start date to allow adequate time for NRC review and approval 

within 12 months after the interval start date (i.e., submit the updated program at least 3 - 6

months before the start date.)

In accordance with the regulations, when updating a program to a later edition of the ASME 

Code, licensees must implement the updated program at the beginning of a 120-month interval. 

The regulations state that, where a pump or valve test specified by the Code is determined to be

impractical and is not included in the revised IST program, the licensee must demonstrate the 

basis for the determination to the satisfaction of the Commission not later than 12 months after 

each 120-month interval. However, experience has shown that licensees also identify 

impractical test provisions throughout the interval. In such cases, the staff recommends that

licensees should request relief as soon as they identify the condition. Because the 

requirements are impractical, the licensee would test the applicable components using the

method proposed in the relief request in the period of time from the beginning of the new 

interval (or from the time of identification) until the NRC staff completes its evaluation (e.g., if a

licensee identifies a solenoid valve that is within the scope of the IST program and is stroke-time 

tested but has no position indication, the licensee cannot meet the Code requirements because 

of design limitations and an alternative method may not comply with the Code requirements). 

Relief requests that relate to proposed alternatives to the Code requirements (rather than to 

“impractical” requirements) shall not be implemented until the NRC staff completes its 

evaluation (e.g., if a licensee proposes to implement a pump vibration program based on using 

spectral analysis, rather then the Code-specified method, the licensee must continue to meet 

the Code requirements until the NRC staff completes its evaluation).

Basis for Recommendation

When licensees update their IST programs to a revised edition and addenda of the Code, the

NRC staff recognizes that changes might be completed over a period of time to allow for 

adequate review and approval; however, the staff recommends completing the procedural 

revisions in a timely manner. The regulations do not allow a licensee to continue with a

previous program while waiting for NRC review and approval of the relief requests and

proposed alternatives for the next interval.

3.3.4 General Comments on Inservice Testing Intervals

The NRC has received requests for IST programs and partial submittals that lack the dates of

the intervals or the Code edition in use. Some individuals responsible for the IST programs 

were not aware that the Code of Federal Regulations is updated throughout the year through 

issuance of the Federal Register. Therefore, when those individuals revised their programs, 

they used the bound version of 10 CFR Part 50 to determine the Code edition cited in 10 CFR

50.55a(b) 12 months before the interval start date. However, a more recent edition may have 
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been incorporated by reference in 10 CFR 50.55a(b) as noticed in the Federal Register, which 

may have resulted in the program being developed to an incorrect edition of the Code.

Additionally, several plants have asked questions concerning phasing in the updated program. 

Generally, this is an acceptable approach for testing if the program does not involve any

requests for relief from Code requirements.

NRC Recommendation

The NRC staff recommends that licensees should include the interval dates and Code edition in

each IST submittal, regardless of whether it is for an entire program or only a partial submittal 

containing new or revised relief requests. The staff must ensure that the interval dates are

correct and that the evaluation is performed using the edition of the Code from which the

licensee is requesting relief. The staff also recommends that the individuals responsible for

developing and maintaining the IST program should be aware of the regulatory changes made 

in 10 CFR 50.55a throughout the year, and should review any new or revised requirements for

any effect on the IST program.

For phasing in a new edition or addenda of the Code before the start of a new interval (or during

an ongoing interval), the staff recommends that licensees should submit a plan and schedule to 

the NRC. If there are no issues that require NRC review, the testing can be phased into the 

appropriate edition of the Code (1) during the 12 months prior to the interval start date, or (2) 

during any time period identified by the licensee up to an interval start date, if the phasing-in 

begins in the middle of an interval and a licensee wants to use an edition of the Code that is 

more recent than that incorporated by reference in 10 CFR 50.55a(b).

Basis for Recommendation

The NRC has noted incorrect interval dates and Code editions cited in IST program submittals. 

The Code stipulates that licensees shall calculate the inspection interval according to the

number of calendar years that have passed since the power unit was placed into commercial

service. For information purposes, the annual “NRC Information Digest” (NUREG-1350) lists 

the licensing and commercial operation dates for nuclear power plants. For convenience, the 

licensees for several plants have established the initial interval as beginning on the date of their 

operating licenses or some other unspecified milestone. However, if the NRC revised 

10 CFR 50.55a after the interval start date cited by the licensee and before the date of the 

operating license, and if the revision of 10 CFR 50.55a incorporated a later edition of the Code, 

the regulations may have required use of a more recent edition than the licensee actually used. 

Therefore, it is important that the IST program document state the Code edition and addenda 

used to develop the program.

3.4 Skid-Mounted Components and Component Subassemblies

The Code-class piping systems at a plant may include skid-mounted components or component 

subassemblies, such as valves in diesel air-start subassemblies, diesel skid-mounted fuel oil 

pumps and valves, steam admission and trip throttle valves for HPCI or auxiliary feedwater 

(AFW) pump turbine drivers, steam traps, and air supply system check valves and solenoidoperated valves for main steam isolation valves. If the licensee’s safety analysis report (SAR) 

identifies
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these components as ASME Code Class 1, 2, or 3, they are subject to IST. By contrast, if the 

SAR does not identify these components as ASME Code Class 1, 2, or 3 (or indicates that they 

are maintained as Code class, but are not required to be Code class), they are not subject to 

IST in accordance with 10 CFR 50.55a. Nonetheless, these components may be subject to 

periodic testing in accordance with Appendices A and B to 10 CFR Part 50.

NRC Recommendation

Subsections ISTB-1200(c) and ISTC-1200(c) define the components that are subject to IST. 

The staff has determined that testing the major component is an acceptable means to verify the 

operational readiness of the skid-mounted components and component subassemblies if the 

licensee discusses this approach in the IST program document. Licensees should consider and 

document the specific measurements and attributes of major component testing which relate to 

the assessment of skid-mounted component condition. In addition, various continuous and 

periodic observations of the major components (such as System Monitoring Walkdowns or 

Operator Logs) may also support assurance of skid-mounted component readiness. This is

acceptable for both Code class components and non-Code class components that are tested 

and tracked by the IST program.

Basis for Recommendation

Various pumps and valves that are procured as part of larger component subassemblies are

often not designed to meet the requirements for components in ASME Code Classes 1, 2, and 

3. In Regulatory Guide 1.26, “Quality Group Classifications and Standards for Water-, Steam-, 

and Radioactive-Waste-Containing Components of Nuclear Power Plants,” the NRC gives 

guidance on classifying components for quality groups A, B, C, and D (Code Classes 1, 2, and 

3, and ASME VIII/ANSI B31.1, respectively). (For additional guidance, licensees should review 

Section 3.9.6 of NUREG-0800, the NRC’s Standard Review Plan.) When many of the 

components were procured, the requirements for IST did not apply and, thus, the components 

may not have included features for IST. Licensees may, therefore, elect to use the IST program 

for testing these components and state in the IST program document that the surveillance tests 

of these components adequately test the skid-mounted components.

The OM Code addresses both components that are physically mounted on the skid, and those 

that are not mounted on the skid but function the same as skid-mounted components (e.g., 

check valves in the service water system that supply cooling water to a pump), provided that 

testing the major component is adequate to test the function of the system component.

For components that are outside the scope of 10 CFR 50.55a, relief requests are not necessary. 

The NRC’s position concerning testing components that are outside the scope of

10 CFR 50.55a is discussed in Section 2.2.3. Testing of skid mounted check valves are 

specifically discussed in Section 4.1.10.

3.5 Pre-Conditioning of Pumps and Valves

3.5.1 Background

The regulations in 10 CFR 50.55a require licensees to test pumps and valves at nuclear power 

plants to assess their operational readiness within the scope of the ASME OM Code. Criterion
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XI, “Test Control,” in Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel 

Reprocessing Plants,” to 10 CFR Part 50 specifies that licensees must establish a test program 

to ensure that all testing required to demonstrate that structures, systems, and components 

(SSCs) will perform satisfactorily in service is identified and performed in accordance with 

written test procedures that incorporate the requirements and acceptance criteria contained in 

applicable design documents. Criterion XI further requires that test procedures must include 

provisions to ensure that (1) all prerequisites for the given test have been met, (2) adequate test 

instrumentation is available and used, and (3) the test is performed under suitable 

environmental conditions. Criterion XI then requires licensees to document and evaluate the 

test results to ensure that the test requirements have been satisfied. In order to effectively 

assess operational readiness, the performance of the tested pump or valve, and the conditions 

under which the pump or valve must be capable of performing its safety function, need to be 

fully understood. Any maintenance activities performed before actual inservice testing are

called “preconditioning” or “grooming” and this will adversely affect the validity of the test results.

3.5.2 NRC Guidance

In Information Notice (IN) 97-16, “Preconditioning of Plant Structures, Systems, and 

Components Before ASME Code Inservice Testing or Technical Specification Surveillance 

Testing,” the NRC staff discussed the longstanding concern regarding unacceptable 

preconditioning of plant SSCs before testing. The NRC staff noted that experience has 

demonstrated that some testing cannot be performed without disturbing or altering the 

equipment. The staff also indicated that any such disturbance or alteration would be expected 

to be limited to the minimum necessary to perform the test and to prevent damage to the 

equipment. In addition, the staff alerted licensees that, in certain cases, the safety benefit of 

some preconditioning activities might outweigh the benefits of testing in the as-found condition.

The staff has provided guidance to the NRC’s regional offices and inspectors with respect to 

preconditioning of plant equipment prior to ASME Code and TS testing. This guidance is found

in the following documents (see Section 9 of Appendix A for the locations of these documents):

• NRC memorandum, dated July 2, 1996, from Frederick J. Hebdon, Director, Project 

Directorate II-3, Division of Reactor Projects I/II, Office of Nuclear Reactor Regulation, to 

Jon R. Johnson, Acting Director, Division of Reactor Projects, Region II, in response to 

Technical Assistance Request TIA 96-007: “Regulatory Acceptability of Lubricating 

Valves Prior to Surveillance Testing”

• NRC Inspection Manual, Part 9900, “Technical Guidance: Maintenance -

Preconditioning of Structures, Systems, and Components Before Determining 

Operability”

• Attachment 22, “Surveillance Testing,” to IP 71111, “Reactor Safety: Initiating Events, 

Mitigating Systems, Barrier Integrity”

The guidance provided in these documents is instructive for nuclear plant personnel in providing

assurance that testing conducted as part of the IST program is capable of assessing the 

operational readiness of pumps and valves within the scope of the ASME Code.
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NRC Inspection Manual, Part 9900, defines preconditioning as the “alteration, variation, 

manipulation, or adjustment of the physical condition of an SSC before Technical Specification 

surveillance or ASME Code testing.” The licensee may consider several factors in determining 

whether an activity constitutes acceptable preconditioning of a pump or valve to be tested. For

example, an activity would constitute acceptable preconditioning of a pump or valve if it is

performed to protect personnel or equipment, or to meet the manufacturer s recommendations 

(or based on industry-wide operating experience). If a preventive maintenance activity (such as

valve stem lubrication or pump venting) periodically occurs prior to testing, the licensee might 

justify the acceptability of this infrequent preconditioning of a pump or valve if the licensee 

evaluates the effect of the activity on the overall ability to assess the operational readiness of 

the pump or valve, and to trend degradation in its performance. As noted in the inspection 

guidance, the licensee should have evaluated and documented the activity as acceptable 

preconditioning before performing the testing.

Some activities would constitute unacceptable preconditioning of a pump or valve to be tested 

under the IST program. NRC Inspection Manual, Part 9900, defines unacceptable 

preconditioning of an SSC as an activity that alters one or more of the SSCs operational 

parameters and, thereby, results in acceptable test results. For example, a preventive 

maintenance activity might constitute unacceptable preconditioning of a pump or valve if the 

licensee routinely conducts the activity prior to testing. NRC Inspection Manual, IP 71111, 

Attachment 22, instructs NRC inspectors to evaluate the acceptability of any preconditioning of 

equipment in preparation for surveillance tests. Similarly, NRC inspectors also verify that

licensees do not routinely schedule preventive maintenance activities prior to testing in order to 

help ensure that the test is passed satisfactorily. In addition to activities related to an individual 

pump or valve, maintenance or surveillance activities involving several SSCs, including the 

scheduling or timing of such activities, can inadvertently result in unacceptable preconditioning 

of a pump or valve.

NRC Inspection Manual, Part 9900, provides a series of questions that NRC inspectors should

consider when evaluating the acceptability of an activity that appears to involve preconditioning 

of a plant SSC. With respect to pumps and valves, those questions can be interpreted as 

follows:

• Does the practice performed ensure that the pump or valve will meet its testing 

acceptance criteria?

• Would the pump or valve have failed the test without the preconditioning?

• Does the practice bypass or mask the as-found condition of the pump or valve?

• Is preventive maintenance routinely performed on the pump or valve just before testing?

• Is preventive maintenance on the pump or valve performed only for scheduling 

convenience?

According to NRC Inspection Manual, Part 9900, an activity constitutes unacceptable 

preconditioning if an affirmative answer is determined in response to any of these questions, 

and the activity meets the definition of unacceptable preconditioning provided in the inspection 
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guidance. Licensees are encouraged to consider such questions as part of their determination 

of whether an activity related to a pump or valve in their IST program constitutes unacceptable 

preconditioning.

3.5.3 ASME Code Guidance

The ASME Code relies on the licensee to determine whether an activity would constitute 

acceptable or unacceptable preconditioning of a pump or valve prior to testing under its IST

program, except in a few limited instances. One such instance is found in Sections I-3300 and 

I-7300 of the OM Code’s mandatory Appendix I, “Inservice Testing of Pressure Relief Devices in 

Light-Water Reactor Nuclear Power Plants,” which specify that no maintenance, adjustment, 

disassembly, or other activity that could affect as-found set-pressure or seat tightness data for 

pressure relief devices is permitted before testing. Another instance is found in Section 3.3 of 

ASME Code Case OMN-1, “Alternative Rules for Preservice and Inservice Testing of Certain 

Electric Motor-Operated Valve Assemblies in Light-Water Reactor Power Plants, OM Code1995, Subsection ISTC,” which specifies that inservice tests of motor-operated valves shall be 

conducted in the as-found condition, and that maintenance activities shall not be conducted if 

they might invalidate the as-found condition for inservice testing. The 2009 Edition of the ASME 

OM incorporates Code Case OMN-1 (including the as-found testing provision), and OMN-11, 

“Risk Informed Testing of Motor-Operated Valves,” as Appendix III to the OM Code, to replace 

quarterly MOV stroke-time testing with periodic exercising and diagnostic testing (Note: The 

details related to the ASME OM-2009 is for information only). Where the ASME Code does not 

specify provisions for as-found testing of a pump or valve, the licensee is responsible for 

determining whether an activity constitutes acceptable or unacceptable preconditioning of a 

pump or valve to be tested under its IST program.

3.5.4 NRC Recommendation

The NRC staff has provided examples of acceptable and unacceptable preconditioning of plant

components prior to testing in such documents as IN 97-16 and NRC Inspection Manual, Part

9900. Where the ASME Code does not provide specific provisions related to as-found testing of 

a pump or valve in the IST program, the staff considers acceptable preconditioning to include 

such activities as (1) periodic venting of pumps, which is not routinely scheduled directly prior to 

testing but may occasionally be performed before testing; (2) pump venting directly prior to 

testing, provided that the venting operation has proper controls with a technical evaluation to 

establish that the amount of gas vented would not adversely affect pump operation; 

(3) occasional lubrication of a valve stem prior to testing of the valve, where stem lubrication is

not typically performed prior to testing; and (4) unavoidable movement attributable to the setup 

and connection of test equipment. In each instance of acceptable preconditioning, the staff will 

expect the licensee to have available a documented evaluation of the preconditioning activity 

and a justification for continued confidence in the capability of the IST program to assess the 

operational readiness of the pump or valve. Generic evaluations may be acceptable as long as 

the evaluation bounds the conditions of specific activity performed on the SSC.

By contrast, the staff considers unacceptable preconditioning of pumps and valves in the IST

program to include such activities as (1) routine lubrication of a valve stem prior to testing the

valve; (2) operation of a pump or valve shortly before a test, if such operation could be avoided 

through plant procedures with personnel and plant safety maintained; and (3) venting a pump
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immediately prior to testing without proper controls and scheduling. Licensees may evaluate 

applicable NRC staff documents to determine whether specific activities prior to testing 

constitute acceptable or unacceptable preconditioning of a pump or valve in the IST program. 

The NRC staff encourages licensees to contact their NRC resident inspector or NRR project 

manager if questions arise regarding potential preconditioning of a pump or valve to be tested 

under the IST program.

3.6 Testing in the As-Found Condition

The intent of IST is primarily to detect and monitor the degradation or rate of changes of a 

component after a period of operation, or stand-by conditions. I staff review of IST programs 

during 1985 through 1991, section XI, IWP and IWV, and Part 6 and Part 10 of OM-1987 had 

been interpreted to require that IST be performed in the “as-found” condition. However, later 

OM Codes, with the exception of Appendix I for safety/relief valves, do not specifically require 

licensees to test components in the “as-found” condition. If “as-found” test were not performed, 

degradation mechanisms or rate of degradation resulting from previous period of operation or 

stand-by condition can not be identified and obtained. Therefore, this section should ensure

that the intent of IST be retained. 

The Code does not specifically require licensees to test components in the “as-found” condition 

(except for safety and relief valves and, in the 2009 Edition of the ASME OM Code, MOVs)

(Note: The details related to the ASME OM-2009 are for information only). Sections 1300, 1350 

and 3300 of the ASME OM Code’s mandatory Appendix I require licensees to measure the 

initial lift of safety relief valves to determine whether additional valves are to be tested. In 

addition, Sections 3300 and 7300 of Appendix I specify that licensees must periodically test all

pressure relief devices and may not perform any maintenance, adjustment, disassembly, or

other activity that could affect the as-found set pressure or seat tightness data before testing.

The “as-found” condition is generally considered to be the condition of a valve without prestroking or maintenance. The ASME OM Code does not require stroke-time testing or check 

valve stroking prior to maintenance; however, degradation mechanisms may not be identified if

the licensee does not perform any as-found testing. However, the staff encourages licensees to 

perform as-found testing, where practical. The staff also cautions licensees to consider the 

timing of maintenance with regard to the required test intervals and the potential for 

pre-conditioning. Post-maintenance testing is required when the maintenance could have 

affected the valve s performance. As-found testing may also apply to pumps in a similar 

fashion. Most inservice testing is performed in a manner that generally represents the condition 

of a standby component if it were actuated in the event of an accident (i.e., no pre-conditioning 

prior to actuation).

3.7 Testing at Power

In an effort to shorten refueling outages, many licensees are trying to perform as much 

maintenance, testing, and surveillance as possible with the nuclear power plant on line. For

example, several licensees have submitted relief requests to obtain NRC approval to conduct 

inservice testing once per refueling cycle, rather than during the refueling outage as prescribed 

by the Code. In preparing (and evaluating) such relief requests, licensees (and the NRC staff) 

should consider several factors to ensure that the licensee s proposed alternative provides an 

acceptable level of quality and safety.
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If a licensee is testing a particular pump or valve during refueling outages, the licensee may

have determined that it is impractical to test the pump or valve quarterly during operation. The 

licensee’s IST program document should, therefore, discuss the basis for deferring the testing 

from quarterly (and during cold shutdowns) to refueling outages. Relief requests to perform 

testing once each refueling cycle with the nuclear power plant on line should be prepared in lieu 

of the refueling outage justification for each affected valve or group of valves. If necessary, the 

licensee should revise the refueling outage justification to be consistent with the relief request.

Licensees (and the NRC staff) should also consider whether the testing can be accomplished 

within the allowed outage time permitted by any applicable technical specification. In general, 

the time necessary to complete the testing should be significantly less than the allowed outage 

time. This is to preclude TS violations or the need to issue exigent TS amendments or notices 

of enforcement discretion (NOEDs). In addition, licensees should not conduct non-corrective 

maintenance/testing activities at power if the associated post-maintenance testing cannot

reasonably be accomplished until the next outage.

Sometimes, there is a tradeoff between testing these components at power (e.g., when they 

could be needed to mitigate the consequences of an accident) and testing them during outages 

(e.g., when there may be greater reliance on shutdown cooling or when other equipment is

necessarily out-of-service). Licensees should quantitatively or qualitatively address the risks 

associated with testing components on line, rather than testing during the refueling outage. If

the proposed testing could have a significant risk impact, or if its justification includes riskrelated arguments, the relief request should be prepared and reviewed in accordance with 

RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on

Plant-Specific Changes to the Licensing Basis,” and Appendix D to SRP Chapter 19, as

applicable. Licensees should also identify any compensatory measures to be established as a 

means to reduce the impact (e.g., risk and operational worker safety) of testing with the nuclear 

power plant at power.1 

If relevant, licensees should also provide information on how testing at power (rather than 

testing during refueling outages) will affect scheduled maintenance work windows for the 

applicable system (i.e., whether the testing can be completed within the work windows or 

whether it will extend either the shutdown or at-power work windows). In addition, licensees will 

need to develop a new estimate of the maintenance unavailabilities that reflects the increased 

maintenance activities at power, and will need to document the basis for the new estimate (e.g., 

use plant logs or maintenance data to include in the current estimate of the maintenance 

unavailabilities those activities that were being performed during shutdown that will now be 

performed at power) 1

.

1 It should be noted that the assessment of risk resulting from performance activities as required by 10 CFR 

50.65(a)(4) of the Maintenance Rule is not sufficient justification for testing components at power. This 

assessment is required for maintenance activities performed during power operations or during shutdowns. 

This configuration risk management does not address the relative merits of testing at power verses testing 

during refueling outages.
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At times, testing (or the disassembly and inspection of components) during refueling outages 

can be more advantageous than at-power operations from a worker safety perspective (for

example, systems may be cold and depressurized). When requesting NRC approval to perform 

testing with the nuclear power plant on line, licensees should consider worker safety and should 

discuss whether the applicable components can be adequately isolated and restored.

In Section 11.2.3 of NUMARC 93-01, Rev. 2, “Industry Guideline for Monitoring the 

Effectiveness of Maintenance at Nuclear Power Plants,” the Nuclear Management and 

Resources Council (NUMARC, now NEI) provided additional guidance for conducting online 

maintenance and testing. It states, in part –

Online maintenance [and testing] should be carefully managed to achieve a balance 

between the benefits and potential impacts on safety, reliability or availability. For 

example, the margin of safety could be adversely impacted if maintenance is performed 

on multiple equipment or systems simultaneously without proper consideration of risk, or 

if operators are not fully cognizant of the limitations placed on the plant due to out of 

service equipment. Online maintenance should be carefully evaluated, planned and 

executed to avoid undesirable conditions or transients, and to thereby ensure a 

conservative margin of core safety.

3.8 Potential Adverse Impact on Plant Components from Flow-Induced Vibration

Nuclear power plant operating experience has revealed that plant components can be adversely 

affected by flow-induced vibration caused by hydrodynamic and acoustic loading in nuclear 

power plant fluid systems. For example, BWRs have experienced damage to plant components 

in main steam systems as a result of vibration caused by acoustic resonance during power 

uprate operation. Flow-induced vibration has also resulted in damage to components in the 

reactor coolant system at a PWR plant. Licensees of new nuclear power plants, and licensees 

of currently operating plants planning to implement a power uprate, should ensure that the 

potential for adverse effects on plant components from flow-induced vibration is addressed 

during plant startup or power uprate ascension.

3.9 Repair of Pumps and Valves for New Reactors Under Construction

In a final rule issued on August 28, 2007 (72 FR 49352), the NRC amended its regulations to 

clarify the applicability of various requirements to each of the licensing processes by making 

necessary conforming amendments throughout NRC’s regulations to enhance the NRC’s 

regulatory effectiveness and efficiency in implementing its licensing and approval process for 

new reactors licensed under 10 CFR Part 52. One of these changes included a revision to 

10 CFR 50.55a in which the NRC clarified how the regulations in § 50.55a apply to approvals, 

certifications, and licenses issued under 10 CFR Part 52. The NRC modified § 50.55a to state 

that each combined license for a utilization facility is subject to the conditions in § 50.55a, but is 

only subject to the conditions in §§ 50.55a(f) and (g) after the NRC makes the finding under § 

52.103(g). This 52.103(g) finding is commonly associated with the date of initial fuel loading. 

Essentially, the effect of these changes to § 50.55a is that combined license plants are subject 

to the requirements of Section III of the ASME Boiler and Pressure Vessel Code (ASME B&PV 

Code) from the date of issuance of the combined license during which the plant is under 

construction, and are subject to the requirements of Section XI of the ASME Code only after 

initial fuel loading.
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As a result of these changes to § 50.55a, the question, “Which repair requirements (i.e., those 

under Section III or XI) apply to a combined license plant under construction?” was identified by 

industry and NRC staff in public meetings on ITAAC maintenance. In operating plants, repairs 

to ASME Code components are made using the rules of Section XI of the ASME Code; not 

Section III. The industry questioned whether the requirements of Section XI may be used when 

making repairs to an installed component in a plant under construction after the system or 

component has received its Code Symbol Stamp, but before the Commission makes its finding 

under § 52.103(g). As a result, the NRC staff brought these questions to the ASME Boiler and 

Pressure Vessel Code committees for their interpretation and possible action.

In 2010, the ASME Boiler Code committees examined the history and background of this issue 

and found that the ASME Boiler and Pressure Vessel Code permits repairs using either Section 

III or XI. However, the ASME Boiler and Pressure Vessel Code did not provide clear 

requirements on how repairs are to be accomplished under Section III. As a result, the Code 

committees developed definitive rules clarifying how repairs are to be performed under Section 

III for a plant under construction. Although the initiative was originally intended for application to 

new reactors, a similar and more urgent concern also existed for the Watts Bar Nuclear Plant, 

Unit 2 which recently restarted its construction activities and committed to use Section III for its 

plant’s repairs during construction. These definitive rules were issued in ASME Code Cases 

N-801 and N-802 to address the concerns at Watts Bar, Unit 2. ASME Code Case N-802 

clarifies how repairs shall be made at the plant site by the organization that originally 

manufactured or certified the component, and ASME Code Case N-801 clarifies how repairs 

shall be made at the plant site by organizations other than the original component manufacturer. 

ASME Code Case N-801 was needed at Watts Bar, Unit 2 because many manufacturers of 

components installed in the Watts Bar plant are no longer in the nuclear business. ASME Code 

Case N-801 is equally important for new reactors in which major plant components may be 

manufactured and certified in other countries, and for which it might be impracticable for those 

original foreign manufacturers to perform repairs once the component is installed in a U.S. plant. 

The NRC, therefore, finds Code Cases N-801 and N-802 to be applicable and appropriate for 

use by new reactors licensed under Part 52. 

The NRC is currently preparing an amended rule in 10 CFR 50.55a to require the use of ASME 

Code Cases N-801 and N-802 by new reactors under construction and licensed under 

10 CFR Part 52. The amended rule might include certain conditions on its use that will be 

clarified when the final rule is issued in early 2013.
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4. SUPPLEMENTAL GUIDANCE

ON INSERVICE TESTING OF VALVES

The NRC staff has developed the following recommendations for valves that may be a part of 

an IST program at nuclear power plants. The types of valves discussed herein are covered by 

the ASME OM Code. Specifically, these include check valves (Section 4.1), power-operated 

valves (Section 4.2), safety/relief valves (Section 4.3), and other miscellaneous valves 

(Section 4.4).

4.1 Check Valves

The NRC staff considers check valves, and other automatic valves designed to close without 

operator action after an accident and for which flow is not blocked, as “active” valves that would

be classified as such in the IST program (reference, for example, Section B 3.6.3 of the

NUREG-1431, Revision 3 Westinghouse Standard Technical Specifications). Similar criteria 

could be applied to the opening function of a check valve. The flow through a check valve 

would be blocked by any condition that precludes flow through the system. For example, 

installing a flange or closing another valve (other than a check valve) in the line would block 

flow. A valve that is “positively held in place” would be one that has an operator or other 

auxiliary device that maintains the disk in an open or closed position, such as a stop check 

valve. SECY-77-439, “Single Failure Criterion,” dated August 17, 1977, which was referenced 

in several plants’ licensing bases, discusses the failure of a check valve to move to its correct 

position as a passive failure; however, this does not correspond to the issue of “active” versus 

“passive” for the purpose of IST.

The ASME Code defines valves that are self-actuating in response to some system 

characteristic, such as flow direction, for fulfillment of the required function(s) as Category C 

valves. The Code also defines valves for which seat leakage is limited to a specific maximum 

amount in the closed position for fulfillment of their required function(s) as Category A valves. 

Those check valves (Category C valves) that must also be leak-tight (Category A valves) would

be designated as Category A/C in the IST program.

Whereas the Code only requires licensees to exercise Category C check valves on a periodic 

basis, Category A/C check valves must be leak tested in addition to being exercised. The NRC

staff has found that, in many instances, licensees are not assigning check valves to Category 

A/C, despite the fact that the licensees take credit for the check valve providing an essentially 

leak-tight function.

The categorization of a check valve is not solely dependent on the function performed by the

valve, such as whether it is a containment isolation valve. When determining the proper

categorization of a check valve, a licensee should consider all applicable aspects. For example, 

the licensee should determine (1) whether the flow requirements for connected systems can be 

achieved with the maximum possible leakage through the check valve, (2) the effect of any 

reduced system flows resulting from the leakage on the performance of other systems and 

components, (3) the consequences of the loss of water from the system, (4) the effect that 

backflow through the valve may have on piping and components, such as the effect of high 

temperature and thermal stresses, and (5) the radiological exposure to plant personnel and the 

public caused by the leak. If any of the above considerations indicate that Category C testing 
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may not be adequate, licensees should assign the check valve to Category A/C and should 

comply with the associated leak testing requirements.

Licensees may refer to NRC Information Notice 91-56, “Potential Radioactive Leakage to Tank 

Vented to Atmosphere,” dated September 19, 1991, for information on the categories assigned 

to valves that function to close. These valves may also function to prevent leakage above an

assumed limit to prevent the plant from exceeding the limits in 10 CFR Part 100. Section 4.1.1 

herein discusses backflow testing of check valves in series.

Subsection ISTC-3550 of the OM Code discusses valves in regular use and states that valves 

that operate in the course of plant operation at a frequency that would satisfy the exercising 

requirements need not be additionally exercised, provided that the observations otherwise 

required for testing are made and analyzed during such operation, and recorded in the plant 

record at intervals no greater than specified in Subsection ISTC-3510. Even if licensees 

“exercise” check valves in accordance with Subsection ISTC-3550, they need to be included in 

the valve list in the IST program document, and the record (e.g., plant log, test procedure) 

needs to indicate that the test requirements are met.

For grouping valves in multiple nuclear power plants of like design and construction (e.g., Point

Beach Nuclear Plant, Units 1 and 2), if the plants are “identical” and the grouped valves have 

similar operational experience and otherwise meet the grouping criteria, it is acceptable for

licensees to group valves from multiple plants. If a potentially generic problem is identified 

through disassembly and inspection during a refueling outage, the licensee must inspect all 

valves in the group in that plant during the refueling outage. If the other plant is not also in a 

refueling outage, inspection of the valves in the group that are installed in that plant may be 

deferred to the next refueling outage unless the licensee s evaluation of the problem indicates 

that it could impact the safety of continued operation. “Grouping” may also be applied to the 

use of nonintrusive techniques as discussed in Section 4.1.2 (below), although the focus is 

slightly different, in that all of the valves in the group are tested, while the nonintrusive 

techniques are applied to only one valve of the group; therefore, all valves in the group must be 

in the same nuclear power plant.

The NRC issued the following information notices (INs) on IST for check valves:

• IN 82-08 “Check Valve Failures on Diesel Generator Engine Cooling System”

• IN 83-03 “Check Valve Failures in Raw Water Cooling System of Diesel Generators”

• IN 83-54 “Common Mode Failure of Main Steam Isolation Non-Return Check Valves”

• IN 88-70 “Check Valve Inservice Testing Program Deficiencies”

• IN 2000-21 “Detached Check Valve Disk Not Detected by Use of Acoustic and 

Magnetic Nonintrusive Test Technique”

4.1.1 Closure Verification for Series Check Valves without Intermediate Test 

Connections

Some plants have piping configurations that include two check valves in series with no provision 

for verifying that each valve can close. These valves may perform a safety function in the

closed position. For example, the valves may be required to prevent the gross diversion of flow 

rather than to be leak-tight. The Code requires that each valve that performs safety functions 

must be stroked to the position(s) required for the valve to perform those functions. The
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requirements for testing two check valves in a series configuration are addressed in Subsection

ISTC-5223, as follows:

ISTC-5223, “Series Valve Pairs,” If two check valves are in a series configuration without 

provisions to verify individual reverse flow closure (e.g., keepfill pressurization valves) and the 

plant safety analysis assumes closure of either valve (but not both), the valve pair may be 

operationally tested closed as a unit. If the plant safety analysis assumes that a specific valve 

or both valves of the pair close to perform the safety function(s), the required valve(s) shall be 

tested to demonstrate individual valve closure.

NRC Recommendation

Both valves in a series pair must be tested to demonstrate individual valve closure if the plant 

safety analysis credits or otherwise requires both valves. For example, GDC 14 in Appendix A 

to 10 CFR Part 50 requires licensees to test the valves in the reactor coolant pressure boundary 

to demonstrate extremely low probability of abnormal leakage. Pressure isolation valves are a 

special case of reactor coolant pressure boundary valves, which are generally required to be 

individually leakage tested at a frequency specified by the TSs and the Code.

Systems containing series pair valves may have provisions for verifying that at least one valve is

capable of closing. These provisions enable the licensee to measure or observe operational 

parameters such as leakage, pressure, or flow during each quarter, each cold shutdown outage, 

or each refueling outage. However, testing the series pair as a unit provides no assurance that

both valves close. The only indication of a problem would be the failure of both valves in the 

series. If the valve pair is operationally tested closed as a unit, as allowed by Subsection

ISTC-5223, because the plant safety analysis assumes closure of either valve (but not both), 

and the tested unit closure capability is questionable, both valves must be declared inoperable 

and corrective actions must be taken for both valves.

If it is not practical to flow test the pair of valves in accordance with the Code, the licensee may

demonstrate the closure safety function of each valve by other positive means, such as 

nonintrusive testing, or disassembly and inspection. However, licensees must not use these

methods to verify leak tightness, which requires Category A valve testing. Relief is not required 

to perform testing of each valve in the valve pair.

Basis for Recommendation

The requirements of Subsections ISTC-5223 are contained in the NRC’s endorsement of the 

ASME OM Code 2004 Edition, as published in a regulatory amendment to 10 CFR 50.55a.

Subsections ISTC-5221(a) and/or ISTC-5221(c) allow the use of nonintrusive examination, 

disassembly and inspection, or other positive means, for check valves that have no provision for 

testing individual valves or practical means to demonstrate the closure capability of each valve 

by flow, but not for verifying leak-tightness.

Keep-fill valves are a special case, in that they are redundant valves in a system in which only 

one valve of a series is actually necessary to perform a system’s intended function. Licensees 

have proposed to exclude the upstream valve from the IST program. However, recognizing that 

neither valve can individually demonstrate a closure function, and that the Code alternative 
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allows the valve pair to be operationally tested closed as a unit, the NRC staff previously 

determined that both valves must be included in the IST program and must be operationally 

tested as a pair to prevent reverse flow. The NRC staff specified that upon observing leakage, 

the licensee must disassemble, inspect, and repair or replace both valves (as necessary) before 

return to service.

4.1.2 Exercising Check Valves with Flow and Nonintrusive Techniques

The Code requires both an open and close test for check valves and that licensees exercise 

check valves to the position(s) required to fulfill their safety function(s). To verify the disk 

position of check valves that do not have external disk position indication, the Code allows 

licensees to use indirect evidence (such as changes in system pressure, flow, temperature, or 

level) or other positive means. An acceptable test method must demonstrate by positive means 

that a check valve disk moves to the position necessary to fulfill its safety function. The 

demonstration may not require that the valve be exercised “full-open” to the backstop. The 

full-stroke to the open position may be verified either by passing design flow or by other positive 

means such as nonintrusive techniques. The “other positive means” must be repeatable to 

meet the intent of the Code.

NRC Recommendation

The licensee may use nonintrusive techniques for IST of check valves. Relief is not required 

except as would be necessary for the testing frequency if the test interval extends beyond each

refueling outage as allowed by the OM Code. The licensee may use nonintrusive techniques to 

verify the valve’s capability to open, close, and fully stroke if it is qualified for the application in 

accordance with the plant’s quality assurance program requirements. A qualified nonintrusive 

technique is one that has been successfully and reliably demonstrated for the examination 

method and specific valve application. The licensee may qualify the technique and application 

on its own equipment, subcontract it to a vendor, or rely on the results of the Nuclear Industry 

Check Valve Group (NIC) evaluation of nonintrusive diagnostic techniques for check valves. 

Personnel training and qualification performed by a vendor in accordance with the licensee’s 

quality assurance program may be acceptable; however, the technique must also be qualified 

as described above. Records of techniques and qualification documentation shall be 

maintained in accordance with the licensee’s quality assurance program, and the NRC inspector 

may examine the licensee’s records.

Basis for Recommendation

The NRC previously determined that nonintrusive testing methods appropriate for certain valve

applications are acceptable to verify the capability of the valve to open, close, and fully stroke, 

provided that the licensee properly qualifies the testing methods used for the valve application in 

accordance with the plant’s quality assurance program requirements. These techniques are 

considered “other positive means” in accordance with Subsection ISTC-5221(a)(3). It is the 

licensee’s responsibility to qualify and document the results in accordance with the plant’s 

quality assurance program requirements. Appendix B to 10 CFR Part 50 provides for the quality 

assurance program, which includes, in part, requirements to ensure that nondestructive testing 

is controlled and accomplished by qualified personnel using qualified procedures in accordance 

with applicable codes, standards, specifications, criteria, and other special requirements. The 
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NIC conducted an experimental research and testing program to evaluate available nonintrusive 

technologies to determine their acceptability and reliability for use in check valve testing 

applications. Information on qualification of nonintrusive testing is provided in the summary of 

the NRC’s public workshops on the revision of NRC Inspection Procedure 73756, dated July 18, 

1997 (Actual July 27, 1995). In response to a question about expectations for qualification of a 

nonintrusive test method, the NRC indicated that a qualified nonintrusive test method is a 

technique that has been successfully and reliably demonstrated for the examination method and 

for the specific valve application. Other expectations discussed in NRC IN 2000-21 are 

summarized in the following paragraphs.

Qualification includes establishing a performance baseline when the check valve is known to be

in good operating condition. A check valve’s performance can be assessed against this

baseline trace. One means to determine if a nonintrusive test method or technique will provide 

accurate, reliable, and repeatable results for a specific check valve is to qualify the method prior 

to its use. The qualification process may reveal that certain techniques or methods give 

inconclusive results for a particular application. Acoustic techniques and test methods are 

susceptible to other plant system noise being transmitted and masking or affecting the desired 

sound pattern and results. The NIC suggests the use of more than one technique to verify 

questionable results.

4.1.3 Full Flow Testing of Check Valves

The ASME OM Code requires both an open and close test for check valves and that licensees 

exercise check valves to the position(s) in which they perform their safety function(s). A check 

valve’s full-stroke to the open position may be verified by passing the maximum required 

accident condition flow through the valve. This is considered an acceptable full-stroke. Any 

flow rate less than this is considered a partial-stroke exercise. A valid full-stroke exercise by 

flow requires that the flow through the valve must be known. Knowledge of only the total flow 

through multiple parallel lines does not provide verification of flow rates through an individual 

valve and may not be a valid full-stroke exercise without further analysis.

4.1.3.1 Alternative to Direct Flow Measurement

Flow through a check valve must be known for a valid full-stroke exercise test, but an alternative 

to direct flow rate instrumentation may be acceptable. Any quantitative measure that has 

acceptance criteria that demonstrate the required flow through the check valve may be used to

satisfy the full-stroke requirement. An indirect measure of flow may be acceptable. For

example, a change in tank level over a specified period could be used. In another case, the

acceptance criterion could be based on a change in flow rate of an instrumented line when flow 

is admitted from a non-instrumented line containing the check valve being tested. In any event, 

some form of quantitative criteria should be established to demonstrate full-stroke capability.

4.1.3.2 Flow through Parallel Lines

Knowledge of total flow through multiple parallel lines does not provide indication of flow through 

each individual path. The objective of inservice testing is to evaluate and investigate the 

possibility of degradation of individual components and to take corrective action before a

component fails. Verification of total header flow rate might not identify a problem, developing 
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or occurring, with an individual check valve in one of the parallel flowpaths. With respect to the 

balancing of flow, TS requirements are based on the flow from one loop being lost through a 

postulated break. Consequently, that flowpath is restricted or throttled to minimize significant 

diversion of flow. TS surveillance requirements were not intended to verify individual check 

valve operability. The licensee is expected to justify the use of a test method that does not 

verify full-stroke of individual check valves.

For example, in a safety evaluation dated January 24, 1992, the NRC informed the licensee of

the results of an evaluation of flow through parallel lines and stated that a flow test through

parallel lines without individual flow measurement may not be sufficient to indicate that the 

check valves in the lines are full-stroke exercised. Knowledge of only total flow through multiple 

parallel lines does not provide verification of flow rates through an individual valve and may not 

be a valid full-stroke exercise without further analysis.

4.1.3.3 Accident Condition Flow

The phrase “maximum required accident condition flow” is the largest flow rate for which a

licensee takes credit in a safety analysis for this component in any flow configuration. The

safety analyses are those contained in the plant’s final safety analysis report (FSAR), or

equivalent, but are not limited to the accident and transient analyses.

4.1.3.4 Check Valves Not Required to Be Fully Opened

For check valves that are never required to open fully (i.e., thermal expansion or siphon 

breakers), verification of design (safety) function is testing to confirm the capability of forward 

flow through the system. In addition to verifying its safety function performance, licensees 

should develop quantifiable acceptance criteria for the testing of these components. Verifying 

that a system is full is an acceptable means for verifying that the keep-fill check valves are 

capable of opening to provide flow when necessary.

4.1.3.5 Impractical Full-Flow Testing

Full-flow testing of a check valve may be impractical for certain valves. It may be possible to

qualify other techniques to confirm that a valve is exercised to the position required to perform 

its safety function. To substantiate the acceptability of any alternative technique for meeting the 

ASME Code requirements, licensees must, as a minimum, address and document the following 

items in the IST program. (Any alternative techniques for meeting ASME Code requirements 

must be submitted to the NRC for approval pursuant to 10 CFR 50.55a (a)(3)(i).)

(1) the impracticality of performing a full-flow test

(2) a description of the alternative technique used and a summary of the procedures being

followed

(3) a description of the method and results of the program to qualify the alternative 

technique for meeting the ASME Code

(4) a description of the instrumentation used and the maintenance and calibration of the

instrumentation
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(5) a description of the basis used to verify that the baseline data has been generated when

the valve is known to be in good working order, such as recent inspection and

maintenance of the valve internals [components]

(6) a description of the basis for the acceptance criteria for the alternative testing and a

description of corrective actions to be taken if the acceptance criteria are not met

The NRC position for full-flow testing of check valves allows licensees flexibility in qualifying 

alternatives to full-flow testing. In general, licensees should demonstrate that the alternative 

test is quantifiable and repeatable. The alternative test must meet the intent of the ASME OM 

Code. This qualification of the alternative test should be documented by the licensee and

should be available for review by NRC inspectors. Any alternative techniques for meeting

ASME Code requirements must be submitted to the NRC for approval pursuant to 10 CFR

50.55a(a)(3)(i).

The 1994 Edition of the ASME OM Code and subsequent editions include the use of

nonintrusive testing as other positive means for demonstrating check valve exercising. The

criteria listed in the NRC position for full-flow testing of check valves could be applied to the

nonintrusive techniques.

4.1.4 Disassembly and Inspection Alternative to Flow Testing

Guidance regarding disassembly and inspection of certain check valves is included in 

Subsection ISTC-5221(c) for use if the test methods of Subsections ISTC-5221(a) and 

ISTC-5221(b) are impractical, or sufficient flow cannot be achieved.

NRC Recommendation

Guidance regarding testing check valves by disassembly is as follows:

The sample disassembly and inspection program involves grouping similar valves and 

testing a different valve in each group during each refueling outage. The sampling 

technique requires that each valve in the group be the same design (manufacturer, size, 

model number, and materials of construction) and have the same service conditions 

including valve orientation.

Additionally, at each refueling outage sampling the licensee must verify that the 

disassembled valve is capable of full-stroking and that the internals of the valve are 

structurally sound (no loose or corroded parts). Also, if the disassembly is to verify the 

full-stroke capability of the valve, the disk should be manually exercised. While the valve 

is in a partially disassembled condition the valve internals should be inspected and the 

condition of the moving parts evaluated. This inspection and evaluation should include 

verification that the valve disk is free to move. Following reassembly, a partial flow test 

is expected to be performed.

A different valve of each group is required to be disassembled, inspected, and manually 

full-stroke exercised at each successive refueling outage, until the entire group has been 

tested. If the disassembled valve is not capable of being full-stroke exercised or there is
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binding or failure of valve internals, the remaining valves in that group must also be 

disassembled, inspected, and manually full-stroke exercised during the same outage.

Once this is completed, the sequence of disassembly must be repeated unless the valve 

group is in the Condition Monitoring Program alternative and an extension of the interval 

can be justified.

At least one valve from each group shall be disassembled and examined at each 

refueling outage; all valves in each group shall be disassembled and examined at least 

once every 8 years.

The personnel performing the disassembly/inspection must be qualified to evaluate the

condition of the valve and to assess its continued operability. The licensee is responsible for 

the development and implementation of a program to ensure that IST personnel are

appropriately trained and qualified to perform the valve disassembly/inspections. Licensees 

must implement the provisions of ANSI/ASME N45.2.6, “Qualifications of Inspection, 

Examination, and Testing Personnel for Nuclear Power Plants.” The NRC staff encourages 

those licensees that review the ANSI standard for guidance in developing a program for the 

qualification of IST personnel.

Basis for Recommendation

In ASME OM Code, Subsection ISTC-5221(c) provides for the use of sample disassembly and 

inspection of check valves. Specifically, Subsection ISTC-5221(c) provides requirements for 

sample disassembly and inspection of certain check valves, if the test methods of Subsections

ISTC-5221(a) and ISTC-5221(b) are impractical, or if sufficient flow cannot be achieved or

verified. ISTC-5221(c)(3) provides the requirements that every valves in each group must be 

examined at least once every 8 years. The NRC endorsed these Code requirements in its 

regulatory amendment to 10 CFR 50.55a.

4.1.5 Reverse Flow Testing of Check Valves

The OM Code requires that Category C check valves (those that are self-actuated in response 

to some system characteristic such as pressure or flow direction) that perform a safety function 

in the closed position to prevent reverse flow must be tested in a manner that proves that the

disk travels to the seat on cessation or reversal of flow. In addition, for Category A/C check 

valves (those that have a specified leak rate limit and are self-actuated in response to a system 

characteristic), seat leakage must be limited to a specific maximum amount in the closed 

position for fulfillment of their function. Verification that a Category C valve is in the closed 

position can be achieved through visual observation, by an electrical signal initiated by a 

position-indicating device, by observation of appropriate pressure indication in the system, by 

leak testing, or by other OM Code-defined positive means, i.e. ISTC-5221.

ASME Code-class check valves that perform a safety function in the closed position and are

frequently not reverse flow tested include the following examples:

(a) main feedwater header check valves

(b) pump discharge check valves on parallel pumps
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(c) keep-fill check valves

(d) check valves in steam supply lines to turbine-driven AFW pumps

(e) main steam non-return valves

(f) chemical and volume control system (CVCS) volume control tank outlet check valves

4.1.5.1 Closure Capability of Check Valves that Do Not Have Defined Seat Leakage 

Limits

When performing a test to verify closure capability of a check valve that does not have a defined 

seat leakage limit, the achievement of the necessary system flow rate through the intended 

flowpath might be an adequate demonstration of the closure capability of a check valve. For 

example, when verifying the closure capability of the check valves on the discharge of parallel 

pumps, achievement of the required safety flow rate from one running pump with the idle 

pump’s discharge check valve providing the barrier for recirculation flow would be considered an 

acceptable test configuration. In addition, the licensee should evaluate the consequences of 

reverse flow through the check valve. This evaluation should consider the loss of water from 

that system and connecting systems, the effect that the leakage might have on components and 

piping downstream of the valve, and any increase in radiological exposure resulting from the 

leakage.

A plant’s safety analysis may include a leakage limit for a particular valve, or may only require 

that the valve closes to inhibit gross leakage. When a valve has a safety-related function to

close to prevent diversion of flow between trains of a system, there may be a leakage limit 

based on the total system requirements. The Code does not specifically require these valves to

be Category A. The basis for assigning valves to categories should be available for inspection.

Licensees may refer to NRC Information Notice 91-56, “Potential Radioactive Leakage to Tank 

Vented to Atmosphere” regarding information concerning the categories assigned to valves that 

function to close to prevent leakage above an assumed limit thus preventing a plant from 

exceeding the limits of 10 CFR Part 100.

4.1.5.2 Listed Systems

In reference to the examples of valves that are not frequently backflow tested (listed in Section

4.1.5, above), all of the listed systems do not necessarily apply to each plant. As a minimum, a 

licensee should evaluate the listed systems to determine whether they apply to its facility and 

should make any necessary modifications to its IST program. Examples (d), (e), and (f) are 

specific to PWRs, while Example (a) may apply to both PWRs and BWRs. Example (f) may 

serve an important safety function at some PWRs to separate the non-safety-grade water 

source from the safety-grade source.

4.1.5.3 Permissible Leak Rates

Subsection ISTC-3630(e) requires that the licensee must specify a permissible leak rate for a 

specific valve. If the licensee does not specify leak rates, permissible leak rates are provided in 

Subsection ISTC-3630(e). It should be noted that the OM Code does not provide either criteria 

or guidance concerning the methods licensees should use to establish or specify the 

permissible leak rate of a particular valve. The OM Code recognizes that the leak behavior of a

 
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valve varies according to valve type and size, vendor, service conditions, safety-related 

functions, and other factors, and there is no simple leak rate rule that applies to all valves.

In general, licensees should set the leak rate limits within certain bounds. If the leak limits are

too low, unnecessary valve repairs or adjustments can result. Leak limits that are too high could 

result in failure of the tests required by Appendix J to 10 CFR Part 50, thereby leading to

concerns regarding the leak-tight integrity of the containment. Appropriate permissible leak

rates can only be developed and refined by analyzing and trending the leak rate data for specific 

valves or for similar valves at other plants. Therefore, the NRC staff has not provided specific 

guidance concerning leak rates. Licensees should document their methods for establishing the 

initial permissible leak rates and procedures for verifying compliance with the leak rate limits.

4.1.5.4 Closure Testing of Stop Check Valves

If a stop-check valve does not perform a safety-related function in the closed position, valve

closure is only necessary to ensure a repeatable starting point for opening testing. Valves may

be closed by using a handwheel or a hand switch. (Note:1996 Addenda, and later Editions to

the OM Code requires bidirectional testing.)

If the use of a handwheel or hand switch to close a valve achieves the safety-related function of

the system, exercising the valve by this method meets the ASME Code requirements of

Subsection ISTC-5221. By contrast, if closure of a stop check valve on cessation or reversal of 

flow is required to accomplish a safety-related function, its closure must be verified by reverse 

flow testing or other positive means, such as acoustic monitoring or radiography.

When no other means of verification are possible, licensees may disassemble valves to verify 

valve closure. However, disassembly provides limited information on valve capability to seat on

cessation or reversal of flow. Furthermore, if the method involves extensive disassembly, a

post-reassembly test would be necessary in accordance with Subsection ISTC-3310, because 

disassembly and inspection can increase the probability of human error when the valve is

reassembled. Licensees may investigate the use of nonintrusive testing techniques and may

implement such techniques if they are demonstrated to be effective to assess closure capability, 

degradation, and incipient failure. Infrequent disassembly and inspection of the valves are 

appropriate to assess overall check valve condition, while reverse flow testing and nonintrusive 

testing provide an assessment of continued operational readiness.

If closure testing using flow can only be done at an extended interval, the licensee should close

the valve using a hand wheel or hand switch on a quarterly frequency, in conjunction with the 

flow test on the extended interval, if practical.

4.1.5.5 Other Positive Means of Verification

Subsection ISTC-5221 allows for other positive means of verification. Examples from IST 

programs include verifying that a parallel centrifugal pump does not spin in reverse to verify

closure of a pump discharge check valve, monitoring an upstream pressure indicator, 

monitoring a tank level, measuring the flow rate of a redundant train, or opening an upstream 

vent and drain valve.
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4.1.6 Extension of Test Interval to Refueling Outage for Check Valves Verified Closed 

by Leak Testing

When it is impracticable for the licensee to verify check valve closure during plant operation or

cold shutdown, it is acceptable for the licensee to extend the check valve quarterly exercise test 

(both open and close) to the refueling outage when the closure verification may be performed in

conjunction with the Type C leak rate test conducted in accordance with Option A of Appendix J 

to 10 CFR Part 50. Licensees may also perform the open exercise test during the refueling 

outage or anytime during the fuel cycle interval.

NRC Recommendation

If no other practical means are available, it is acceptable for licensees to extend the quarterly 

closure exercise test to a refueling frequency. In such instances, the licensee must develop a

refueling outage justification describing the impracticality of performing the quarterly closure test 

during plant operation or cold shutdown. The NRC staff has determined that the need to set up

test equipment constitutes adequate justification to defer reverse flow testing of a check valve to 

a refueling outage. By referencing the refueling justification in the IST program document, the 

licensee may perform the closure exercise test during each reactor shutdown for refueling. A 

seat leak test is one method to verify that the obturator has traveled to the seat. All 

requirements of each individual valve category are applicable, although repetition of a common 

testing requirement is not required. Therefore, when the required performance of the Appendix 

J leak rate test coincides with a refueling outage exercise seat leak test, only the Appendix J 

test is required.

The OM Code states that open and close tests need only be performed at an interval “when it is

practicable to perform both tests.” The OM Code also states that licensees are not required to 

perform open and close tests at the same time if they are both performed during the same 

interval. Therefore, since the closure test is extended to the refueling outage by the refueling 

justification, the quarterly open exercise test may also be extended to the refueling outage or 

may be performed anytime during the fuel cycle interval.

Basis for Recommendation

OM Subsection ISTC-5221(a) states that valve obturator movement observations shall be made 

by a direct indicator (e. g., a position-indicating device) or other positive means including seat 

leak testing. Therefore, a seat leak test is one method to verify that the obturator has traveled 

to the seat.

OM Subsection ISTC-3522(c) states that if exercising is not practical during plant operation and 

cold shutdowns, it shall be performed during refueling outages. A refueling outage justification 

shall document the extension of the exercise test to the refueling outage.

OM Subsection ISTC-3522(a) states that open and close exercise tests need only be performed 

at an interval when it is practical to perform both tests. This Code section also states that open 

and close tests are not required to be performed at the same time if they are both performed 

during the same interval.
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4.1.7 Testing and Examination of Check Valves Using Manual Mechanical Exercisers

In ASME OM, Subsection ISTC-5221(b), “Valve Obturator Movement,” in part, requires that if a

manual mechanical exerciser is used to test the check valve, the force(s) or torque(s) required 

to move the obturator to fulfill its safety function(s) shall meet the acceptance criteria specified 

by the owner. (1) Exercise test(s) shall detect a missing obturator, sticking (closed or open), 

binding (throughout obturator movement), and the loss or movement of any weights. (2)

Acceptance criteria shall consider the specific design, application, and historical performance. 

(3) If it is impractical to detect a missing obturator or loss or movement of any weight(s), other 

positive means may be used (e.g., seat leakage tests and visual observations to detect 

obturator loss and the loss or movement of external weight(s), respectively).

NRC Recommendation

In the Code requirements through the OMa-1996 Addenda, Subsection ISTC 4.5.4(b) requires 

that, if a manual mechanical exerciser is used for IST movement of the obturator, the force or 

torque to initiate disk movement shall not vary by more than 50 percent from an established 

reference value. Licensees have continuously experienced difficulty with this IST acceptance 

criterion. The manual mechanical exerciser assembly includes a packing gland to seal the 

hinge pin penetration of the valve body. The hinge pin seal packing introduces conditions that 

produce variations in friction forces over time. These variations make it difficult to establish a 

reference value that would be continually consistent and appropriate for use in IST.

The NRC has received a number of requests for relief from the requirement that force or torque 

to initiate disk movement not vary by more than 50 percent of the reference value. Licensees 

have also requested that the ASME OM Working Group on Check Valves (WGCV) reexamine 

this requirement. The WGCV reexamination resulted in the Code change in OMa-2000, 

Subsection ISTC-5221(b), “Valve Obturator Movement,” which is discussed above. This 

change requires the owner to specify the acceptance criteria within certain Code-defined

expectations. In establishing an acceptance criterion for IST when using a manual mechanical 

exerciser, the owner should consider the interactions, wear and effects of the valve parts on 

friction forces, and the valve preventive maintenance activities.

Basis for Recommendation

Mechanical exercisers are attached to a hinge pin that is fixed to the disk and penetrates the 

valve body. Many of these valves involve swing check valves that manufacturers supplied with

a lever arm and counterweight modification. The counterweight is used to affect the opening or 

closing response of the disk to flow conditions, depending upon the lever arm’s location relative 

to the disk. The counterweight modification involves the use of a packing gland to seal the 

hinge pin penetration of the valve body. The seal packing introduces variations over time with 

regard to the required disk opening force and opening and closing responses of the disk, 

depending upon the type of packing material used, its condition, friction changes, leakage 

control adjustments, and the packing gland tightening procedure. Any wear of the hinge pin and 

bearing interfaces may exacerbate these variations. Disk opening and closing friction forces 

may also change as a result of valve preventive maintenance activities. ASME OM, Subsection 

ISTC-5221(b), “Valve Obturator Movement,” states that if a mechanical exerciser is used to 

exercise the check valve, the force(s) or torque(s) required to move the obturator to fulfill its 

safety function(s) shall meet the acceptance criteria specified by the owner. 
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4.1.8 Check Valve Bidirectional Testing and Condition Monitoring Program

Bidirectional testing ensures that a check valve is adequately tested, regardless of its safety 

function. Such testing also improves the IST capability to detect valve degradation prior to valve

failure. Two significant OM Code changes, in Subsections ISTC 4.5.4(a) and ISTC 4.5.5, 

respectively, were introduced in OMa-1996. Specifically, those changes included (1) a

requirement for bidirectional exercise testing of the disk movement of check valves, and (2) a 

voluntary provision to use the condition monitoring program as an alternative to IST exercise 

testing for certain check valves. This integral two-part improvement to the Code provides 

interrelated requirements. The condition monitoring program allows licensees certain IST 

flexibility in establishing the types of test, examination, and preventive maintenance activities 

and their associated intervals, when justified based on the valve’s performance and operating 

condition. These Code changes were developed so that licensees who elect not to implement 

condition monitoring in their IST program, would be required to bidirectionally test check valves 

as a default set of testing and examination requirements.

OMa-1996, Subsection ISTC 4.5.4(a), “Valve Obturator Movement,” in part, requires that the

necessary obturator movement during exercise testing shall be demonstrated by performing

both an open and close test, and observations shall be made by a direct indicator (e.g., a 

positive-indicating device) or by other positive means (e.g., changes in system pressure, flow, 

level, temperature, seat leakage, testing, or nonintrusive testing and examination).

OMa-1996, Subsection ISTC 4.5.5, Condition Monitoring Program, in part, provides an option 

for the owner to establish a check valve condition monitoring program alternative to the testing 

and examination requirements of Subsections ISTC 4.5.1 - 4.5.4. The purpose of the program 

is to improve valve performance and optimize testing, examination, and preventive maintenance 

activities in order to maintain acceptable check valve performance. The program must be 

implemented in accordance with Appendix II, “Condition Monitoring Program.”

NRC Recommendation

The required testing or examination of the check valve obturator movements to both the open 

and closed positions, as required by Subsection ISTC-3522, is necessary to assess the valve’s 

operational condition, confirm the acceptability of its performance, and detect degradation prior 

to failure. Single-direction flow testing of check valves will not always detect functional 

degradation of the valves.

The NRC staff considers the condition monitoring program approach of ASME OM, Appendix II, 

for check valve IST with the regulatory modifications 10 CFR 50.55a(b)(3)(iv), to be an 

improvement over present Code requirements, and encourages licensees to implement the 

condition monitoring alternative.

Basis for Recommendation

The NRC incorporated OMa-1996 by reference in 10 CFR 50.55a on September 22, 1999. In

OMa-1996, Subsection ISTC 4.5.4(a) included the requirement that the necessary obturator 

movement during exercise testing must be demonstrated by performing both open and closed 

tests. The NRC agrees with the need for a required demonstration of bidirectional exercise 

testing of the movement of the check valve disk. Single-direction flow testing will not always 
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detect degradation of the valve. The classic example of the flawed single-direction testing 

strategy is that the loss of the disk would not be detected during forward flow tests. The 

detached disk could be lying at the bottom of the valve body or another part of the system, and 

could move to block flow or disable another valve or component. The most recent example of 

an undetected detached check valve disk lying at the bottom of the valve body is captured in the

discussion of the event described in IN 2000-21. The NRC considers testing or examination of

the check valve obturator movement to both the open and closed positions necessary to assess 

its condition and confirm acceptable valve performance.

The use of the ASME OM Code, Appendix II, “Condition Monitoring Program,” as incorporated 

by reference in 10 CFR 50.55a (published in the Federal Register on September 22, 1999), 

includes requirements that apply when extending check valve IST intervals, with regard to 

consideration of the plant safety significance, justification by trending of current and historical 

valve condition and performance data, maximum IST interval limits, stepwise IST interval limits, 

and bidirectional testing or examination. These requirements provide the licensee with 

knowledge of the valve’s operating condition, monitor and verify valve performance over 

extended intervals, and provide a process to justify prudent IST interval extensions to reduce 

the burden of unnecessary IST.

The ASME OM Code Committee, Sub-Group on Check Valves (SGCV), has completed 

proposed changes to Subsection ISTC of the ASME OM Code to address the regulatory 

modification issues. These changes were issued as part of OMb-2003. All changes discussed 

above are incorporated in the 2004 Edition of ASME OM Code.

4.1.9 Instrumentation Requirements

Instruments used to verify that check valves fulfill their safety function(s) are not subject to the

same range and accuracy requirements as instrumentation used for pump-related IST. 

However, OM-1998 added Subsection ISTC-3800, “Instrumentation,” to provide specific 

requirements for instrumentation that is used in the testing and examination of valves. 

Specifically, in OMa-1998, Subsection ISTC-3800 requires that instrumentation, including both 

measuring and test equipment and permanent plant instrumentation, used for valve testing and

examination activities, shall (1) be properly controlled, calibrated, and adjusted in accordance 

with the owner s quality assurance program, and (2) have the accuracy, range, and repeatability 

characteristics necessary to verify compliance with the requirements of this subsection. In

addition, instrumentation accuracy shall be considered when establishing valve test acceptance 

requirements.

NRC Recommendation

In OMb-2000 thru OM-2004 with 2005 and 2006 addendas, ISTC-3800, “Instrumentation,”

specifies requirements for instrumentation that is used in performing IST for check valve testing 

and examination activities. Instrumentation that is used in IST must be controlled and 

calibrated, and must have the accuracy, range, and repeatability necessary to verify compliance 

with the requirements. Accuracy and repeatability of the instrumentation are important 

considerations in the IST of safety-related check valves. IST should be performed in a way that 

permits the results to be compared for indications of valve degradation trends. When 

instrumentation that is used in valve testing and examination is not properly controlled and 

calibrated, any valve degradation indications may be masked, thereby diminishing the 

usefulness of the valve test results.
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Basis for Recommendation

The instrumentation requirements and quality assurance activities specified in Subsection ISTC3800, are needed to properly verify compliance with Code requirements and to detect and trend 

any operational degradation for assurance that the check valves will perform satisfactorily until 

the next IST. Appendix B to 10 CFR Part 50 provides for the quality assurance program and 

includes requirements for IST of safety-related components, test control, and control of 

measuring and test equipment.

4.1.10 Skid-Mounted Valves and Component Subassemblies

The exercising or examination of each check valve that is contained in (or part of) a 

skid-mounted package or major component is not always practical, particularly if the valve is 

enveloped within the structure of the package or component. Although the valve’s performance 

may support a safety function, the practicality of exercising or examining each valve separately, 

as required by the OM Code, was not addressed prior to OMa-1996. Subsection ISTC-1200(c) 

of OMa-1996 exempts skid-mounted valves from the individual testing requirement and allows

the valves to be tested as part of the overall package or major component. Specifically, 

Subsection ISTC-1200, states, in part, “Skid-mounted valves and component subassemblies 

are excluded from this Subsection, provided they are tested as part of the major component and 

are determined by the Owner to be adequately tested.”

NRC Recommendation

Testing major components that include check valves that are an integral portion of the major 

component and that support the major component s performance of its safety function, requires

that the owner must determine that the check valves are adequately tested. The NRC staff 

expects that, as part of the owner s testing and determination responsibility, the check valves 

will be identified in the IST plan, along with an explanation of how the testing of the major 

component adequately tests the valves. The owner should review the safety significance of the 

identified valves to ensure that the IST is adequate to demonstrate their continued operability. 

The documentation that provides assurance of the continued operability of the valves through 

the performed tests should be available at the plant site.

Basis for Recommendation

GDC 1 in Appendix A to 10 CFR Part 50, in part, requires that components important to safety 

must be tested to quality standards commensurate with the importance of the safety functions 

performed.

4.1.11 Check Valves in New Reactors

RG 1.206 provides guidance for COL licensees to develop IST programs for check valves. New 

nuclear power plants should provide the means for bi-directional testing of check valves within 

the scope of the IST program. As part of its review of design certification and COL applications, 

the NRC staff is evaluating whether new nuclear power plants will be designed to allow 

bi-directional testing of check valves. The staff will conduct inspections of the development and 

implementation of the IST program (including bi-directional testing of check valves) during 

construction and operation of the new nuclear power plants. 
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4.2 Power-Operated Valves

Power-operated valves (POVs) are equipped with actuators that use motive force to change the

position of the valve obturator. The types of actuators may include, for example, motor 

actuators, pneumatic actuators, hydraulic actuators, solenoid actuators, and pyrotechnic 

actuators. In addition, in the ASME OM Code, Subsection ISTC defines a power-operated relief 

valve (PORV) as a POV that can perform a pressure-relieving function and is remotely actuated 

by either a signal from a pressure-sensing device or a control switch. Certain valves, such as 

main steam isolation valves and valves that have a fail-safe function, may actuate open (or 

closed) on spring force. The ASME Code includes provisions for exercising, stroke-time testing, 

leak testing, and position-verification testing of POVs in the IST program. In the following 

sections, the NRC staff provides guidance concerning the implementation of specific Code 

provisions and associated regulatory requirements.

4.2.1 Stroke-Time Testing for Power-Operated Valves

In the ASME OM Code, Subsection ISTC-5113 specifies that active POVs shall have their 

stroke times measured when exercised in accordance with the nominal 3-month schedule 

specified in Subsection ISTC-3500. Stroke-time testing may indicate degradation in the 

performance of POVs. NRC requirements and guidance for supplementing the ASME Code 

provisions for stroke-time testing are discussed later in this document.

The ASME Code includes provisions for establishing reference values and limiting values for

POV stroke times. Subsection ISTC-3300 states that reference values shall be determined from 

the results of preservice testing or inservice testing. Subsections ISTC-5113, 5121, 5131, 5141, 

and 5151 also state that limiting values for stroke time for various types of POVs shall be 

specified by the owner. In addition, Subsections ISTC-5114, 5122, 5132, 5142, and 5152 

include a set of acceptance criteria for the reference value of the stroke time for POVs, and 

specify various corrective actions to be taken if those criteria are not satisfied. If the limiting 

value of stroke time is exceeded, Subsections ISTC-5115, 5123, 5133, 5143, and 5153 state 

that the POV shall be immediately declared inoperable.

The Code does not specify provisions for establishing the limiting value for stroke times, and it

does not identify the relationship that should exist between those limits and the reference values 

for stroke time or any limits identified in the plant Technical Specifications (TS) or safety 

analysis. 

NRC Recommendation

The limiting value of full-stroke time should be based on the reference (or average) stroke time 

of a POV when it is known to be in good condition and operating properly. The limiting value 

should be a reasonable deviation from this reference stroke time, based on the size and type of

the valve and power actuator. The deviation should not be so restrictive that it results in a POV 

being declared inoperable as a result of reasonable stroke time variations. However, the

deviation used to establish the limiting value should be such that corrective action would be 

taken to provide assurance that the POV would remain capable of performing its safety function.

The limiting value for stroke time of a POV should be that point at which the licensee seriously 

questions continued operability. It is expected to be a value that is determined to be reasonable
A 4-17

for the individual POV based on its characteristics and past performance, but not to exceed any

safety analysis requirements. The value should not be based solely on the system 

requirements or values specified in safety analyses for system performance. When the 

identified limiting value is exceeded, the licensee shall declare the component inoperable and

shall enter any applicable TS limiting condition for operation (LCO). After declaring the valve 

inoperable, the licensee should perform an analysis to identify the cause of the problem with the 

POV. If this analysis clearly demonstrates that the POV remains capable of performing its

safety function, the analysis might constitute the corrective action required by the Code. The

analysis must be documented.

Licensees should establish reference values that reflect the stroke time of the specific POV 

when in good condition and operating under applicable conditions. A licensee may establish 

additional sets of reference values as discussed in Subsection ISTC-3320, such as reference 

values that reflect test conditions of fluid pressure or flow in the system.

Licensees may use a quantitative multiplier on a reference time as a means of establishing a

limiting value for stroke time. The licensee should document the justification for its selection of 

reference values for the stroke time of each POV, and should have this justification available at

the plant site for review by NRC personnel.

Basis for Recommendation

The purpose of the limiting value of full-stroke time for a POV is to establish a value for taking 

corrective action on a degraded POV before it reaches the point where there is a high likelihood 

of failure to perform its safety function. While the TSs provide assurance that important plant 

systems are capable of performing their safety functions in a timely manner during selected 

plant transient accidents and anticipated operational occurrences, the provisions of the ASME 

OM Code are intended to ensure the continued operability of particular plant components. The 

distinct bases for these two documents (i.e., TSs and ASME Code) lead to criteria that may 

differ significantly. Nonetheless, the TSs and ASME Code are both needed to provide 

confidence that the nuclear power plant can be operated safely. Therefore, licensees must 

follow the more restrictive criteria of the two documents, even though this might result in a 

component or system being declared inoperable. For example, if the TSs or safety analysis 

limit for a POV is less than the IST value established using the above guidelines, the TSs or 

safety analysis limit should be used as the limiting value of full-stroke time. When the TSs or 

safety analysis limit for a POV is greater than the IST value established using the above 

guidelines, the limiting value of full-stroke time should be based on the above guidelines instead 

of the TSs or safety analysis limit. The TSs and safety analysis limits are useful for analyzing 

data when a POV has indicated degraded performance and been declared inoperable. In 

accordance with Subsections ISTC-5115, 5123, 5133, 5143, and 5153 the data may be 

analyzed to verify that the new stroke time represents acceptable POV operation.

4.2.2 Stroke-Time Measurements for Rapid-Acting Valves

In the ASME OM Code, Subsections ISTC-5114, 5122, 5132, 5142, and 5152 allow licensees to

establish the limiting stroke time of 2 seconds for POVs that stroke in less than 2 seconds. The 

Code also eliminates the acceptance criterion related to the reference value for stroke-time for

those POVs. However, new technologies and new applications of existing technologies enable 

licensees to time the strokes of rapid-acting valves with accuracy measured in milliseconds. 
A 4-18

Using new technology, licensees could establish an appropriate limiting stroke time based on a

multiple of a reference value to ensure that corrective actions are taken if degrading conditions 

are identified.

The traditional method of stroke timing POVs was to use stopwatches to measure the stroke 

time from initiation of the signal at the handswitch to the change in position-indicating lights 

(switch to light). The traditional method includes signal processing time from the switch to the

valve actuator. Monitoring stroke times for valves that stroke in milliseconds using diagnostic 

equipment that measures only actual valve travel is acceptable to indicate degrading trends; 

however, the method does not indicate increases that could occur in the signal to the valve, 

which may be important in meeting safety analysis limits for certain valves. Typically, the valves 

that would benefit from this monitoring are rapid-acting valves. The traditional method would 

have a set limit of 2 seconds, which masks any signal processing time unless a gross change 

occurs. If measuring the stroke times locally needs to be supplemented by a periodic test to 

include the signal processing times, a periodic 2-second limit test could be performed to 

augment the IST. The Code does not specify a particular method, so there would be no conflict 

in using more than one method.

NRC Recommendation

The NRC staff recommends that licensees should determine whether continued reliance on the

2-second limiting stroke time criterion in the ASME Code is appropriate when the actual stroke 

time can be measured in milliseconds.

Basis for Recommendation

The 2-second limiting stroke time for rapid-acting valves was based on measurement of stroke 

times using a stopwatch. Updated technology may improve the monitoring of the condition of 

these POVs or verify that a valve operates within a safety analysis limit that is less than 

2 seconds.

4.2.3 Stroke Time for Solenoid-Operated Valves

The NRC is often asked to approve relief from the ASME Code provisions to allow licensees not

to measure the stroke times of enclosed solenoid-operated valves (SOVs) that do not have 

position indication. If the licensee cannot time the stroke of an SOV by the conventional method 

using position indication, the licensee needs to propose a method to time the stroke of the valve 

or otherwise monitor the POV for degrading conditions to provide adequate assurance of its

operational readiness. If the frequency provisions of the Code are met, the licensee does not 

need to request relief to use methods such as acoustics or diagnostic systems for stroke timing. 

If the licensee intends to apply a method to monitor for degradation other than by measuring 

stroke time, NRC authorization of the alternative is required pursuant to 10 CFR 50.55a(a)(3). 

For example, an enhanced maintenance program or periodic replacement may be acceptable 

when testing methods cannot be used effectively.

NRC Recommendation

The NRC staff recommends that licensees should use advanced diagnostic techniques to obtain 

stroke-time measurements in accordance with the frequency provisions of the Code, and should 
A 4-19

also use those advanced techniques or maintenance programs to monitor the degradation of 

SOV performance. In addition, the staff recommends that the technique should evaluate actual 

disk movement and not only movement of the pilot valve or valve stem.

Basis for Recommendation

In NUREG-1275, Vol. 6, “Operating Experience Feedback Report: Solenoid-Operated Valve 

Problems,” the NRC described common-mode SOV problems that could significantly reduce 

plant safety. Several methods are available to measure stroke time or monitor the condition of

SOVs using parameters such as the acoustic effects of disk movement, electric resistance, and 

the temperature of the coil. These advanced diagnostic techniques provide more precise 

means of monitoring SOV performance.

4.2.4 Supplement to the POVs Stroke-Time Test Provisions of the ASME OM Code

Operational experience and valve testing programs have revealed weaknesses in the ability of

stroke-time testing to assess the operational readiness of POVs to perform their safety 

functions. In response to those weaknesses, ASME, the NRC, and various industry groups 

have taken action to provide improved confidence in the capability of POVs to perform their 

safety functions under design-basis conditions.

With respect to motor-operated valves (MOVs), the NRC’s regulations in 10 CFR 50.55a require

that licensees who are implementing the ASME OM Code (beginning with the 1995 Edition) 

must supplement the provisions for MOV stroke-time testing specified in the Code with a 

program to ensure that the MOVs continue to be capable of performing their design-basis safety 

functions. In a Federal Register notice (64 FR 51370) dated September 22, 1999, the NRC 

discussed the implementation of MOV programs in response to GL 89-10, “Safety-Related 

Motor-Operated Valve Testing and Surveillance,” (June 1989), and GL 96-05, “Periodic 

Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves,” (September 

1996), as a means of satisfying the requirement to supplement MOV stroke-time testing.

The NRC established Generic Safety Issue (GSI) 158, “Performance of Safety-Related 

Power-Operated Valves Under Design-Basis Conditions,” to evaluate whether additional 

regulatory actions were necessary to address performance issues for air-operated valves 

(AOVs), hydraulic-operated valve (HOVs), and solenoid-operated valves (SOVs). In Regulatory 

Issue Summary (RIS) 2000-03, “Resolution of Generic Safety Issue 158, ‘Performance of Safety 

Related Power-Operated Valves Under Design-Basis Conditions’,” dated March 15, 2000, the

NRC closed GSI-158 on the basis that current regulations provide adequate requirements to

ensure verification of the design-basis capability of POVs, and no new regulatory requirements 

are needed. In RIS 2000-03, the staff provided attributes for an effective POV testing program 

that incorporates lessons learned from MOV research and testing programs. 

NRC Recommendation

The NRC staff recommends that licensees should apply their MOV programs established and

implemented in response to GL 89-10 and GL 96-05 to supplement the provisions in the ASME 

OM Code for MOV stroke-time testing in accordance with the requirements in 10 CFR 50.55a. 

The staff also recommends that licensees consider information provided in RIS 2000-03, as well 

.
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as lessons learned from their own MOV programs, to improve confidence in the capability of 

other POVs to perform their safety functions.

Basis for Recommendation

In the 1980s and 1990s, operating experience at nuclear power plants revealed that 

weaknesses in the ability of stroke-time testing to assess the operational readiness of POVs 

allowed performance deficiencies to remain undetected for an extended period of time. In

10 CFR 50.55a, the NRC requires that licensees whose code of record is the 1995 Edition (or a

later edition or addenda) of the ASME OM Code must supplement their stroke-time testing of

MOVs with programs to ensure that the MOVs are capable of performing their design-basis 

safety functions. As discussed in RIS 2000-03, the NRC staff determined that current 

requirements and guidance indicate the need for licensees to have confidence in the capability 

of all safety-related POVs to perform their design-basis functions. In RIS 200-03, the staff 

discusses industry activities to improve POV performance.

In GL 89-10, the NRC asked licensees to ensure that MOVs in safety-related systems have the

capability to perform their intended functions by reviewing MOV design bases, verifying MOV 

switch settings initially and periodically, testing MOVs under design-basis conditions where 

practical, improving evaluations of MOV failures and necessary corrective actions, and trending 

MOV problems. The NRC subsequently issued GL 96-05 to request that licensees establish a 

program, or ensure the effectiveness of their current program, to verify, on a periodic basis, that 

safety-related MOVs continue to have the capability to perform their safety functions within the 

current licensing basis of the facility. The NRC staff reviewed licensees’ activities in response to 

GL 89-10 and GL 96-05 through plant-specific inspections and reviews of submitted information.

Licensees have completed their GL 89-10 programs for the operational nuclear power plants. In

response to GL 96-05, the owners’ groups of nuclear power plant licensees established the

Joint Owners’ Group (JOG) Program on MOV Periodic Verification as an industry-wide effort to

evaluate potential degradation of MOV operating requirements. Most licensees committed to 

implement the JOG MOV program as part of their response to GL 96-05. In a safety evaluation 

dated September 2006 and its supplement dated September 2008, the NRC staff accepted the 

application of the JOG Program on MOV Periodic Verification as described in the SE and its 

supplement. As discussed in the SE, the JOG program addresses the operating requirements 

of valves and, therefore, licensees are responsible for justifying the output capability of MOVs 

within the scope of the JOG program. Further, licensees are responsible for justifying the 

long-term periodic verification programs for MOVs or their applications that are outside the 

scope of the JOG program. The staff considers the MOV design-basis capability verification 

performed by licensees in response to GL 89-10 and the MOV periodic verification program 

being conducted by licensees under GL 96-05, including any appropriate modifications in 

response to the JOG topical report and the NRC safety evaluation, to satisfy the requirement in 

10 CFR 50.55a to supplement the MOV stroke-time test provisions of the ASME OM Code.

Although the NRC has not established new regulatory requirements to address performance 

issues for POVs other than MOVs, current NRC regulations and documents contain 

requirements and guidance intended to provide assurance that safety-related POVs are capable 

of performing their safety-related functions. For example, the regulations in Appendices A and

B to 10 CFR Part 50 require that licensees must provide confidence that safety-related
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equipment (including POVs) is capable of performing its safety functions under design-basis 

conditions. Further, the regulations in 10 CFR 50.65 require that licensees must monitor the 

performance of structures, systems, and components (SSCs) in a manner sufficient to provide 

reasonable assurance that such SSCs are capable of fulfilling their intended functions. With 

respect to AOV air systems, the NRC staff issued GL 88-14, “Instrument Air Supply System 

Problems Affecting Safety-Related Equipment,” to ask licensees to verify (by test) that AOVs will 

perform as expected in accordance with all design-basis events. The staff provided the results 

of studies of POV issues in NUREG-1275, “Operating Experience Feedback Report,” Volumes

2, 6, and 13; NUREG/CR-6644, “Generic Issue 158: Performance of Safety-Related 

Power-Operated Valves Under Operating Conditions,” and NUREG/CR-6654, “A Study of 

Air-Operated Valves in U.S. Nuclear Power Plants.” In RIS 2000-03, the staff provided a list of 

attributes of a successful POV design capability and long-term periodic verification program 

based on lessons learned from staff reviews of valve programs and plant visits. The staff also

prepared several information notices (INs) to alert licensees to IST issues related to POV 

performance. Specifically, these included IN 86-50, “Inadequate Testing To Detect Failures of 

Safety-Related Pneumatic Components or Systems ‘; IN 85-84, “Inadequate Inservice Testing 

of Main Steam Isolation Valves”; and IN 96-48, “Motor-Operated Valve Performance Issues,”

which described lessons learned from MOV programs that are applicable to other POVs.

ASME has initiated efforts to improve the Code provisions for assessing the operational 

readiness of POVs in IST programs at nuclear power plants. For example, ASME has 

incorporated ASME OM Code Cases OMN-1, “Alternative Rules for Preservice and Inservice 

Testing of Certain Electric Motor-Operated Valve Assemblies in LWR Power Plants, “ and 

OMN-11, “Risk Informed Testing of Motor-Operated Valves,” as Appendix III in the 2009 Edition 

to the ASME OM to replace quarterly MOV stroke-time testing with periodic exercising and 

diagnostic testing (Note: The details related to the ASME OM-2009 for information only). ASME 

also developed OM Code Case OMN-12 that provides guidance for an alternative to quarterly 

stroke-time testing for air- and hydraulic-operated valves. OMN-12 includes risk-informed 

provisions to allow licensees to obtain more precise performance data for use in assessing the 

operational readiness of POVs that are determined to have higher safety significance, while 

allowing licensees to obtain less precise data for POVs of lower safety significance. 

In addition to the NRC and ASME, the nuclear industry is taking action to address POV

performance issues. As discussed above, the industry developed the JOG Program on MOV

Periodic Verification to share resources among licensees and to establish an improved

response to the MOV issues to be addressed under GL 96-05. In addition, a Joint Owners’

Group on Air-Operated Valves (JOG AOV) has established a voluntary program to improve 

confidence in the capability of safety-related AOVs to perform their design-basis functions. In 

RIS 2000-03, the NRC staff noted that it provided comments on the JOG AOV program in a 

letter to the NEI dated October 8, 1999. The staff also stated that it would continue to monitor 

licensees’ activities to ensure that POVs are capable of performing their safety-related functions 

under design-basis conditions. If the industry does not adequately address POV functionality 

under design-basis conditions, the NRC staff indicated in RIS 2000-03 that additional regulatory 

action may be necessary.

4.2.5 Alternatives to POVs Stroke-Time Testing

Given the weakness in stroke-time testing for assessing the operational readiness of POVs, 

ASME developed alternatives to the Code provisions for stroke-time testing of POVs.
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As an alternative to MOV stroke-time testing, ASME developed Code Case OMN-1, “Alternative

Rules for Preservice and Inservice Testing of Certain Electric Motor-Operated Valve Assemblies 

in LWR Power Plants,” which provides periodic exercising and diagnostic testing for use in 

assessing the operational readiness of MOVs. In Code Case OMN-11, “Risk-Informed Testing 

of Motor-Operated Valves,” ASME provides additional guidance for use with Code Case OMN-1 

to emphasize the testing provisions for MOVs in the IST program that are determined to have 

high safety significance, while allowing less precise testing for MOVs that are determined to 

have lower safety significance. ASME has incorporated these code cases into Appendix III to 

the 2009 Edition of the ASME OM Code to replace quarterly MOV stroke-time testing with 

periodic exercising and diagnostic testing. (Note: The details related to the ASME OM-2009 are 

for information only). 

With respect to AOVs and HOVs, ASME developed Code Case OMN-12, “Alternate 

Requirements for Inservice Testing Using Risk Insights for Pneumatically and Hydraulically 

Operated Valve Assemblies in Light-Water Reactor Power Plants (OM Code 1998, Subsection

ISTC), to provide an alternative to the Code stroke-time testing provisions that incorporates risk 

insights to focus on AOVs and HOVs in the IST program that are determined to have the 

highest safety significance, while allowing less emphasis on AOVs and HOVs that have lower 

safety significance.

NRC Recommendation

The NRC regulations in 10 CFR 50.55a(b)(3) require nuclear power plant licensees 

implementing the ASME OM Code incorporated by reference in 10 CFR 50.55a to supplement 

the quarterly MOV stroke-time testing provisions in the ASME OM Code with a program to 

periodically demonstrate the design-basis capability of safety-related MOVs. The NRC staff 

considers that the MOV programs developed in response to GLs 89-10 and 96-05 together will 

satisfy the regulatory requirement to supplement the MOV stroke-time provisions in the ASME 

OM Code. The staff also considers the provisions in ASME OM Code Case OMN-1 as 

accepted in RG 1.192 to satisfy the regulatory requirement to supplement the quarterly MOV 

stroke-time provisions. ASME has replaced the quarterly MOV stroke-time testing provisions in 

Appendix III to the 2009 Edition of the ASME OM Code that incorporates the provisions in 

ASME OM Code Cases OMN-1 and OMN-11. (Note: The details related to the ASME OM-2009 

are for information only). The NRC staff also considers the alternative approach to quarterly 

stroke-time testing of AOVs and HOVs in ASME OM Code Case OMN-12 as accepted in 

RG 1.192 to provide an acceptable alternative to the ASME OM Code stroke-time testing 

provisions for those POVs. Therefore, the NRC staff recommends that licensees implement 

ASME Code Cases OMN-1, OMN-11, and OMN-12, as accepted by the NRC (with certain 

conditions) in RG 1.192, as alternatives to the stroke-time testing provisions in the ASME Code 

for applicable POVs. When the 2009 Edition to the ASME OM Code is incorporated by 

reference in the NRC regulations, licensees will be responsible for modifying their MOV testing 

programs in accordance with the 10-year update requirements for IST programs. (Note: The 

details related to the ASME OM-2009 are for information only). 

Basis for Recommendation

As part of the ASME Code Committee process, industry experts have developed ASME Code

Cases to address weaknesses in the ability of stroke-time testing to assess the operational 

readiness of POVs in IST programs at nuclear power plants. The ASME Code Cases
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incorporate risk insights to emphasize IST provisions for POVs that are determined to have the

highest safety significance. The NRC has reviewed and accepted several of these Code Cases 

with certain conditions. For example, RG 1.192 allows licensees with an applicable code of 

record to implement ASME Code Case OMN-1 (in accordance with the provisions in the

regulatory guide) as an alternative to the Code provisions for MOV stroke-time testing, without 

submitting a request for relief from their code of record. In RG 1.192, the staff also accepts 

(with certain conditions) the use of the risk-informed provisions in Code Case OMN-11 by 

applicable licensees, in conjunction with Code Case OMN-1. RG 1.192 also allows licensees 

with an applicable code of record to implement Code Case OMN-12 for AOVs and HOVs (with 

certain conditions) in lieu of the Code provisions for stroke-time testing, without the need to 

submit a relief request. Licensees with a code of record that is not applicable to the acceptance 

of these Code Cases may submit a request to apply those Code Cases, consistent with the 

indicated conditions, as an alternative to the ASME OM Code that provides an acceptable level

of quality and safety. ASME has incorporated ASME OM Code Cases OMN-1 and 11 as 

Appendix III to the 2009 Edition of the ASME OM Code. Therefore, the NRC regulations will 

require licensees to modify their MOV testing programs to apply the provisions of Appendix III to 

the ASME OM Code in accordance with the 10-year update requirements for IST programs. 

(Note: The details related to the ASME OM-2009 are for information only). 

4.2.6 Main Steam Isolation Valves

In IN 85-84, “Inadequate Inservice Testing of Main Steam Isolation Valves,” the NRC staff 

described an inadequacy in the IST of MSIVs Specifically, the staff stated that two different 

licensees were testing their MSIVs using the nonsafety-related instrument air to achieve 

closure. Fail-safe IST of MSIVs as required by Subsection ISTC-3560 necessitates the removal 

of the instrument air supply and electric power. Concerns related to MSIVs are described in

IN 94-08, “Potential for Surveillance Testing To Fail To Detect an Inoperable Main Steam 

Isolation Valve,” and IN 940-44, “Main Steam Isolation Valve Failure To Close on Demand 

Because of Inadequate Maintenance and Testing.”

NRC Recommendation

The staff recommends that licensees should review their inservice and fail-safe testing to

ensure compliance with Code requirements.

Basis for Recommendation

The practice of performing IST of components that are relied on to mitigate the consequences of 

accidents using sources of power that were not considered in the safety analyses is inconsistent 

with the objective of periodic IST for fail-safe testing. In IN 85-84, the NRC staff alerted 

licensees that, with low or no steam flow, the MSIV might not close unless instrument air is

available to power the actuator.

In its Service Information Letter 477, General Electric (GE) described a related concern for

boiling-water reactors (BWRs) in which excessive tightening of gland flanges in the MSIV can 

prevent the valve from closing in response to spring force alone. During a postulated 

design-basis accident in which a recirculation line breaks with the MSIVs open, containment 

pressure may increase significantly, and may exert an opening force on the valve actuators 

inside containment. Under such circumstances, the MSIV springs alone will not close the MSIV
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unless the spring force can overcome the combination of the opening force caused by 

containment pressure and the resistive force caused by stem packing friction. GE

recommended a review of packing chamber maintenance practices, “springs-only” full-stroke 

closing tests, a force balance in which containment pressure is considered, a leak-tightness test 

of the MSIV actuator and accumulator, and a modification of the applicable licensing-basis 

documents. GE noted that this would necessitate measurement of the actual valve stem travel 

because the final 10 percent of stem travel coincides with the weakest spring force. GE stated 

that, by monitoring position switches alone, a utility could not determine that the valve is fully 

closed because the switches monitor the valve only when it is 90-percent open or 90-percent 

closed. One BWR licensee identified that the MSIVs would not pass local leak rate testing after 

closing on spring force only.

4.2.7 Verification of Remote Position Indication for Valves by Methods Other Than 

Direct Observation

ASME OM Code, Subsection ISTC-3700, “Position Verification Testing,” states

Valves with remote position indicator shall be observed locally at least once every 

2 years to verify that valve operation is accurately indicated. Where practicable, this local 

observation should be supplemented by other indications such as the use of flowmeters 

or other suitable instrumentation to verify obturator position. These observations need 

not be concurrent. Where local observation is not possible, other indications shall be 

used for verification of valve operation.

The requirement that valves with remote position indicators must be observed at least once 

every 2 years to verify that valve position is accurately indicated has been specified in the 

ASME OM Code and the previous ASME/ANSI OM Part 10 for many years. The additional 

guidance for supplementing the local observation was not present in older versions of 

OM Part 10, such as in the 1983 versions. 

Many valves have no provision for verifying the position by direct observation. To verify the 

position by observation, licensees can disassemble the valve, which could introduce additional 

valve failure mechanisms. Other methods (such as nonintrusive techniques, causing the flow to 

begin or cease, leak testing, and pressure testing) can yield a positive indication of position.

NRC Recommendation

ISTC-3530, “Valve Obturator Movement,” allows obturator movement to be determined by 

indicating lights in the control room when exercising the valve to meet the quarterly stroke-time 

testing requirement of the ASME OM Code. The valve position verification testing required by 

ISTC-3700 provides confirmation on a 2 year frequency that the indicating lights reflect actual 

valve operation. ISTC-3700 allows flexibility to licensees in verifying that operation of valves 

with remote position indicators is accurately indicated. The extent of verification necessary for 

valve operation to satisfy ISTC-3700 will depend on the type of valve, the sophistication of the 

diagnostic equipment used in testing the valve, possible failure modes of the valve, and the 

operating history of the valve and similar valve types. After such consideration, the licensee will 

be responsible for developing and implementing a method to verify that valve operation is 

accurately indicated as required by ISTC-3700.
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Basis for Recommendation

In Section 4.2.5, “Verification of Remote Position Indication for Valves by Methods Other Than 

Direct Observation,” of NUREG 1482 (April 1995), the NRC staff noted that the Code requires 

that valves with remote position indicators be observed at least once every 2 years to verify that 

valve position is accurately indicated. The staff stated that if remote valve position cannot be 

verified by local observation at the valve, an acceptable approach is for the licensee to observe 

operational parameters such as leakage, pressure, and flow that give positive indication of the 

valve’s actual position. The staff indicated its interpretation of the Code requirement by stating 

that for certain types of valves that can be observed locally, but for which valve stem travel does 

not assure the stem is attached to the disk, the local observation must be supplemented by 

observing an operating parameter as required in the Code. In the basis discussion, the staff 

stated that accurate position indication for safety-related valves is important for reactor 

operation during all plant conditions. Therefore, the staff noted that the Code requires 

verification of the accuracy of the remote position indication for all valves in the IST program 

with remote position indication. The staff indicated that many positive ways are available to 

verify the indication that a valve is open or closed. For example, the staff referenced leak-rate 

testing, in-line flow rate instrumentation, and system and differential pressures for indication of 

valve position.

In Section 4.2.7 of Revision 1 to NUREG-1482 (January 2005), the NRC staff discussed its 

interpretation of the ASME OM Code for position indication verification. The staff noted that 

ISTC-3700 requires verification of the accuracy of the remote position indication for all valves in 

the IST program with remote position indication. The staff indicated that if licensees cannot 

verify remote valve position by local observation at the valve, an acceptable approach is for the 

licensee to observe operational parameters (such as leakage, pressure, and flow) that give a 

positive indication of the valve’s actual position. In Revision 1, the staff modified the discussion 

on supplementing local observation of valves for which stem travel does not assure that the 

stem is attached to the disk with observation of operating parameters (i.e., changing “must” to 

“should”) because Revision 1 to the NUREG removed the preapproval of this approach and 

required Commission approval pursuant to 10 CFR 50.55a(f)(4)(iv). 

The Code requires licensees to verify the accuracy of the remote position indication for all 

valves in the IST program that have remote position indication. Subsection ISTC-3700 states 

that where local observation is not possible, licensees shall use other indications to verify 

operation. Nuclear power plant operating experience has revealed that reliance on indicating 

lights and stem travel are not sufficient to satisfy the requirement in ISTC 3700 to verify that 

valve operation is accurately indicated for those valves where the integrity of the stem-to-disk

connection cannot be assured. Therefore, licensees are responsible for developing a method to 

verify that valve operation is accurately indicated to satisfy ISTC-3700.

The ASME OM Code Subsection ISTC-3700 requires that valves with remote position indicators 

must be observed at least once every 2 years to verify that valve position is accurately 

indicated. Many valves (such as sealed solenoid valves and valves with enclosed stems) have

no provision for verifying the position by direct observation. To verify the position by

observation, licensees can disassemble the valve, which could introduce additional valve failure 

mechanisms. Other methods (such as nonintrusive techniques, causing the flow to begin or 

cease, leak testing, and pressure testing) can yield a positive indication of position.
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NRC Recommendation

If licensees cannot verify remote valve position by local observation at the valve, an acceptable

approach is for the licensee to observe operational parameters (such as leakage, pressure, and 

flow) that give a positive indication of the valve s actual position(s). This is consistent with 

Subsection ISTC-3700. The staff determined that the use of this portion of the OM Code would 

not require relief if the licensee implements all requirements of Subsection ISTC-3700. No other 

related requirements apply. However, Commission approval is still required pursuant to 

10 CFR 50.55a(f)(4)(iv).

For certain types of valves that can be observed locally, but for which valve stem travel does not

ensure that the stem is attached to the disk, the local observation should be supplemented by

observing an operating parameter as required by Subsections ISTC-3700 and ISTC-3520.

Basis for Recommendation

Accurate position indication for safety-related valves is important for reactor operation under all

plant conditions. The Code requires licensees to verify the accuracy of the remote position 

indication for all valves in the IST program that have remote position indication. Subsection 

ISTC-3700 states that where local observation is not possible, licensees shall use other 

indications to verify operation. Such indications are also useful to ensure that a valve disk is 

connected to the stem. Licensees are not able to verify the accuracy of remote position 

indication by local observation of many valves, such as those with enclosed stems or sealed 

solenoid valves, where the valves do not have position indicators, such as pointers, on the valve 

actuators.

Many positive means are available to verify the indication that a valve is open or closed. For

example, leak-rate testing may yield positive indication that the disk is in the closed position. In

addition, an in-line flow rate instrument can indicate system flow or flow stoppage. System

pressures or differential pressure across a valve seat may also give a positive indication of

actual valve position.

4.2.8 Requirements for Verifying Position Indication of Passive Valves

The Code does not restrict the verification of position indication to only active valves. The 

ASME OM Table ISTC 3500-1 indicates that the licensee must also locally verify the position

indication for Category B passive valves. As discussed in Section 4.2.7 of this NUREG, the 

licensee is responsible for developing and implementing a method to verify that valve operation 

is accurately indicated as required by ISTC 3700. The extent of verification necessary for valve 

operation to satisfy ISTC 3700 will depend on the type of valve, the sophistication of the 

diagnostic equipment used in testing the valve, possible failure modes of the valve, and the 

operating history of the valve and similar valve types. 
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The Code does not require licensees to verify the indication at the remote panels. However, 

verification at remote panels is a good practice and provides confidence in the remote 

indication.

4.2.9 Control Valves with a Safety Function

In general, control valves that respond to system conditions would be exempt from IST as

discussed in Subsection ISTC-1200. However, some control valves also perform safety or failsafe functions (e.g., fail open, fail closed, fail as-is), and such valves must be tested in

accordance with the requirements for IST. The staff has received many requests for relief from

stroke-time measurement requirements, based on the impracticality of performing the

measurement by the conventional method using position indication lights. Typically, the control 

valves do not have position indication, and testing can only be performed by bypassing control 

signals. To allow stroke timing by bypassing the control signals of those control valves that 

have position indication lights, the licensee may have to drain systems, which might make it 

impractical to test at the Code-defined frequency.

NRC Recommendation

Control valves that perform a safety or fail-safe function must be tested in accordance with the

Code provisions for IST to monitor the valves for degrading conditions. The NRC staff 

recommends that licensees should apply ASME Code Case OMN-8, as accepted in RG 1.192,

if concerns exist regarding IST of control valves with fail-safe functions. Code Case OMN-8 

states that stroke-time testing need not be performed for POVs when the only safety-related 

function of those valves is to fail safe. Any abnormality or erratic action experienced during

valve exercising should be recorded in the test record and an evaluation should be performed.

4.2.10 Pressurizer Power-Operated Relief Valve Inservice Testing

Power-operated relief valves (PORVs) were often not purchased as safety-related valves, and

the function of these valves to provide pressure control for plant transients was not considered 

safety-related. The valves were not designed to serve as overpressure protection devices 

during power operations, as required by ASME Section III, but many have been used as lowtemperature overpressure protection valves.

NRC Recommendation

The staff recommends that licensees should consider previous NRC guidance that the PORVs 

should be included in the IST program as Category B valves and should be tested to the OM 

Code. Recognizing that the PORVs have shown a high likelihood of sticking open and are not 

needed for overpressure protection during power operation, the provisions in Subsections ISTC3500 and ISTC-5100 for exercising quarterly during power operation are not practical and,

therefore, exercising may be performed during cold shutdown conditions. Subsection ISTC3310 requires licensees to perform testing after maintenance or repair. Test methods must 

confirm that the PORV has been reassembled correctly and is capable of performing its design 
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function. There have been instances (see IN 96-02, “Inoperability of PORVs Masked by 

Downstream Indications During Testing”) where improper evaluation of testing failed to identify 

the incorrect reassembly of a PORV.

Previously approved NRC guidance included in GL 90-06 (see below) indicates that, because

the PORVs function during reactor startup and shutdown to protect the reactor vessel and

coolant system from low-temperature overpressurization conditions, they should be exercised 

before system conditions warrant vessel protection, and should also be exercised after the

operational readiness of the block valves is ensured, by exercising and stroke-timing according 

to the following test schedule:

• Perform full-stroke exercising during each cold shutdown or, as a minimum, once each 

refueling cycle.

• Perform stroke timing during each cold shutdown, or as a minimum, once each refueling 

cycle.

• Perform fail-safe testing during each cold shutdown, or as a minimum, once each 

refueling cycle.

• Include the PORV block valves in the IST program, and test them quarterly to ensure 

protection against a small-break LOCA in the event that a PORV fails open.

• If the plant frequently enters cold shutdown mode, testing of the PORVs is not required 

more often than once every 3 months.

Basis for Recommendation

The NRC’s guidance on the IST requirements for PORVs is included in GL 90-06, “Resolution of

Generic Issue 70: Power-Operated Relief Valve and Block Valve Reliability,” and Generic Safety 

Issue 94: Additional Low-Temperature Overpressure Protection for Light-Water Reactors

Pursuant to 10 CFR 50.54(f).” In IN 89-32, “Surveillance Testing of Low-Temperature 

Overpressure-Protection Systems,” the NRC discussed the stroke time assumptions made in 

plants’ safety analyses for these PORVs, and the IST performed for these valves. Stroke times 

of the valves were unacceptable or were not measured in the direction required for lowtemperature overpressure systems to prevent exceeding the limits in Appendix G to 10 CFR

Part 50. Compliance with the guidance in GL 90-06 has been coordinated between the plants 

and the NRC project managers for each plant on a case-by-case basis.

4.2.11 POVs in New Reactors

RG 1.206 provides guidance for COL licensees to develop IST programs for POVs including, for 

example, MOVs, air-operated valves, hydraulic-operated valves, and solenoid-operated valves. 

The NRC staff reviews the consideration of diagnostic testing of POVs as part of the review of 

IST program descriptions provided in support of COL applications. The NRC regulations in 10 

CFR 50.55a require licensees of new nuclear power plants to update their IST programs to the 

most recent edition and addenda of the ASME OM Code incorporated by reference in the NRC 

regulations 12 months before fuel loading. The staff will conduct inspections of the 

development and implementation of the IST program (including POV diagnostic testing) during 

construction and operation of the new nuclear power plants.
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With respect to MOVs, the NRC regulations in 10 CFR 50.55a(b)(3)(ii) will require licensees of 

new reactors to supplement the quarterly MOV stroketime testing provisions in the ASME OM 

Code with a program to periodically demonstrate the design-basis capability of safety-related 

MOVs. COL licensees may apply the JOG Program on MOV Periodic Verification as accepted 

in an NRC safety evaluation dated September 2006 and its supplement dated September 2008 

as part of satisfying 10 CFR 50.55a(b)(3)(ii). As discussed in the SE, the JOG program 

addresses the operating requirements of valves and, therefore, licensees are responsible for 

justifying the output capability of MOVs within the scope of the JOG program. Further, 

licensees are responsible for justifying the long-term periodic verification programs for MOVs or 

their applications that are outside the scope of the JOG program. If the 2009 Edition (or later 

edition or addenda) of the ASME OM Code has been incorporated by reference in the 

regulations, the new reactor licensee will be required to implement Appendix III to the ASME 

OM Code that replaces quarterly MOV stroke-time testing with periodic exercising and 

diagnostic testing. (Note: The details related to the ASME OM-2009 are for information only). 

With respect to POVs other than MOVs, the NRC staff evaluates whether the IST program 

description includes the ASME OM Code provisions for quarterly stroke-time testing. The staff 

also evaluates whether IST program description for POVs other than MOVs incorporates the 

lessons learned from MOV testing and research programs. For example, RIS 2000-03 includes 

attributes of a successful IST program for POVs that incorporates lessons learned from MOV 

operating and testing experience. ASME is developing improved IST provisions for POVs other 

than MOVs for incorporation into the ASME OM Code. See Section 4.4.8 for IST guidance 

regarding pyrotechnic actuated valves.

4.3 Safety and Relief Valves

4.3.1 Scope

Subsection ISTA-1100 defines the scope of the valves subject to IST to include pressure-relief 

devices that protect systems (or portions of systems) that perform a required function in shutting 

down the reactor to the safe shutdown condition, maintaining the safe shutdown condition, or 

mitigating the consequences of an accident that results from overpressure. Pressure-relief 

valves, which are installed in systems to protect against overpressure, may not, of themselves, 

appear to perform a specific function to shut down the reactor, maintain it in a safe shutdown 

condition, or mitigate the consequences of an accident. (Automatic depressurization valves in 

BWRs are an example of relief valves that perform both an overpressure protection function and 

a function to depressurize the primary system when opened on an automatic signal or by an 

operator.) However, they may be required to be included in the IST program and tested 

according to the schedules stipulated in Subsection ISTC and Appendix I of the ASME OM 

Code. Specifically, Subsection ISTC of the ASME OM Code clarifies that its requirements apply 

only to pressure-relief devices required for overpressure protection. The testing of these 

devices is to be included in 120-month updated IST programs. 
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Testing of “thermal relief valves” has been the subject of much discussion over the past several

years. Contributing to some confusion, in many original system designs, so-called thermal relief 

valves were installed to protect isolated segments of piping that could be pressurized as a result 

of heating from some source, but were widely viewed as having no safety-related function in 

mitigating the consequences of accidents or ensuring any other system safety function. Generic 

Letter 96-06, “Assurance of Equipment Operability and Containment Integrity During DesignBasis Accident Conditions,” emphasized the importance of protecting certain isolated segments 

of piping from excessive thermally-induced pressurization, especially where containment 

integrity could be affected. In recent years, ASME has made changes to Appendix I to the OM 

Code to include specific requirements to periodically test or replace thermal relief valves.

The requirement to test safety and relief valves that provide overpressure protection is based on

the requirements of Section III of the ASME Code, as well as the USA Standard Code for 

Pressure Piping (USAS B31.1) and the USA Standard Code for Nuclear Power Piping (USAS

B31.7). If the results of an overpressure protection “re-analysis” for a particular system indicate 

that a relief valve is not necessary, it may be removed from the scope of the IST program.

4.3.2 Method of Testing Safety and Relief Valves

4.3.2.1 BWR Safety/Relief Valve Stroke Testing

In recent years, the NRC staff has received numerous requests for relief and/or TSs changes 

related to the stroke testing requirements for BWR dual-function main steam safety/relief valves 

(S/RVs). Both Appendix I to the OM Code and the plant-specific TSs require stroke testing of

S/RVs after they are reinstalled following maintenance activities. Several licensees have

determined that in situ testing of the S/RVs can contribute to undesirable seat leakage of the 

valves during subsequent plant operation and have received approval to perform stroke testing 

at a laboratory facility coupled with in situ tests and other verifications of actuation systems as 

an alternative to the testing required by the OM Code and TSs.

4.3.2.2 PWR Main Steam Safety Valve Set Pressure Testing

To reduce the need to remove valves from their installed position and the time required to

perform set pressure testing, many PWR licensees perform testing of main steam safety valves 

(MSSVs) using in situ testing equipment with operating steam pressure. One advantage of this 

method is that actual environmental and fluid temperature conditions are used, in lieu of 

duplicating them in a test laboratory. However, this method has introduced inaccuracies 

because the set pressure is determined by a combination of the measured system operating 

pressure and the applied assisting force provided by the testing device. This assisting force is 

applied by pneumatic pressure on a piston or diaphragm and is converted to an equivalent 

additional amount of system pressure by dividing the force by the valve disk area against which 

the system pressure acts. Inaccuracies in the value of the disk area have caused some 

inaccuracies in the set pressure determination, as discussed in IN 94-56, “Inaccuracy of Safety 

Valve Set Pressure Determination Using Assist Devices.”

4.3.3 Jack-and-Lap Process

In IN 91-74, “Changes in Pressurizer Safety Valve Setpoints Before Installation,” the NRC 

stated that the setpoint changes in Dresser pressurizer safety valves could result, in part, from
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changes made during a jack-and-lap procedure that is performed after setpoint testing and 

before installation to reduce seat leakage. This procedure may have lacked adequate controls.

Many licensees avoid performing setpoint testing after jack-and-lap maintenance because this 

testing could lead to leakage. Subsection ISTC and Appendix I of the OM Code require that 

after repairing a valve or performing maintenance that could affect the valve s performance, the 

licensee must demonstrate that the performance parameters are acceptable by testing the valve

before returning it to service. The licensee must test pressure relief devices as required by 

Subsection ISTC and Appendix I following replacement, repair, and maintenance.

The staff recommends that, if a licensee chooses to use the jack-and-lap process and not reverify the set pressure of the valves, the licensee must determine whether the maintenance 

activity is of an extent that a setpoint test is required after the valve is reassembled and 

reinstalled. If the jack-and-lap process is controlled so that the setpoint will not be affected, the 

licensee may not need to perform a test. Action in accordance with this recommendation 

necessitates determination of the effect of this activity and evaluation within the quality controls 

and quality assurance for the process. Controls include limits on the amount of material that is 

removed, the controls to ensure that the settings and adjustments of the valve parts that affect 

the setpoint are not changed, and the requirements in the maintenance procedure to address 

any unusual conditions that occur during the maintenance activity. The licensee may also 

consider industry experience to determine whether changes in the methods of performing this 

activity are necessary as plants accumulate more data. Because the NRC staff cannot make 

this determination by evaluating a relief request, relief is neither appropriate nor available for

this activity.

4.3.4 Maintenance and Inspection of Safety and Relief Valves in Addition to OM Code 

Requirements

Licensees should note that not all maintenance and inspection that may be needed to ensure 

continued functional capability of safety, relief, and pilot valves is necessarily performed as a 

result of inservice testing required by the OM Code. In a recent case involving some BWR 

S/RVs, additional periodic maintenance and inspection of certain internal parts were necessary 

to check for excessive wear and eventual binding of the main disks. This was discovered on

valves that had successfully passed required inservice tests and is discussed further in IN 2003-

01, “Failure of a Boiling-Water Reactor Main Steam Safety/Relief Valve.”

4.3.5 Scheduling of Safety and Relief Valve Testing

Appendix I to the OM Code requires that licensees must test a minimum size sample of valves 

within a valve group within a specified period. A penalty is also applied, in that additional valves 

must be tested when any of the sample fails to meet the necessary acceptance criteria. In

determining the minimum acceptable sample size, fractions of valve numbers resulting from

calculating the number of valves to be tested are to be rounded to the next higher whole 

number. 

4.4 Miscellaneous Valves

The following issues and NRC recommendations apply to miscellaneous types of valves.
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.4.4.1 Post-Accident Sampling System Valves

NUREG-0737, Clarification of TMI Action Plan Requirements, Section II.B.3, details the

requirements and capabilities of post-accident sampling systems (PASSs) for sampling both the 

reactor coolant and the containment atmosphere. The PASS consists of pumps and valves that 

perform these and possibly other functions. The PASS also includes valves that perform a 

containment isolation function

NRC Recommendation

The IST program applies to any PASS valves within the scope of 10 CFR 50.55a and the ASME

OM Code. Such valves in the PASS that perform a containment isolation function must be 

included in the IST program as Category A or A/C and must be tested to Code requirements 

except where relief has been granted.

The remaining valves in the PASS would typically be tested as required by the TSs or other 

documents and need not be included in the IST program. However, the staff recommends that 

if the licensee elects to include these valves in the IST program, a note should be included that

the testing is beyond the scope of 10 CFR 50.55a.

In many cases, a licensee’s TSs have been amended to eliminate the requirements to have and

maintain a PASS. If a PASS valve is eliminated from the TSs but still performs a function within 

the scope of 10 CFR 50.55a and the ASME OM Code, the valve should remain in the IST 

program.

4.4.2 Post-Maintenance Testing After Stem Packing Adjustments and Backseating of 

Valves to Prevent Packing Leakage

Subsection ISTC-3310, “Effects of Valve Repair, replacement, or Maintenance on reference 

Values,” requires that, when a valve or its control system has been replaced, repaired, or has 

undergone maintenance that could affect the valve’s performance, a new reference value shall 

be determined or the previous values reconfirmed by an inservice test before the time it is 

returned to service or immediately if not removed from service. 

Examples of maintenance are provided in a footnote of ISTC-3310 and include: adjustment of 

stem packing, limit switches, or control system valves, removal of the bonnet, stem assembly, 

actuator, obturator, or control system components.

Backseating a valve may also affect its performance (e.g., vause damage to the valve or bind it 

into its back seat)

It may be necessary to adjust the stem packing during power operations in order to stop stem

packing leaks on valves that must remain in position for operations to continue. Examples

include MSIVs and main feedwater isolation valves. If the leakage does not pose a personnel 

safety hazard, licensees may adjust the packing without removing the valves from service. 

Alternatively, backseating a valve may stop packing leakage without the need to take the valve 

out of service. Licensees should exercise caution when performing such maintenance, as 

improper backseating or adjustment of valve stem packing could adversely affect the valve’s 

functional capability.
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NRC Recommendation

The staff has determined that it is acceptable for licensees to perform an engineering evaluation 

of the impact of adjusting valve stem packing or backseating a valve to demonstrate that the

performance parameters are within acceptable limits if a stroke test cannot be performed under 

current plant conditions. If it is necessary to adjust the stem packing or backseat a valve to stop 

packing leakage and if a required stroke test or leak rate test is not practical in the current plant 

mode, the licensee must, at a minimum, justify by analysis that (1) the packing adjustment is 

within manufacturer-specified torque limits for the existing packing configuration, (2) the 

backseating does not deform the valve stem, and (3) the performance parameters of the valve 

are not adversely affected. In addition, the licensee must perform a confirmatory test at the first 

available opportunity when plant conditions allow testing. Packing adjustments beyond the 

manufacturer’s limits may not be performed without (1) an engineering analysis showing that the 

performance parameters of the valve are not adversely affected, and (2) input from the 

manufacturer, unless tests can be performed after adjustments.

Examples of such valves are MSIVs and main feedwater isolation valves, which must remain 

open to continue power operations. The licensee must evaluate any data available from 

previous testing with the packing torqued to the specified limit, and must verify that the valve 

was leak tight and previously stroked within acceptable limits with the packing adjusted to the

higher value, or from previous instances of backseating a valve.

Relief is not appropriate because this action is in accordance with the Code requirements if the

licensee can demonstrate that the performance parameters will not be adversely affected.

In implementing this guidance, licensees must perform a partial-stroke test, if practical, to obtain

further assurance that the valve stem is free to move. At the first opportunity when the plant 

enters an operating mode in which testing is practical, the licensee must test all valves that have 

had packing adjustments or been backseated without post-maintenance testing. The 

maintenance procedure used to adjust the packing must include the limits, and any changes to 

the torque limits are subject to a 10 CFR 50.59 review. Licensees should avoid adjusting 

redundant valves without performing post-maintenance testing. Backseating procedures should 

include precautions to prevent stem deformation.

To implement this guidance, a licensee must evaluate valves individually, unless it has 

established a valve packing program in which designated limits, justified by test data, allow

adjustments that do not affect performance parameters. Specific or general relief is not

appropriate for this activity. If the licensee cannot demonstrate that the packing adjustment 

does not adversely affect performance parameters, the Code requirements must be met for

post-maintenance testing. Therefore, the licensee must consider this issue for each valve

individually.
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Basis for Recommendation

The NRC would not require a licensee to shut down a plant to perform IST unless the licensee 

has no alternative to ensure that the operational readiness of components is maintained or

unless a safety issue exists. The IST requirements do not prohibit or discourage a licensee 

from making limited adjustments to packing to stop a leak that may be adversely affecting the

valve or surrounding components. Therefore, the licensee can perform an analysis of the

packing adjustment and, upon demonstrating that the adjustment does not adversely affect the

stroke time (or leakage rate) such that it would not exceed its limiting value, can make the

adjustment without a post-maintenance stroke time measurement (or leakage test). 

Confirmatory testing must be performed at the first available opportunity when plant conditions 

allow testing. This guidance applies only to valves that need adjustment during power operation 

and cannot be fully stroked in the plant operating mode. The guidance does not apply merely

as a convenience to the licensee and does not supersede any related guidance associated with 

GL 89-10.

NRC IN 87-40, “Back Seating Valves Routinely to Prevent Packing Leakage,” gives information 

related to backseating valves. Both Westinghouse and General Electric had issued guidance 

on performing backseating to minimize deformation to valve stems. Backseating is not listed as

an example of a maintenance activity in ASME OM Subsection ISTC-3310. The licensee would 

have to assess the effect of backseating on valve operation and determine whether postmaintenance testing is required. Test results for MOV programs to address GL 89-10 and GL 

96-05 may indicate whether backseating of a particular valve affects its stroke time. Any 

information would need to be included and documented in an evaluation, and the assessments 

would have to be valve-specific.
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4.4.3 Manual Valves

The staff has received questions about the requirements for including manual valves in the IST

program. The Code includes manual valves that meet the scope requirements of 10 CFR

50.55a. To comply with the OM Code, manual valves must be exercised in accordance with 

applicable requirements of Subsection ISTC if the licensee’s safety analysis credits the manual

valve to perform a specific function in shutting down the reactor to a safe shutdown condition, 

maintaining the safe shutdown condition, or mitigating the consequences of an accident. 

Manual valves that perform only a pressure boundary safety function are omitted from the scope 

of the IST program. If the manual valve is included in actions in emergency operating 

procedures, but is not credited in the safety analysis, it does not fall within the scope of the IST 

program; however, such a valve may be periodically tested at an appropriate frequency to 

ensure that it can function satisfactorily. Applicable tests could include exercising, leak testing, 

and position indication verification, at the frequency specified in the Code. Passive manual 

valves that have position indication could be subject to position indication verification.

Manual valves that perform only a pressure boundary safety function are omitted from the scope 

of the IST program. If the manual valve is included in actions in emergency operating 

procedures, but is not credited in the safety analysis, it does not fall within the scope of the IST 

program; however, such a valve may be periodically tested at an appropriate frequency to 

ensure that it can function satisfactorily. 

4.4.3.1 Manual Valve Exercise Interval

On August 3, 2001, the NRC published a proposed rule in Federal Register notice (66 FR

40626) to amend 10 CFR Part 50 to endorse the ASME OM Code, 1998 Edition with the 1999 

and 2000 Addenda. The final rule was published in Federal Register (67 FR 60520) on 

September 26, 2002. As part of that amendment, 10 CFR 50.55a(b)(3)(vi) included a 

modification to the ASME OM Code 1998. That modification specified the manual valve 

exercise interval to as 2 years, rather than 5 years as specified in the 1999 and 2000 Addenda 

to the ASME OM Code. The 1998 Edition and earlier versions of the ASME Code specified an 

exercise interval of 3 months for manual valves within the scope of the Code. The 1999 

Addenda to the ASME OM Code revised Subsection ISTC- 3540 to extend the exercise 

frequency for manual valves to 5 years; however, the NRC staff did not agree that there is 

sufficient justification to extend the exercise interval for manual valves to 5 years. The staff’s 

review of licensees’ IST programs indicated that manual valves are exercised every 3 months 

except in instances where it is impractical to operate valves during unit operation. In such 

instances, the valves are exercised when the unit is in a cold shutdown or refueling outage 

condition, and the exercise frequency cannot exceed 2 years. Therefore, a 2-year interval is 

justified for exercising manual valves because the available manual valve exercise data 

supports a 2-year interval. Nonetheless, licensees are not prohibited from exercising manual 

valves more frequently than every 2 years. In the 2006 Addenda to the ASME OM Code 

revised ISTC-3540 to change the exercise frequency for manual valves to 2 years. 
A 4-36

4.4.4 Pressure Isolation Valves

Pressure isolation valves (PIVs) are defined as two normally closed valves in series that isolate

the reactor coolant system (RCS) from an attached low-pressure system. PIVs are located at 

all RCS/low-pressure system interfaces. As such, PIVs are located within the reactor coolant 

pressure boundary (RCPB), which is defined in 10 CFR 50.2.

“Event V” PIVs are defined as two check valves in series at an RCS/low-pressure system 

interface, which may result in a LOCA that bypasses containment if they fail. The “Event V”

PIVs comprise a subset of PIVs. “Event V” refers to the scenario described for this event in the 

“Reactor Safety Study” (WASH-1400).

On April 20, 1981, the NRC issued Orders to 32 PWRs and 2 BWRs, which required the 

specified licensees to conduct leak rate testing of their PIVs, based on plant-specific NRCsupplied lists of PIVs, and required the licensees to modify their TSs accordingly. These Orders 

are known as the “Event V Orders,” and the valves listed therein are the “Event V” PIVs.

Currently, plants are required to operate within the Improved Standard Technical Specifications 

(STS) which do not contain a listing of PIVs or Event V PIVs. Therefore, the staff recommends 

that licensees should include a listing of PIVs (including Event V PIVs) in their 10-year IST 

programs to document IST testing requirements for each PIV. Licensees should also review 

their testing procedures to ensure that the PIVs are individually leak rate tested. (This position 

supersedes Position 4 of GL 89-04, because the improved STS no longer contain PIV listings.)

4.4.4.1 PIV Discussion in Generic Letter 87-06

GL 87-06 supersedes Position 4 of GL 89-04, because the STS do not contain PIV listings. The 

staff used licensees’ responses to GL 87-06 as input for the resolution of Generic Issue 105, 

“Interfacing Systems LOCAs at Light-Water Reactors,” which was investigated by the NRC’s 

Office of Nuclear Regulatory Research. The results of studies of interfacing system LOCAs are 

provided in NUREG/CR-5124, “Interfacing Systems LOCA: Boiling-Water Reactors,” and 

NUREG/CR-5102, “Interfacing Systems LOCA: Pressurized-Water Reactors.” Generic Issue 

105, which included the issue discussed in GL 87-06, was closed by memorandum from E. 

Beckjord to J. Taylor, “Technical Resolution of Generic Issue 105 (GI-105), ‘Interfacing Systems 

Loss-of-Coolant Accident (ISLOCA) in LWRs’,” dated June 3, 1993. Pressure isolation valves 

need to be included in and tested by the IST programs if they are not included as part of a 

licensee’s technical specifications.

4.4.4.2 Leak Rate Testing of PIVs

The leak rate testing specified in a plant’s TSs must meet the intent of Subsection ISTC-3600. 

A licensee should ensure that each PIV is individually leak tested (or that the measured leakage 

is adjusted) in accordance with the differential pressure requirements of the OM Code. If the 

TSs are not sufficiently detailed to ensure individual valve leak testing, the licensee is

responsible to ensure that the test procedures are themselves adequate for individual valve leak 

testing. .
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NRC Recommendation

A licensee may consider the leakage testing performed to meet TS requirements to also meet 

IST requirements if the intent of the OM Code is met (e.g., leakage limits are established, 

corrective actions are taken as required, and valves are individually leak tested). However, a

licensee must ensure that the test differential pressure specified in the TSs, if applicable, is

equivalent to the function maximum pressure differential, or that the measured leakage is

adjusted to the function maximum pressure differential in accordance with the formula in 

Subsection ISTC-3630 of the OM Code

Basis for Recommendation

Increasing pressure usually improves the seating of a valve. The Code allows that when leak 

testing those types of valves in which the service pressure will tend to diminish the overall 

leakage channel opening, as by pressing the disk into or onto the seat with greater force, the

test differential pressure may be lower than the function maximum differential pressure. The

resulting leakage is to be adjusted according to the following formula from the OM Code:

where

L = leakage

dP = differential pressure

While the NRC staff has accepted other aspects of the TSs, the licensee must ensure that any

testing requirements that are not specifically detailed in the TSs are, nonetheless, imposed on 

the pressure isolation valves to comply with the OM Code leakage testing requirements. 

Generally, the same test will be used to meet both TS and IST requirements. The major 

difference between TSs and IST requirements are related to the acceptance criteria specified in 

some TSs between a nominal leakage limit and the upper leakage limit. (If allowed by TS, the 

upper leakage limit is considered acceptable as the acceptance criteria for IST.)

If the list of PIVs is removed from the TSs, the leakage testing must be described in detail in the

SAR or must be identified as in accordance with the requirements of the ASME OM Code.

4.4.5 Containment Isolation Valves That Have Other Leak-Tight Safety Functions

Valves that function as containment isolation valves may have additional safety functions (i.e.,

other than isolation), such as pressure isolation, train separation, or preventing diversion of flow. 

The leakage testing for Appendix J may not adequately test these additional functions based on 

the pressure or fluid medium. For such valves, the requirements of both Appendix J and 

Subsection ISTC-3600 apply.
A 4-38

4.4.6 Testing Individual Scram Valves for Control Rods in Boiling-Water Reactors

BWRs are equipped with bottom-entry hydraulically driven control rod drive mechanisms with

high-pressure water providing the hydraulic power. Each control rod is operated by a hydraulic 

control unit (HCU), which consists of valves and an accumulator. The HCU is supplied charging 

and cooling water from the control rod drive pumps, and the control rod operating cylinder 

exhausts to the scram discharge volume. Various valves in the control rod drive system 

perform an active function in scramming the control rods to rapidly shut down the reactor.

The NRC has determined that those ASME Code-Class valves that must change position to

provide the scram function should be included in the IST program and should be tested in

accordance with the requirements of Section ISTC except where relief has been granted in a

safety evaluation report. Bi-directional exercise testing of check valves is required by the 1996 

Addenda to the ASME Code (and later editions and addenda).

The control rod drive system valves that perform an active safety function in scramming the

reactor are the scram discharge volume vent and drain valves, scram inlet and outlet valves, 

scram discharge header check valves, charging water header check valves, and cooling water 

header check valves. With the exception of the scram discharge volume vent and drain valves, 

exercising the other valves quarterly during power operations could result in rapid insertion of

one or more control rods. If practical, licensees should test control rod drive system valves at

the Code-specified frequency. However, for those control rod drive system valves for which 

testing could result in rapid insertion of one or more control rods, the rod scram test frequency 

identified in the facility s TSs may be used as the valve testing frequency to minimize rapid

reactivity transients and wear of the control rod drive mechanisms. This alternative test 

frequency should be clearly stated and documented in the IST program document.

Industry experience has shown that normal control rod motion may verify the cooling water 

header check valve moving to its safety function position. This can be demonstrated because

rod motion may not occur if this check valve were to fail in the open position. If this test method 

is used at the Code required frequency, the licensee should clearly explain in the IST program 

document that this is how these valves are being verified to close quarterly.

Closure verification of the charging water header check valves requires that the control rod drive 

pumps must be stopped to depressurize the charging water header. This test should not be

performed during power operation because stopping the pumps results in a loss of cooling 

water to all control rod drive mechanisms, and seal damage could result. Additionally, this test 

cannot be performed during each cold shutdown because the control rod drive pumps supply

seal water to the reactor recirculation pumps, and one of the recirculation pumps is usually kept 

running. Therefore, the HCU accumulator pressure decay test, as identified in the facility’s TSs 

may be used as the charging water header check valve alternative testing frequency for the 

reasons stated above. If this test is not addressed in the licensee’s TSs, this closure verification 

should be performed at least during each refueling outage, and this alternative test frequency 

should be specifically addressed in the IST program document.
A 4-39

The scram inlet and outlet valves are power-operated valves that full-stroke in milliseconds and

are not equipped with indications for both positions; therefore, it may be impractical to measure 

their full-stroke time as required by the Code. Verifying that the associated control rod meets 

the scram insertion time limits defined in the plant’s TSs can be an acceptable alternative 

method of detecting degradation of these valves. Also, it may be impractical and unnecessary 

to trend the stroke times of these valves because they are indirectly stroke timed, and no 

meaningful correlation may be drawn between the scram time and valve stroke time. 

Furthermore, conservative limits are placed on the control rod scram insertion times. If the 

above test is used to verify the operability of scram inlet and outlet valves, it should be 

specifically documented in the IST program document.

4.4.7 Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests

In the ASME OM Code, Subsection ISTC-3522 requires licensees to exercise Category C 

valves every 3 months. ISTC-3620 also requires licensees to seat leak test Category A valves 

(containment isolation valves) in accordance with Appendix J to 10 CFR Part 50. Specifically,

Option B of Appendix J allows a variable seat leak testing frequency, based on component 

performance, and allows test intervals for valves with acceptable performance to be extended to 

once every three refueling outages. Therefore, for Category A/C valves, the Code requires two 

independent tests, including an exercising test and a seat leakage rate test.

The Code recognizes that when more than one distinguishing category characteristic applies, all

requirements for each of the individual categories apply, although duplication or repetition of

common testing requirements is not necessary. Therefore, a seat leak rate test is one

acceptable method to verify the closure portion of an exercise test.

The Appendix II Check Valve Condition Monitoring Program allows an alternative to the 

exercising testing requirements in the OMa-1996 Addenda. The OMa-1996 Addenda included 

two significant changes to IST of check valves to (1) correct certain anomalies in the way check

valves are currently being exercised, and (2) codify a process for monitoring the valve’s 

operating condition and performance. This integral two-part improvement to the Code provides

interrelated requirements. ASME modified Subsection ISTC 4.5.2, “Exercising Requirements,”

and Subsection ISTC 4.5.4, “Valve Obturator Movement,” to require bidirectional testing to 

improve on the detection of valve degradation and failure. The related change to Subsection 

ISTC 4.5.5, Condition Monitoring Program, allowed the use of a codified condition monitoring 

process as an alternative to the exercising and testing requirements of Subsections ISTC 4.5.1 -

4.5.4. The similar requirements are incorporated in the ASME OM Code, Subsections ISTC3520, “Exercising Requirements,” and ISTC-3530, “Valve Obturator Movements,” to continue 

bidirectional check valve testing. The condition monitoring process, defined in Appendix II, 

“Check Valve Condition Monitoring Program,” gives licensees certain IST flexibility in 

establishing the types of test, examination, and preventive maintenance activities and their 

associated intervals, when justified based on the valve s performance and operating condition.
A 4-40

NRC Recommendation

The use of the alternative Appendix II Condition Monitoring Program, with the regulatory 

modifications, provides the licensee with knowledge of the valve’s operating condition, informed

and verified expectations of the valve s performance over extended intervals, and a process to 

justify prudent IST interval extensions to reduce the burden of unnecessary IST. Therefore, the 

staff recommends that licensees should implement the condition monitoring program to justify 

extending the exercise test interval to the leak test frequencies specified in Option B of 

Appendix J.

Basis for Recommendation

The use of the Appendix-II, “Condition Monitoring Program,” as incorporated by reference in 10

CFR 50.55a, provides licensees with knowledge of the valve’s operating condition, monitors and 

verifies valve performance over extended intervals, and provides a process to justify prudent 

IST interval extensions to reduce the burden of unnecessary IST.

4.4.8 Pyrotechnic-Actuated Valves in New Reactors

Some designs for new nuclear power plants include pyrotechnic-actuated (squib) valves that 

have more safety significance than squib valves in currently operating nuclear power plants. In 

addition, squib valves for new reactors might have different designs and be much larger than

squib valves used in current plants. Paragraph ISTC-5260 in the ASME OM Code specifies that 

at least 20% of the charges in explosively activated valves shall be fired and replaced at least 

once every 2 years. If a charge fails to fire, the ASME OM Code specifies that all charges with 

the same batch number shall be removed, discarded, and replaced with charges from a different 

batch. The NRC staff considers the provisions for IST surveillance of squib valves in the ASME 

OM Code to be insufficient for the design and application of squib valves in some new reactors. 

At this time, reactor vendors for new nuclear power plants have not completed the design and 

qualification of squib valves to be used in their new reactors. The NRC staff is monitoring the 

design and qualification process for squib valves to be used in new reactors by the applicable 

reactor vendors. The staff is also participating in international efforts to provide improved 

design, qualification, and testing for squib valves to be used in new reactors. ASME is 

developing improved OM Code provisions to address new reactors, including the consideration 

of surveillance activities for squib valves. Nuclear power plant licensees for new reactors will 

need to incorporate the lessons learned from the design and qualification process in the 

development of IST surveillance activities for squib valves. For example, in addition to the test 

firing sample of squib valve charges specified in the ASME OM Code, licensees of new nuclear 

power plants should address the following aspects in providing reasonable assurance of the 

operational readiness of squib valves: (1) verification of the structural integrity of external and 

internal parts of the actuator and valve; (2) identification and removal of foreign material, fluid 

and corrosion within the actuator and valve that might interfere with the operation of the actuator 

or valve; and (3) confirmation of the capability of the pyrotechnic charge in the actuator to 

provide the necessary motive force to operate the valve under design-basis conditions without 

damage to the valve body or connected piping. 
A 5-1

5. SUPPLEMENTAL GUIDANCE

ON INSERVICE TESTING OF PUMPS

5.1 General Pump Inservice Testing Issues

In 1995, OM Code Subsection ISTB introduced a new approach to pump testing, in which 

pumps are divided into two basic groups with an enhanced baseline or preservice and three 

periodic tests (i.e., Group A, Group B, and Comprehensive). This modified pump testing 

program is commonly referred to as the Comprehensive Pump Test (CPT). The pump grouping 

criterion of ISTB is based on the way the pumps are operated at the plant.

The CPT allows less-rigorous pump testing to be performed for certain pumps on a quarterly 

frequency while requiring a pump test to be performed with more accurate pressure and 

differential pressure instrumentation every 2 years at 20 percent of pump design flow. The 

CPT was developed with the knowledge that some pumps, such as containment spray pumps, 

cannot be tested at the required high flow rates because of system design limitations. 

Subsection ISTB-3300(e)(1) of the OM Code requires licensees to establish reference values 

within 20 percent of the design flow for the CPT.

5.1.1 Categories of Pumps for Inservice Testing

The OM Code requires that all pumps that the licensee identifies as part of an inservice testing 

(IST) program must be categorized as either Group A or Group B pumps. Subsection ISTB2000 defines Group A as “pumps that are operated continuously or routinely during normal 

operation, cold shutdown, or refueling operation.” By contrast, the Code defines Group B 

pumps as “pumps in standby systems that are not operated routinely except for testing.”

5.1.2 Testing Requirements and Frequency of Inservice Tests

The ASME OM Code identifies four types of tests, including Preservice, Group A, Group B, and 

Comprehensive tests. All pumps receive a Preservice test followed on a quarterly basis by the

test associated with the pump category (Group A test for Group A pumps, and Group B test for 

Group B pumps), and at least once every 2 years by a Comprehensive test. A Comprehensive 

test may also be substituted for a Group A or Group B test. Similarly, a Group A test may be 

substituted for a Group B test, and a Preservice test may be substituted for any inservice test.

ASME OM Code, ISTB-3410, “Pumps in Regular Use,” states -

Group A pumps that are operated more frequently than 3 months need not be run or 

stopped for a special test provided the plant records show the pump was operated at 

least once every 3 months at reference conditions, and the quantities specified were 

determined, recorded, and analyzed per Article ISTB-6000.

ASME OM Code, ISTB-3420, “Pumps in Systems Out of Service,” states -

For a pump in a system declared inoperable or not required to be operable, the test 

schedule need not be followed. Within 3 months before the system is placed in an 

operable status, the pump shall be tested and the test schedule followed in accordance 
A 5-2

with the requirements of this Subsection. Pumps which can only be tested during plant

operation shall be tested within 1 week following plant startup.

ASME OM Code, ISTB-3430, “Pumps Lacking Required Fluid Inventory”

Group B pumps lacking required fluid inventory (e.g., pumps in dry sumps) shall receive 

a comprehensive test at least once every 2 years except as provided in ISTB-3420. The 

required fluid inventory shall be provided during this test. A group B test is not required. 

NRC Recommendation

In 1995, OM Code Subsection ISTB introduced the new CPT, which allows licensees to perform 

less-rigorous pump testing for certain pumps on a quarterly frequency, while requiring licensees 

to perform a pump test with more accurate pressure/differential pressure instrumentation every 

2 years at 20 percent of pump design flow. This section also discusses previously issued 

guidance and experience. Nothing in the ASME OM Code shall be construed to supersede the 

requirements of a Technical Specification.

5.2 Use of Variable Reference Values for Flow Rate and Differential Pressure During 

Pump Testing

Some system designs do not allow for testing at a single reference point or a set of reference 

points. In such cases, it may be necessary to plot pump curves to use as the basis for variable 

reference points. Consequently, the ASME Code Committee introduced Code Case OMN-9, 

Revision 0, “Use of Pump Curves for Testing,” which the NRC staff subsequently included in

RG 1.192. That regulatory guide lists the OM Code Cases that the NRC staff finds acceptable 

for licensees to implement in their IST programs for light-water-cooled nuclear power plants. In

particular, the staff accepted Code Case OMN-9, with the condition that (1) when the repair, 

replacement, or routine servicing of a pump may have affected a reference curve, the licensee 

must determine a new reference curve, or reconfirm an existing reference curve, in accordance 

with Section 3 of Code Case OMN-9; and (2) if it is necessary or desirable, for some reason 

other than that stated in Section 4 of Code Case OMN-9, to establish an additional reference 

curve or set of curves, the licensee must determine the new curves in accordance with Section

3 of Code Case OMN-9. The use of OMN-9 requires relief because OMN-9 is only applicable to 

the ASME OM Code-1990 through ASME OMb Code 1992. It is not applicable to 1995 or later 

Code editions. In ASME OMb-2006 addenda, ASME OM Code committee, developed a new 

Code Case OMN-16, “Use of a Pump Curve for Testing,” to replace Code Case OMN-9. The 

OMN-16 incorporates all the NRC conditions imposed for the use of Code Case OMN-9. 

Therefore, once the Code Case OMN-16 is endorsed in RG-1.192, no relief request is need to 

use OMN-16. 

5.2.1 Reference Values

Licensees shall determine reference values from the results of Preservice testing or the first 

inservice test. The resultant reference values shall be at points of operation that are readily 

duplicated during subsequent tests, and the licensee shall compare all subsequent test results 

to the initial reference values or the new reference values established in accordance with the 

Code. Licensees shall only establish reference values when the pump is known to be operating 

acceptably. If the particular parameter being measured or determined can be significantly 

influenced by other related conditions, these conditions shall be analyzed.
A 5-3

5.2.2 Reference Curves

If the establishment of specific reference values is impractical for a centrifugal or vertical line 

shaft pump, the licensee may establish reference curves. In so doing, the licensee shall 

determine the reference curves from the data measured during Preservice testing or the first 

inservice test. In addition, the licensee shall establish a reference curve from a minimum of five 

data points for each 20 percent of the maximum pump curve range, and the range of the 

reference curve shall be sufficient to bound the points of operation that are expected during 

subsequent tests. The licensee shall then compare all results to the initial reference curves or 

the new values established in accordance with Sections 5.2.3 and 5.2.4, below. In addition, the 

licensee shall only establish reference curves when the pump is known to be operating 

acceptably. If vibration is relatively unaffected by changing differential pressure or flow over the 

reference curve range, the licensee may use a single reference value as the test quantity, 

provided it is at the minimum of the measured data. By contrast, if the licensee uses reference 

curves, the record of the test shall document and justify the reasons for doing so and the 

suitability of the methods used to develop the reference curves and acceptance criteria. (See 

Subsection ISTB-9000.)

5.2.3 Effect of Pump Replacement, Repair, and Maintenance on Reference Values or 

Reference Curves

When the repair, replacement, or routine servicing of the pump may have affected a reference 

value, a set of reference values, or a reference curve, the licensee shall determine a new 

reference value, set of reference values, or a new reference curve or reconfirm the previous 

value (or curve) by an inservice test run before declaring the pump operable. The licensee shall 

then identify any deviation between the previous and new set of reference values (or reference 

curves), and the record of the tests shall document the verification that the new values (or 

curves) represent acceptable pump operations. (See ISTB-9000.)

5.2.4 Establishment of Additional Sets of Reference Values or Reference Curves

If it is necessary or desirable, for some reason other than discussed above, to establish an 

additional set of reference values or curves, the licensee shall run an inservice test under the

conditions of an existing set of reference values, or within the range of existing reference 

curves, and shall analyze the results. If the operation is acceptable in accordance with 

Section 7 of Code Case OMN-9 (Section 16-6200 of Code Case OMN-16), a second test run 

under the new reference conditions shall follow as soon as practicable, and the results of this 

test shall establish the additional set of reference values or reference curves. Whenever a 

licensee establishes an additional set of reference values or reference curves, the record of the 

tests shall document and justify the rationale for doing so. (See ISTB-9000.)

NRC Recommendation

The NRC accepts the use of pump curves for reference values of flow rate and differential

pressure if the licensee clearly demonstrates, in a relief request, that it would be impractical to

establish a fixed set of reference values. A relief request must include a description of the 

methodology to be used in evaluating these pumps. To obtain approval for a proposed method 

of evaluating these pump parameters to detect hydraulic degradation and determine pump 

operability, the licensee must demonstrate that the acceptance criteria are equivalent to the 

Code requirements as specified under test acceptance criteria in ISTB Table ISTB-5121-1 or
A 5-4

ISTB-5221-1(depending on pump type), for allowable ranges using reference values and

curves.

To use this test method, the licensee must plot a valid pump characteristic curve from empirical 

data or obtain one from the pump manufacturer and verify it with measurements taken when the 

pump was known to be in good operating condition. Additional guidance is given in 

Sections 5.2.2, 5.2.3, and 5.2.4 above; the OM Code; and Code Case OMN-9, including 

RG 1.192). 

The use of OMN-9 requires relief because OMN-9 is only applicable to the ASME OM Code1990 through ASME OMb Code 1992. It is not applicable to 1995 or later Code editions. Code 

Case OMN-16 may be used once endorsed in RG 1.192.

Basis for Recommendation

Where it is not practical to return to the same flow configuration for each subsequent inservice 

pump test, the licensee must establish a method for evaluating the operational readiness of 

pumps in variable flow systems. This may be the case for service water or component cooling 

water systems and other systems where temperature or flow is controlled at a variety of 

locations. During quarterly pump testing, the licensee may not be able to manually control each 

of these local stations and duplicate the overall system reference conditions, as required by the 

Code.

Using the manufacturer’s pump-specific curves for flow and differential pressure, the licensee 

may be able to evaluate the pump in as-found system conditions. In implementing this 

guidance, the licensee would confirm these values by performing in situ testing. Another 

method would be to plot pump curves over the range of conditions expected during the system’s 

normal operation. It is also important to develop a method of evaluating pump vibration 

measurements taken with the pump operating over the range of possible as-found conditions, 

since this is a variable pump parameter. By evaluating these measurements of pump vibration, 

the licensee will ensure that a pump that is severely degraded, either hydraulically or 

mechanically, is declared inoperable and appropriate action is taken to address the degradation.

5.3 Allowable Variance from Reference Points and Fixed-Resistance Systems

Certain designs do not allow for the licensee to set the flow at an exact value because of 

limitations in the instruments and controls for maintaining steady flow. The characteristics of 

piping systems in other designs do not allow for the licensee to adjust the flow to exact values. 

The Code does not allow for variance from a fixed reference value, stating only that “the 

resistance of the system shall be varied until either the measured differential pressure or the 

measured flow rate equals the corresponding reference value.” Licensees have requested relief 

to establish a range of values similar to using a pump curve, but with a very narrow band. For 

example, one licensee proposed to use a reference curve with the tolerance around the 

selected value of flow to be 2 percent. Plant implementing procedures may instruct operators 

to set the flow to 1,500 gallon per minute (gpm). When this step is performed, the operator 

would attempt to set the flow as close as possible to 1,500 gpm and maintain it steadily at

approximately 1,500 gpm].
A 5-5

NRC Recommendation

The staff has determined that, if the design does not allow for establishing and maintaining flow 

at an exact value, achieving a steady flow rate or differential pressure at approximately the set 

value does not require relief for establishing pump curves. The allowed tolerance for setting the

fixed parameter must be established for each case individually, including the accuracy of the 

instrument and the precision of its display. This will necessitate verification of the effect of 

precision on accuracy as considered in the design of the instrument gauge. For Group A and 

Group B tests, a total tolerance of 2 percent of the reference value (including instrument 

accuracy) is allowed without prior NRC approval; for Preservice and Comprehensive tests, the 

allowable total tolerance is 1/2 percent (including instrument accuracy) for pressure and 

differential pressure,  2 percent (including instrument accuracy) for flow. For a tolerance 

greater than the allowed percent (which may be necessary depending on the precision of the 

instrument), the licensee may make a corresponding adjustment to acceptance criteria to 

compensate for the uncertainty, or may perform and document an evaluation to justify a greater 

tolerance. In using this guidance, the IST program document or implementing procedures must 

document the variance and the method for establishing it.

The intent is that the variance in the reference value setting may be 2 percent for Group A 

and Group B tests and 1/2 percent for Preservice and Comprehensive tests, 1/2 for 

pressure and differential pressure, and 2 percent for flow without requiring relief. 

Nonetheless, any variance in the setting will have an impact on test results.

Basis for Recommendation

The OM Code does not address the likelihood that it may not be possible to control a flow rate 

or differential pressure to an exact value. When the Code specifies that the system resistance 

must be varied until either the flow or differential pressure equals the corresponding reference 

value, it does not intend the set value to have an acceptable range as stated in the ISTB test 

acceptance criteria, including ISTB Tables ISTB-5121-1 and ISTB-5221-1. The acceptance 

criteria apply only to the parameter being determined after the resistance is varied. Licensees 

should recognize that the reference value for certain pumps can only be achieved within a 

specified tolerance. Licensees may set the repeatable parameter as close as possible to the 

reference value during each test, rather than treating any variance in the value with a pump 

curve. If, upon establishing trends in data, the licensee determines that the parameter varies 

such that the readings are outside the accuracy of the instrument, the licensee may need to 

establish pump curves and propose an alternative to the Code requirements for the applicable 

pumps. (See Section 5.2)

Subsection ISTB-3500 specifies the requirements for instrument fluctuations and describes the 

basis for allowing a variance from the reference value of 2 percent for Group A and Group B

tests, and 1/2 percent for pressure and differential pressure, 2 percent for flow for 

Preservice and Comprehensive tests. In addition, Subsection ISTB-3500 allows the use of 

symmetrical damping devices or averaging techniques to reduce instrument fluctuations to 

within 2 percent or 1/2 percent (as applicable) of the observed reading for values specified 

in the implementing procedures. Greater variances must be justified and acceptance criteria 

adjusted as necessary.
A 5-6

If an analog gauge is used, the required accuracy is percent full scale. For a digital gauge, the

required accuracy is over the calibrated range. For a combination of gauges, the required 

accuracy is loop accuracy.

The ASME is currently in the process of developing a Code change as follows to address 

this issue:

The allowed tolerance for setting the fixed parameter must be established for each case 

individually, including evaluation of throttling capability. Licensees should consider 

improvements in throttling methods where system control is especially poor. A total throttling 

tolerance of + 2 / -1 percent of the flow rate reference value is considered as meeting the 

requirements of the code sections.

For a tolerance greater than + 2 / -1 percent of the flow rate reference value, a corresponding 

adjustment to acceptance criteria shall be made to compensate for the uncertainty, or an 

evaluation would be performed and documented justifying a greater tolerance. The variance 

and the method for establishing the variance must be documented in the IST program 

documents or implementing procedures.

The basis for the Code change is as follows:

The ASME OM Code does not address the possibility that a flow rate or differential pressure 

may not be controllable to an exact value. When the Code specifies that the system resistance 

be varied until either the flow or differential pressure equals the corresponding reference value, 

it does not literally intend that the "set value" be precisely attained without any fluctuations. 

Licensees recognize that the reference value for certain pumps can only be achieved within a 

specified tolerance. Licensees shall attempt to set the repeatable parameter as close as 

possible to the reference value during each test. 

The basis for allowing a variance of + 2 / -1 percent from the flow rate reference value deals 

with instrument fluctuations and system stability issues. The Code allows symmetrical damping 

devices or averaging techniques to be used to reduce instrument fluctuations to within 2 percent 

of the observed reading for values specified in the implementing procedures. Greater variances 

must be justified and acceptance criteria adjustments made as necessary. The limitation of 1% 

in the negative direction reduces the non-conservative impact on the variable parameter. The 

total 3% allowable variance provides for a reasonable throttling control range while minimizing 

the impact on trendability of the variable parameter.

Licensees should ensure that performance trending of pumps is capable of detecting 

degradation as early as possible. Larger variances in the reference parameter will induce 

scatter in the variable parameter data. Techniques such as data normalization, where recorded 

test data is corrected by the known pressure to flow relationship, should be used when 

necessary to provide for accurate short term trending. 
A 5-7

5.4 Monitoring Pump Vibration in Accordance with ISTB

OM Subsection ISTB allows licensees to monitor pump vibration in units of either pump 

displacement (peak-to-peak) or pump velocity (peak), and includes acceptance criteria for both 

units of measurement. As specified in OM Table ISTB-3000-1, the measurement of pump 

vibration is required for Preservice, Group A, and Comprehensive tests. However, the Code 

does not require vibration measurements for Group B tests. The staff has determined that if the 

licensee uses OM Subsection ISTB as the basis for monitoring vibration in the IST program, the 

program must include all of the requirements for such monitoring. Licensees may update their

programs in accordance with this position without further relief if they meet all related 

requirements for monitoring vibration in Subsections ISTB-3540, ISTB-5000, and ISTB-6000. 

However, Commission approval to use a later Code edition or addenda is still required pursuant 

to 10 CFR 50.55a(f)(4)(iv). See sections 1.1 and 2.1 for further guidance.

In following this guidance, the frequency response range of the instrumentation must be as 

specified in Subsection ISTB-3510(e) for both low-speed and high-speed pumps, unless the 

licensee demonstrates that the information gained at the low-frequency response does not 

apply for the bearing design of the pumps. In that event, the licensee must still provide an 

acceptable alternative to the required testing. Although the instruments in the low-frequency 

response ranges may not be widely used, the unavailability of instruments does not constitute 

sufficient justification for either obtaining relief from the frequency response range requirements 

of Subsection ISTB, or obtaining approval of an alternative to the requirements.

Basis for Recommendation

As shown in the ASME OM Code Figure ISTB-5223-1, “Vibration Limits,” licensees may choose 

to use units of velocity, rather than displacement, in measuring vibration in pumps that operate 

above 600 revolutions per minute (rpm). Such an approach would enable the licensee to more 

rapidly detect wear in the anti-friction bearing and other types of pump degradation and, thus, 

effect repairs in a more timely manner.

Pump bearing degradation results in increased vibration at frequencies of 5 to 100 times the 

rotational speed of the pump. These high-frequency bearing vibrations may not significantly 

increase the measured displacement of pump vibration and could go undetected. However, the

high-frequency vibrations would significantly increase the measured velocity of pump vibration, 

which could indicate the need for corrective action before the bearing fails. Because pump 

bearings vibrate at high frequencies, the measured vibration velocity indicates the mechanical 

condition of the pumps and reveals pump bearing degradation much more accurately than does 

measured vibration displacement.

Advantages of measuring vibration velocity, rather than displacement, to monitor the 

mechanical condition of pumps (with the exception of low-speed pumps) are widely 

acknowledged in the industry. Many licensees measure pump vibration velocity to detect pump 

degradation and obtain advance warning of incipient pump bearing failure. Upon obtaining this 

advance warning, the licensee can plan and prepare for maintenance during scheduled outages 

instead of suffering losses resulting from unplanned outages to repair failed critical equipment.
A 5-8

ASME OM Code, Subsection ISTB includes a set of allowable ranges for inservice pump 

vibration velocity and measured pump vibration displacement. These ranges are based on an 

evaluation of empirical data and various acceptance criteria for pump vibration velocity 

established by U.S. industries, academia, international industry, and foreign agencies.

The minimum frequency response range requirement is established from one-third of the 

minimum pump shaft rotational speed to at least 1000 Hz in order to encompass all noise 

contributors that could indicate degradation. Instruments with a frequency response range that 

meets these requirements for slow-speed pumps may not be widely used. However, the 

unavailability of instruments, alone, does not constitute adequate justification for obtaining relief 

or approval of an alternative; however, it may be a significant element in the justification. The 

NRC has observed that, because of technology advancement and research in the field of 

instrumentation, vibration measuring transducers meeting the Code requirements can now be 

procured from various suppliers at reasonable costs. Additionally, frequencies less than running 

speed may not be indicative of problems for certain types of bearings; however, subharmonic 

frequencies may be indicative of rotor rub, seal rub, loose seals, or coupling damage. The type 

of bearings and other subharmonic concerns would typically be discussed in the justification for 

relief.

ASME OM Code, Subsection ISTB requires licensees to measure vibration in units of either 

pump displacement (peak-to-peak) or pump velocity (peak). Digital equipment can measure 

directly in peak units. The 10-year update of the ISI and IST programs reflects the need for

licensees to incorporate new technologies that have been incorporated into the codes and 

standards. The ASME OM Code Committee responded to an inquiry (Interpretation 95-4, File 

OMI 94-2) explaining that the intent of the ASME OM Code is to allow vibration to be measured 

in root mean square (rms) and mathematically converted to peak readings. Licensees are 

cautioned that the Code vibration acceptance criteria are in peak or peak-to-peak units, and the 

use of rms is not acceptable without a mathematical conversion. To comply with the 

requirements, licensees that use rms values for recording data must adjust the limits of OM 

Subsection ISTB, or convert the data to peak values.

Several plants have requested an alternative to the vibration acceptance criteria of Section ISTB 

for smooth-running pumps, and the NRC has approved such requests. However, licensees with 

such approval must continue to assess the vibration data and monitor increases that may be 

indicative of a change. In one reported incident, a pump with very low vibration experienced an 

increase in vibration levels over three successive tests, although the levels remained below the 

criteria for smooth-running pumps. Upon investigating the cause of the increase, the licensee 

determined that the bearing had degraded and required replacement.

5.5 Pump Flow Rate and Differential Pressure Instruments

The NRC has received requests for relief to continue using instruments that do not meet either 

the range or accuracy requirements of the Code. The Code requires each analog instrument to

have a full-scale range that is three times the reference value or less, while each digital 

instrument must be such that the reference values do not exceed 70 percent of the calibrated 

range of the instrument. The NRC has accepted Code Case OMN-6 as specified in RG 1.192, 

which allows each digital instrument to be such that the reference values do not exceed 90

percent of the calibrated range of the instrument. For Group A and Group B pumps, OM

Subsection ISTB-3510 requires an accuracy of 2 percent of full-scale for analog instruments, 

2 percent of total loop accuracy for a combination of instruments, or 2 percent of reading 
A 5-9

over the calibrated range for digital instruments. For Preservice and Comprehensive tests, the 

required instrument accuracy is 1/2 percent for pressure and differential pressure instruments.

5.5.1 Range and Accuracy of Analog Instruments

NRC Recommendation

When the range of a permanently installed analog instrument is greater than three times the 

reference value, but the accuracy of the instrument is more conservative than that required by 

the Code, the staff may grant relief when the combination of the range and accuracy yields a 

reading that is at least equivalent to that achieved using instruments that meet the Code 

requirements (i.e., up to 6 percent for Group A and B tests, and 1.5 percent for pressure 

and differential pressure instruments for Preservice and Comprehensive tests). The use of a 

test gauge (in lieu of a permanent instrument) is acceptable if the reading is at least equivalent 

to that required by the Code. When using temporary instruments, the staff recommends that the 

licensee’s IST records should include an instrument number for use in tracing each instrument 

and a calibration data sheet for use in verifying that the instruments are accurately calibrated. 

The licensee need not obtain relief if the temporary instruments meet the range and accuracy 

requirements of the Code. If relief is requested, the licensee would typically describe the effect 

on each group of applicable pumps and would typically discuss adjustment of acceptance limits 

to account for the inaccuracies.

Basis for Recommendation

Because the IST requirements originally specified an instrument range of 4 times the reference 

values or less, the permanent instruments in many early-licensed plants do not meet the current 

requirements of the Code for an instrument range of three times the reference values or less. 

The NRC does not generally consider instrument installation or replacement an undue burden, 

and compliance with the instrument requirements in later editions of the Code does not 

constitute a backfit.

This position applies to the early-licensed plants, but not for the purchase of replacement 

instruments that can be procured to meet the current requirements of the Code; therefore, for

new instrument installations, licensees must meet the accuracy and range requirements 

(although the Code does not prohibit like-for-like instrumentation for the existing installation).

The licensee is not relieved of its responsibility to make modifications to comply with changes to 

IST as a result of changes to the Code. Instrument modifications are considered practical in the 

context of 10 CFR 50.55(a)(f)(4). However, the use of any available instruments that meet the 

intent of the Code requirements for the actual reading would yield an acceptable level of quality 

and safety for testing. Licensees are required to submit a relief request in this case.

When the licensee submits a relief request, it should separately address each group of affected 

pumps if the instruments are permanently installed. By contrast, a general relief request may be 

acceptable for temporary instrumentation. If the instruments do not meet the intent of the Code 

requirements, the NRC may require the licensee to adjust acceptance limits to account for the 

instrument inaccuracies.
A 5-10

Licensees are cautioned that the CPT requires more accurate instruments than those specified 

in earlier editions of the Code. As a result, licensees must verify that instrument accuracy is 

appropriate for the type of test being performed (Group A or Group B versus a Comprehensive 

test). Licensees should also note that previously acceptable instruments may no longer be 

acceptable when updating to a more recent edition of the Code.

5.5.2 Use of Tank Level to Calculate Flow Rate for Positive Displacement Pumps

The NRC has received requests for relief to use the tank level to calculate the flow rate in a 

system with a positive displacement pump when the system was not designed with a flow meter 

in the flow loop.

OM Subsection ISTB-3550 requires licensees to measure the pump flow rate using a rate or 

quantity meter installed in the pump test circuit. If the meter does not directly indicate the flow 

rate, the record of the test shall identify the method used to reduce the flow data. In addition, 

Subsection ISTB-5300(a) requires a 2-minute run time in order to achieve stable pump 

performance parameters before recording data during the test.

NRC Recommendation

When flow meters are not installed in the flow loop of a system with a positive displacement 

pump, it is impractical to directly measure flow rate for the pump. The staff has determined that, 

if the licensee uses the tank level to calculate the flow rate as described in Subsection ISTB

3550, the implementing procedure must include the calculational method and any test 

conditions needed to achieve the required accuracy. Specifically, the licensee must verify that 

the reading scale for measuring the tank level and the calculational method yield an accuracy 

within 2 percent for Group A and B tests, and Preservice and Comprehensive Tests. If the 

meter does not directly indicate the flow rate, the record of the test shall identify the method 

used to reduce the flow data.

Basis for Recommendation

The OM Code requires licensees to measure the pump flow rate in order to determine the 

extent of any pump degradation. A minimum pump run time of 2 minutes is required in order to

achieve stable performance parameters before recording data during the test.

Requiring licensees to install a flow meter to measure the flow rate and to guarantee the test 

tank size, such that the pump flow rate will stabilize in 2 minutes before recording data would be 

a burden because of the design and installation changes to be made to the existing system. 

Therefore, compliance with the Code requirements would be a hardship.

The average flow rate is calculated by measuring the change in test tank level over a period of

time and converting it to flow rate using the following standard formula:

Q (GPM) =  ΔL (inch) / Δt (Second)

Where:Q is flow rate

 is a constant which reflects tank dimensions and unit conversions

ΔL is the measured change in level in the tank in time Δt.
A 5-11

Pump discharge pressure will match system pressure up to the shutoff head of the positive 

displacement pump. Because of the characteristics of a positive displacement pump, there 

should be virtually no change in pump discharge flow rate as a result of the rising tank level. 

Therefore, rising tank level will not have an impact on test results. By having approximately the

same level in the tank at the beginning of each test, licensees can achieve repeatable results. 

In addition, the suction would be from a large source at a constant pressure, which will allow 

pump performance parameters to stabilize quickly. This method would provide reasonable 

assurance of operational readiness, provided that the licensee measures the test tank level in 

accordance with the accuracy requirements of OM Table ISTB-3500-1. The implementing 

procedures should document the calculational method and test conditions required to achieve 

this accuracy. Therefore, the proposed alternative of using the tank level to calculate the flow 

rate provides reasonable assurance of operational readiness. Licensees must submit a relief 

request to implement this proposed alternative.

5.5.3 Use of Tank or Bay Level to Calculate Differential Pressure

The NRC has received requests for relief to use the tank or bay level to calculate differential 

pressure when a direct measurement of inlet pressure or differential pressure is not available.

NRC Recommendation

When inlet pressure gauges are not installed in the inlet of a vertical line shaft pump, it is 

impractical to directly measure inlet pressure for use in determining differential pressure for the 

pump. The staff has determined that, if the licensee uses the bay level to calculate the suction 

(inlet) pressure as described in Subsection ISTB-3520(b), the implementing procedure must 

include the calculation. The licensee must also verify that the reading scale for measuring the 

level and the calculational method yield an accuracy within 2 percent for Group A and B tests, 

and 1/2 percent for Preservice and Comprehensive tests. If direct measurements are 

impractical for other types of pumps with suction from a tank, the licensee must apply similar 

controls. The Code allows the licensee to determine differential pressure by obtaining the 

information from a differential pressure gauge or differential pressure transmitter, or by 

determining the difference between the pressure at a point in the inlet pipe and the pressure at 

a point in the discharge pipe (Subsection ISTB-3520(b)). Therefore, the licensee may 

implement a calculational method without obtaining relief because the ASME Code allows for 

the determination of differential pressure from the discharge pressure and the pressure in the 

pump inlet.

Basis for Recommendation

The method is in accordance with a determination of differential pressure allowed by the Code. 

Although the inlet pressure is not directly measured, it is “measured” for the purpose of 

determining the pressure at a point in the inlet. By including the calculation in the implementing 

procedures, the licensee can determine the differential pressure in a manner that is consistent 

and repeatable from test to test. This method will yield the information needed for monitoring 

the hydraulic condition of the applicable pumps without the need to install suction (inlet) 

pressure gauges, which may not be practical, depending on the design limitations in the inlet of 

the pump.
A 5-12

5.5.4 Accuracy of the Flow Rate Instrument Loop

As clarified in OM Code Interpretations 95-7 (OM-1990, Subsection ISTB 4.6.1 and Table ISTB 

4.6.1.1; OM-1987 with OMa-1988, Part 6, Para. 4.6.11 and Table 1, Instrument Accuracy ) and 

Inquiries IN 91-3 (OM-1987 through OMc-1990, Part 6, Para. 4.6.1.1) and IN 91-037 (ASME 

Section XI, 1977 Edition Through Later Editions and Addenda through the 1987 Addenda, Table 

IWP-4110-1, Instrument Accuracy-Flowrate), the accuracy requirements of analog instruments 

that are used to measure process flow apply only to the reference calibration of the instrument, 

such as that supplied by the instrument manufacturer, in determining loop accuracy. In 

determining instrument accuracy, the Code does not explicitly require the licensee to consider 

physical attributes (such as orifice plate tolerances), tap locations, environmental effects (such 

as temperature, radiation or humidity), vibration effects (such as seismic) or process effects 

(such as temperature). However, factors associated with attributes that could affect the 

measurements include the effects of wear, accumulation of dirt or grease on an annubar flow 

coefficient, and the reversed installation of a one-direction orifice plate.

NRC Recommendation

The Code requirements for instrument accuracy ensure that the instrument loop accuracy is 

adequate for monitoring pumps for degrading conditions. The accuracy for analog instruments 

specified in OM Subsection ISTB 3500 applies only to the calibration of the instruments. The 

staff recommends that, when test results indicate that conditions in the pump or the test circuit 

have changed, licensees should consider corrective action for other attributes that could affect 

the overall loop accuracy of the measurements.

Basis for Recommendation

In ASME Code Interpretation 95-7 and ASME Code Inquiries IN 91-3 and IN 91-037, the ASME 

Code Committee states that the requirements for the final indication of flow rate on an analog 

instrument to be within 2 percent of full scale of actual process flow rate applies only to the 

calibration of the instrument and does not take into account physical attributes, environmental 

effects, vibration effects, or process effects.

5.6 Operability Limits of Pumps

For details see NRC Inspection Manual Part 9900: Technical Guidance, Appendix-C, “Specific 

Operability Issues,” Section C.8, “TS Operability vs. ASME OM Code Criteria.” 

5.7 Duration of Tests

Subsections ISTB-5100(a), ISTB-5200(a), and ISTB-5300(a), “Duration of Tests,” requires that 

for measuring parameters as specified in Table ISTB-3000-1, each pump shall be run for at 

least 2 minutes after pump conditions are stable as the system permits. This duration is 

applicable to Group A tests and CPTs. The staff recommends that this duration also be applied 

to Group B pump tests.
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Basis for Recommendation

The 2-minutes run time is adequate after pump operation becomes stable. This 2-minute run 

time minimizes overheating of pumps that are tested using the minimum flow recirculation line. 

The NRC recommends that licensees should minimize the time pumps are operated on the 

minimum flow recirculation line. (See NRC Bulletin 88-04, “Potential Safety-Related Pump 

Loss.”)

5.8 Adjustments for Instrument Inaccuracies

If the accuracy of plant instrumentation used for IST is not well understood, the test results may 

not be adequate to meet the licensee s safety analysis, even if they meet the Code 

requirements. For example, TSs or the safety analysis report require a pump to produce 1,000

gpm at 500 pounds per square inch differential (psid), but the IST reference values are 1,000 

gpm (fixed) and 550 psid. The low end of the acceptable range for differential pressure from 

ISTB Table ISTB-5121-1 for Group A and Group B tests (0.90) would be 495 psid, although 

conservatively set at 500 psid. If this test is also to prove operability of the pump in addition to 

meeting IST requirements, and the 2 percent instrument inaccuracies were taken into account 

for flow rate and differential pressure, there is the possibility that the pump is putting out less 

than the required values. In this example, the instrument accuracies would need to be taken 

into account if they were not already considered when the design parameters were developed.

When pump test procedures are developed, limits in the safety analysis cannot be ignored. The 

IST requirements are written generally. If specific plant limits are more conservative, to ensure 

compliance with design-basis assumptions, such limits must be clearly indicated as the 

“operability” limits and used for acceptance criteria of IST. For example, when obtaining values 

using instrumentation that meets the accuracy requirements specified for “information only” or 

for IST, the value as-read would be used. If a licensee is attempting to perform a critical test, 

more accurate instrumentation may be necessary; however, the value recorded would be the 

value read if the accuracy of the instrumentation met the specified accuracy. Only when 

instruments are used that cannot meet the specified accuracy for a test would an adjustment be 

necessary to meet the Code. Design analyses may not account for instrument accuracy 

readings; however, when the pump selection is made, the designer generally selects from a 

catalog of available sizes and chooses one with margin above the analyses numbers. 

In the determination of loop accuracy, it is intended that only the instrument manufacturer’s 

reference accuracy be considered. It is not necessary to consider all uncertainties (such as 

environmental effects, process effects, vibration effects, etc.).

5.9 Pump Testing Using Minimum Flow Return Lines With or Without Flow Measuring 

Devices

The NRC has received relief requests from licensees requesting approval of pump testing by 

using minimum flow return lines with or without measuring devices as an alternative to the IST 

requirements as specified in the Subsection ISTB of the OM Code.

As specified in Section 5.1.2 above, the Code identifies four types of tests, including Preservice, 

Group A, Group B, and Comprehensive tests. All pumps receive a Preservice test followed on a 

quarterly basis by the test associated with the pump category (Group A test for Group A pumps, 

and Group B test for Group B pumps), and at least once every 2 years by a Comprehensive 
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test. A Comprehensive test may also be substituted for a Group A or Group B test. Similarly, a 

Group A test may be substituted for a Group B test, and a Preservice test may be substituted for 

any inservice test.

Subsections ISTB-5100(b), ISTB-5200(b), and ISTB-5300(b) of the ASME OM Code allow the 

use of a bypass test loop for Group B tests, provided that it is designed to meet the pump 

manufacturer’s operating specifications (e.g., flow rate, time limitations) for minimum flow 

operation. The bypass test loop may be used for Group A or Comprehensive tests, provided 

that the flow rate through the loop meets the requirements specified in Subsection ISTB 3300.

An inservice pump test requires that the pump parameters shown in Table ISTB-3000-1 must be 

measured and evaluated to determine pump condition and detect degradation. Pump

differential pressure and flow rate are two parameters that are measured and evaluated 

together to determine pump hydraulic performance.

In cases where only the minimum-flow return line is available for pump testing, regardless of the 

test interval, the staff’s position is that flow instrumentation that meets the requirements of 

Subsection ISTB-3500 should be installed in the mini-flow return line. Installation of this 

instrumentation is necessary to provide flow rate measurements during pump testing so that this 

data can be evaluated with the measured pump differential pressure to monitor for pump 

hydraulic degradation. 

When testing a pump using a minimum flow recirculation line, the guidance provided in NRC 

Bulletin 88-04 applies. Licensees should review operating conditions to ensure that a pump is 

not subject to dead head conditions and that the minimum flow line is adequately sized and that 

operation will not damage the pump.

5.10 Alternative to ASME OM Code Comprehensive Pump Testing Requirements

The NRC has received relief requests from licensees requesting approval of alternatives to the 

Comprehensive Pump Testing (CPT) requirements specified in Subsection ISTB-3300, 

“Reference Values” of the OM Code.

The CPT was developed with the knowledge that some pumps, such as containment spray 

pumps, cannot presently be tested at the required high flow rates because of system design 

limitations. Consequently, Subsection ISTB-3300(e)(1) requires licensees to establish 

reference values within 20 percent of the design flow for the CPT.

Some designs do not allow a CPT at a pump design flow of 20 percent because of the original 

system design configuration. In such cases, it may be necessary to use the pump’s 

recirculation line for IST; however, recirculation lines are not typically designed to accommodate 

20 percent of the design flow.

NRC Recommendation

The NRC accepts the use of lower flow (reference values) other than 20 percent of the design 

flow, as specified by OM Subsection ISTB-3300(e)(1) for CPT, if the licensee’s relief request 

clearly demonstrates the impracticality of establishing a reference value within 20 percent of 

the design flow for the CPT.
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To obtain approval for a proposed alternative method of performing the CPT with a flow other 

than as specified in Subsection ISTB-3300(e)(1) and measuring pump parameters to detect 

hydraulic degradation and determine pump operability, the licensee must demonstrate that the 

acceptance criteria are equivalent to the CPT Code requirements in Subsection ISTB3300(e)(1).

To show the impracticality, the licensee should include (as a minimum, but not limited to) the

following information in the submitted relief request:

(1) Provide reason(s) for not performing the CPT at the required flow of 20 percent of 

pump design flow.

(2) Specify the maximum flow at which the CPT can be performed.

(3) Provide the estimated cost of any temporary or permanent system modification required 

to enable a CPT to be performed at 20 percent of pump design flow, along with any 

difficulty associated with implementing the modification.

(4) Provide all details (e.g., temporary modifications of piping, containment sump, etc.), 

including pump performance curves, if a full-flow test was performed during preservice 

or service of the plant.

(5) Provide pump performance curves and any other data associated with the pump’s shop 

testing provided by the manufacturer.

(6) Provide the records and history of maintenance and repairs performed on the pump.

(7) Provide any appropriate compensatory actions being proposed to supplement the 

alternative testing, such as (but not limited to) the following examples:

(a) testing at the best efficiency point (BEP) on a longer interval; BEP is defined as 

the capacity and head at which the pump efficiency is at its maximum

(b) commitment to perform additional performance monitoring

(c) adjustment of acceptance criteria

(d) continuation of the previous Code testing, including taking overall vibration data 

quarterly

(e) periodic sampling and analysis of the lube oil

Additional guidelines are included in NUREG/CP-0152, Volume 4, Proceedings of the Seventh 

NRC/ASME Symposium on Valve and Pump Testing, entitled “Comprehensive Pump Testing 

Based on ASME OM Code Requirements and its Alternative and Related Relief Requests.”

This guidance requires relief because the Code does not allow for a reference value of flow for 

CPTs, other than at flow rates within 20 percent of pump design flow. The NRC will review 

any relief requests on a case-by-case basis. 

5.11 Motor Drivers for Pumps

Pump drivers are outside the scope of the OM Code, with the exception of vibration testing for 

vertical line shaft pumps where the driver is an integral part of the pump. Most pumps are 

driven by electric motors, which are connected via coupling shafts. Motor vibration attributable 

to coupling misalignment may not be realized or measured at the pump, and small changes in 

the vibration of a motor can have significant effects on pump operation and the operational 

readiness.
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Institute of Electrical and Electronics Engineers (IEEE) Standard 741-2007, “IEEE Standard 

Criteria for the Protection of Class 1E Power Systems and Equipments in Nuclear Power 

Generating Stations,” briefly address the vibration issue, and refers to IEEE C37.96-2000, “IEEE 

Guide for AC Motors Protection,” for motors. IEEE 741-2007 includes the testing and the 

surveillance requirements and lists several standards in the reference section for testing. IEEE 

C37, 96-2000 extensively address the vibration issue on electric motors because of its 

significant impact on bearings, lubricants, protective devices, etc. 

5.12 Pumps in New Reactors

RG 1.206 provides guidance for the development of IST programs for pumps by COL licensees. 

In Commission papers SECY-90-016 and 93-087, the NRC indicated that new nuclear power 

plants should be designed to accommodate full flow testing of pumps within the scope of the 

IST program. As part of its review of design certification and COL applications, the NRC staff is 

evaluating whether new nuclear power plants will be designed to allow full flow testing of pumps 

in the IST program. The capability to perform full flow testing should be reflected in the 

development of the IST activities for pumps in new reactors. For example, ASME is considering 

a revision to the ASME OM Code testing provisions for pumps in new reactors to provide 

comprehensive pump testing described in the ASME OM Code on a quarterly frequency. The 

staff will conduct inspections of the development and implementation of the IST program 

(including full flow testing of pumps) during construction and operation of new nuclear power 

plants. 
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6. REVISED STANDARD TECHNICAL SPECIFICATIONS

Details about Standard Technical Specifications see NUREG-1430, through NUREG-1434.

The Administrative Controls TS 5.5 includes a requirement to establish, implement and 

maintain a program entitled “Inservice Testing Program.” This program provides controls for 

inservice testing of ASME Code Class 1, 2, and 3 components. 
A 7-1

7. IDENTIFICATION OF CODE NONCOMPLIANCE

7.1 Nonconforming Conditions

For details see Section 5.4, “Identification of Code Noncompliance” in the Main Section of this 

NUREG.
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7.2 Starting Points for Time Periods in Technical Specification Action Statements

For details see Standard Technical Specification NUREG-1430 through NUREG-1434, 

LCO 3.0.2.
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8. RISK-INFORMED INSERVICE TESTING

8.1 Introduction

10 CFR 50.55a, “Codes and standards,” paragraph (f), “Inservice Testing Requirements,”

requires, in part, that Class 1, 2, and 3 pumps and valves must meet the requirements of the

ASME OM Code. 

RG 1.175, “An Approach for Plant-Specific, Risk-Informed Decision-Making: Inservice Testing,”

August 1998, describes an acceptable alternative approach for applying risk insights from 

probabilistic risk assessment (PRA), in conjunction with established traditional engineering 

information, to make changes to a nuclear power plant’s IST program. The approach described 

in RG 1.175 addresses the high-level safety principles specified in RG 1.174 and attempts to 

strike a balance between defining an acceptable process for developing risk-informed inservice 

testing programs without being overly prescriptive. The resultant risk-informed IST programs 

will have improved effectiveness with regard to the utilization of plant resources, while still 

maintaining acceptable levels of quality and safety. However, licensees may propose other 

approaches for consideration by the NRC staff. It is intended that the approach presented in 

RG 1.175 should be regarded as examples of acceptable practices, and that licensees should 

have some degree of flexibility in satisfying regulatory requirements on the basis of their 

accumulated plant experience and knowledge.

8.2 Discussion

Until such time as a risk-informed alternative to the current Code requirements is incorporated 

by reference into the regulations, the alternative approach described in RG 1.175 must be 

authorized by the NRC pursuant to 10 CFR 50.55a(a)(3)(i) on a plant-specific basis prior to 

implementation. Because 10 CFR 50.55a(a)(3)(i) places no restrictions on the scope of 

alternatives that may be authorized, licensees may propose risk-informed alternatives to their

entire IST program, or may propose alternatives that are more limited in scope (e.g., for a

particular system or group of systems, or for a particular group of components). In either case, 

the staff expects the licensee’s proposal to address the principles described in RG 1.175, 

including those related to implementation and monitoring.

If a licensee proposes a risk-informed alternative to the ASME Code test requirements, the

application should contain a summary description of the proposed alternative. The summary 

description should specify the key technical and administrative aspects necessary to describe 

and control the risk-informed alternative. The NRC staff will review and approve this summary 

description pursuant to 10 CFR 50.55a(a)(3)(i) and, as such, the summary description will serve 

as the framework within which the licensee may make future changes to its risk-informed 

alternative without having to resubmit it for NRC approval.

8.3 Online Inservice Testing

In an effort to shorten refueling outages, many licensees are trying to perform as much 

maintenance, testing, and surveillance as possible with the nuclear power plant on line. For

example, several licensees have submitted relief requests to obtain NRC approval to conduct 

inservice testing once per refueling cycle, rather than during the refueling outage as prescribed 

by the Code. In preparing (and evaluating) such relief requests, licensees (and the NRC staff) 
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should consider several factors to ensure that the proposed alternative provides an acceptable 

level of quality and safety.

If a licensee is testing a particular pump or valve during refueling outages, the licensee may

have determined that it is impractical to test the pump or valve quarterly during operation. The 

licensee s IST program document should, therefore, discuss the basis for deferring the testing 

from quarterly (and during cold shutdowns) to refueling outages. Relief requests to perform 

testing once each refueling cycle with the nuclear power plant on line should be prepared in light 

of the refueling outage justification for each affected valve or group of valves. If necessary, the 

licensee should revise the refueling outage justification to be consistent with the relief request.

Licensees (and the NRC staff) should also consider whether the testing can be accomplished 

within the allowed outage time permitted by any applicable Technical Specification (TS). In

general, the time necessary to complete the testing should be significantly less than the allowed 

outage time. This is to preclude TS violations or the need to issue exigent TS amendments or 

notices of enforcement discretion (NOEDs). In addition, licensees should not conduct noncorrective maintenance/testing activities at power if the associated post-maintenance testing 

cannot reasonably be accomplished until the next outage.

Sometimes, there is a tradeoff between testing these components at power (e.g., when they 

could be needed to mitigate the consequences of an accident) and testing them during outages 

(e.g., when there may be greater reliance on shutdown cooling or when other equipment is

necessarily out-of-service). Licensees should quantitatively or qualitatively address the risks 

associated with testing components on line, rather than testing during the refueling outage. If

the proposed testing could have a significant risk impact, or if its justification includes 

risk-related arguments, the relief request should be prepared and reviewed in accordance with 

RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on

Plant-Specific Changes to the Licensing Basis,” and Appendix D to Standard Review Plan 

(SRP) Chapter 19, (NUREG 0800) as applicable. Licensees should also identify any 

compensatory measures to be established as a means to reduce the impact (e.g., risk and 

operational worker safety) of testing with the nuclear power plant at power.2

 If relevant, 

licensees should also provide information on how testing at power (rather than testing during 

refueling outages) will affect scheduled maintenance work windows for the applicable system 

(i.e., whether the testing can be completed within the work windows or whether it will extend 

either the shutdown or at-power work windows). In addition, licensees will need to develop a 

new estimate of the maintenance unavailabilities that reflects the increased maintenance 

activities at power, and will need to document the basis for the new estimate (e.g., use plant 

logs or maintenance data to include in the current estimate of the maintenance unavailabilities 

those activities that were being performed during shutdown that will now be performed at 

power).

2 In should be noted that the assessment of risk resulting from performance of maintenance activities as 

required by 10 CFR 50.65(a)(4) of the Maintenance Rule is not sufficient justification for testing components 

at power. This assessment is required for maintenance activities performed during power operations or 

during shutdowns. This configuration risk management does not address the relative merits of testing at 

power verses testing during refueling outage.
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At times, testing (or the disassembly and inspection of components) during refueling outages 

can be more advantageous than at-power operations from a worker safety perspective (for

example, systems may be cold and depressurized). When requesting NRC approval to perform 

testing with the nuclear power plant on line, licensees should consider worker safety and should 

discuss whether the applicable components can be adequately isolated and restored.

In Section 11.2.3 of NUMARC 93-01, Rev. 2, “Industry Guideline for Monitoring the 

Effectiveness of Maintenance at Nuclear Power Plants,” the Nuclear Management and 

Resources Council (NUMARC, now NEI) provided additional guidance for conducting online 

maintenance and testing. It states, in part -

Online maintenance [and testing] should be carefully managed to achieve a balance 

between the benefits and potential impacts on safety, reliability or availability. For 

example, the margin of safety could be adversely impacted if maintenance is performed 

on multiple equipment or systems simultaneously without proper consideration of risk, or 

if operators are not fully cognizant of the limitations placed on the plant due to out of 

service equipment. Online maintenance should be carefully evaluated, planned and 

executed to avoid undesirable conditions or transients, and to thereby ensure a 

conservative margin of core safety.

8.4 ASME Risk-Informed Code Cases

Over the past several years, ASME has developed a series of risk-informed Code Cases related

to testing pumps and valves that include risk-informed provisions, including the following 

examples:

• OMN-1, “Alternative Rules for Preservice and Inservice Testing of Certain MotorOperated Valve Assemblies in Light-Water Reactor Power Plants.”

• OMN-3, “Requirements for Safety Significance Categorization of Components Using 

Risk Insights for Inservice Testing of LWR Power Plants.”

• OMN-4, “Requirements for Risk Insights for Inservice Testing of Check Valves at LWR 

Plants.”

• OMN-7, “Alternative Requirements for Pump Testing.”

• OMN-11, “Motor-Operated Valve Risk-Based Inspection Code Case.”

• OMN-12, “Alternative Requirements for Inservice Testing Using Risk Insights for 

Pneumatically and Hydraulically Operated Valve Assemblies in Light-Water Reactor 

Power Plants.”

The Code Cases listed as approved in Tables 1 and 2 of RG 1.192 have been incorporated by

reference into 10 CFR 50.55a. Licensees may voluntarily use these Code Cases, without 

additional staff approval, as an alternative to complying with the ASME Code provisions that 

have been incorporated by reference into 10 CFR 50.55a, provided that the licensee uses the 

Code Cases with the conditions specified in RG 1.192 (i.e., the Code Case is generally 
A 8-4

acceptable, but the NRC staff has determined that the alternative requirements must be 

supplemented in order to provide an acceptable level of quality and safety).

When using ASME’s risk-informed Code Cases, licensees must perform the testing and 

performance monitoring of individual components as specified in the risk-informed component 

Code Cases (e.g., OMN-1, OMN-3, OMN-4, OMN-7, OMN-11, and OMN-12), as modified by

any conditions specified in RG 1.192.

The ASME OM, Subsection ISTE in the 2009 Edition to the ASME OM addresses the riskinformed inservice testing. The NRC staff will review Subsection ISTE as part of the 

incorporation by reference of the 2009 Edition to the ASME OM Code in 10 CFR 50.55a with 

applicable conditions (if any). Note: ASME OM Code 2009 Edit details are provided for 

information only.
A 9-1

9. HISTORICAL REFERENCES

9.1 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section 

III, Division 1, Subsection NC, “Class 2 Components,” New York, 1986.

9.2 ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection ND, “Class 3

Components,” New York, 1986.

9.3 ASME Boiler and Pressure Vessel Code, New York

9.3.1 Inquiry IN 91-3, “OM Code 1987 through OMc-1990, Part 6, Para 4.6.1.1.”

9.3.2 Inquiry IN 91-037, “ASME Section XI, 1977 Edition through Later Editions and Addenda 

through 1987 Addenda, Table IWP-4110-1, ‘Instrument Accuracy Flowrate’.”

9.3.3 Inquiry IN 91-045, “Section XI, IWV-3200; Valve Stroke-Time Test,” March 10, 1992.

9.3.4 Inquiry IN 92-025A, “Section XI, IWV-3410 and IWV-3520; Valve Testing - Extended 

Shutdown,” February 9, 1993.

9.3.5 Inquiry IN 92-031, “Section XI, IWA-3200; Valve Testing - Adjustment of Packing,”

August 27, 1992.

9.3.6 Interpretation XI-78-01.

9.3.7 Interpretation XI-1-79-19, “Section XI, Division 1, Operability Limits of Pumps, IWP3210,” File Number BC-79-150, December 12, 1979.

9.3.8 Interpretation XI-1-89-10, “Section XI, Division 1, IWV-3300, IWV-3412, and IWV-3413; 

Valve Exercising Test,” File Number IN88-015, November 14, 1988.

9.3.9 Interpretation XI-1-89-55, “Section XI, IWP-4110, Table IWP-4110-1, and IWP-4120; 

Pump Instrument Accuracy,” File Number IN90-021, January 9, 1991.

9.3.10 Interpretation XI-1-89-65, “Section XI, IWV-3512 and IWV-3514; Valve Set Point 

Adjustments,” File Number IN90-002, January 15, 1991.

9.4 Performance Test Code, PTC 25.3-1976, “Safety and Relief Valves,” New York, 1976.

9.5 American Society of Mechanical Engineers/American National Standards Institute 

(ASME/ANSI), Operations and Maintenance Standards, New York, 1987.

9.5.1 Part 1 (OM-1), “Requirements for Inservice Performance Testing of Nuclear Power Plant 

Pressure Relief Devices,” 1981 and 1987.

9.5.2 Part 6 (OM-6),” Inservice Testing of Pumps in Light-Water Reactor Power Plants,” 1988

and 1989 Addenda.

9.5.3 Part 10 (OM-10), “Inservice Testing of Valves in Light-Water Reactor Power Plants,”

1988 Addenda.
A 9-2

9.6 American Society of Mechanical Engineers/American National Standards Institute 

(ASME/ANSI), Code for Operation and Maintenance of Nuclear Power Plants, New 

York, 1990.

9.6.1 Interpretation 92-2, “OM-1-1981, Paragraphs 1.3.3.1.5, 1.3.4.1.5, and 1.3.1.3; 

Adjustment of Valve Setpoint - Corrective Action,” File Number OMI-91-2, March 24, 

1992.

9.6.2 Interpretation 92-4, “OM-1-1981, Paragraphs 8.1.1.9 and 8.1.3.8; Set Pressure Testing,”

File Number OMI-91-3d, March 24, 1992.

9.6.3 Interpretation 92-5, “OM-1987 With Addenda Through OMc-1990, Part 1, Paragraphs

1.1.2, 7.1.2.3, 7.2.2.3, 7.3.2.4, and 7.4.2.4; Applicability – Class 2 and 3 Vacuum Relief 

Valves,” File Number OMI-91-4, March 24, 1992.

9.6.4 Interpretation 92-6, OM-1987 With Addenda Through OMc-1990, Part 6, Paragraph 5.2 

and Table 3a: Pump Testing, File Number OMI-91-5, March 24, 1992.

9.6.5 Interpretation 93-1, OM-1987 With OMa-1988 Addenda, Part 10, Paragraphs 4.2.1 and

4.3.2; OM Code-1990, ISTC 4.2 and ISTC 4.5; Valve Testing During Extended

Shutdown, File Number OMI-92-4, January 7, 1993.

9.6.6 Interpretation 95-7, OM Code-1990, ISTB 4.6.1 and Table ISTB 4.6.1.1.; OM-1987 with

OMa-1988, Part 6, Para. 4.6.11 and Table 1, Instrument Accuracy. 

9.7 American Society of Mechanical Engineers/American National Standards Institute 

(ASME/ANSI), Code for Operation and Maintenance of Nuclear Power Plants, New 

York, 1995 Edition through 2004 Edition including 2005 and 2006 addendas.

9.8 American Society of Mechanical Engineers/American National Standards Institute 

(ASME/ANSI), Standard N45.2.6, Qualification of Inspection and Examination and 

Testing Personnel for Nuclear Power Plants. 

9.9 Nuclear Energy Institute (NEI), White Paper, “Standard Format for Requests from

Commercial Reactor Licensees Pursuant to 10 CFR 50.55a,” Rev 1 June 2004.

9.10 Nuclear Management and Resources Council (NUMARC, now NEI), NUMARC 93-01, 

Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power 

Plants, Revision 2, April 1996.

9.11 Oak Ridge National Laboratory, Utility Survey PWR Safety Injection Accumulator Tank 

Discharge Check Valve Testing, ORNL/NRC/LTR-94/04, February 17, 1994.

9.12 U.S. Nuclear Regulatory Commission, Bulletin 88-04, Potential Safety-Related Pump 

Loss, Washington, DC, May 5, 1988.

9.13 U.S. Code of Federal Regulations, Title 10, Energy, Chapter 1, Part 50, Domestic 

Licensing of Production and Utilization Facilities. 
A 9-3

9.13.1 Federal Register, Vol. 41, No. 30, “Codes and Standards for Nuclear Power Plants (10

CFR Part 50),” February 12, 1976, p. 6256.

9.13.2 Federal Register, Vol. 57, No. 3152, “Codes and Standards for Nuclear Power Plants 

(10 CFR Part 50),” August 6, 1992, p. 34666.

9.13.3 Federal Register Notice (56 FR 36175), July 31, 1991.

9.13.4 Federal Register Notice (64 FR 51370), September 22, 1999.

9.13.5 Federal Register Notice (66 FR 40626), August 3, 2001.

9.13.6 Federal Register Notice (69 FR 58804), October 1, 2004.

9.14 Generic Issues and Generic Letters

9.14.1 “Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant 

Accidents in GE-Designed Operating Plants and Near-Term Operating Licensee 

Applications,” January 1980.

9.14.2 “Generic Issue 105, “Interfacing Systems LOCAs at Light-Water Reactors.”

9.14.3 Generic Letter 87-06, Periodic Verification of Leak-Tight Integrity of Pressure Isolation 

Valves, March 13, 1987.

9.14.4 Generic Letter 87-09, Sections 3.0 and 4.0 of the Standard Technical Specifications 

(STS) on the Applicability of Limiting Conditions for Operation and Surveillance 

Requirements, May 4, 1987.

9.14.5 Generic Letter 88-14, Instrument Air Supply System Problems Affecting Safety-Related 

Equipment, August 8, 1988.

9.14.6 Generic Letter 89-04, Guidance on Developing Acceptable Inservice Testing Programs, 

April 3, 1989.

9.14.7 Generic Letter 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance, 

June 28, 1989.

9.14.8 Generic Letter 90-06, Resolution of Generic Issue 70, Power-Operated Relief Valve and 

Block Valve Reliability, and Generic Issue 94, Additional Low-Temperature 

Overpressure Protection for Light-Water Reactors, Pursuant to 10 CFR 50.54(f), June 

20, 1990.

9.14.9 Generic Letter 91-18, Information to Licensees Regarding NRC Inspection Manual 

Sections on Resolution of Degraded and Nonconforming Conditions, Revision 1, 

October 8, 1997 (GL 91-18 superseded see RIS 2005-20).

9.14.10 Generic Letter 96-05, Periodic Verification of Design-Basis Capability of Safety-Related 

Motor-Operated Valves, September 18, 1996.

9.14.11 Generic Letter 96-06, Assurance of Equipment Operability and Containment Integrity 

During Design-Basis Accident Conditions, September 30, 1996.
A 9-4

9.14.12 Generic Safety Issue 158, Performance of Safety-Related Power-Operated Valves 

Under Design-Basis Conditions. 

9.15 Regulatory Information Summary (RIS) 2000-03, Resolution of GSI 158 Performance of

Safety-Related Valves Under Design Basis Conditions, March 15, 2000, including NRC 

staff comments on JOG Program on AOV periodic testing.

9.16 Information Notices

9.16.1 Information Notice 82-08, Check Valve Failures on Diesel Generator Engine Cooling

Systems, March 26, 1982.

9.16.2 Bulletin No. 83-03, Check Valve Failures in Raw Water Cooling System of Diesel 

Generators, March 10, 1983.

9.16.3 Information Notice 83-54, Common Mode Failure of Main Steam Isolation Non-Return

Check Valves, August 11, 1983.

9.16.4 Information Notice 85-84, Inadequate Inservice Testing of Main Steam Isolation Valves, 

October 30, 1985.

9.16.5 Information Notice 86-50, Inadequate Testing to Detect Failures of Safety-Related 

Pneumatic Components or Systems, June 18, 1986.

9.16.6 Information Notice 87-01, RHR [Residual Heat Removal] Valve Misalignment causes

Degradation of ECCS in PWRs, January 6, 1987.

9.16.7 Information Notice 87-40, Back Seating Valves Routinely to Prevent Packing Leakage, 

August 31, 1987.

9.16.8 Information Notice 88-70, Check Valve Inservice Testing Program Deficiencies, August

29, 1988.

9.16.9 Information Notice 89-32, Surveillance Testing of Low-Temperature OverpressureProtection Systems, March 23, 1989.

9.16.10 Information Notice 89-62, Malfunction of Borg-Warner Pressure Seal Bonnet Check

Valves Caused by Vertical Misalignment of Disk, August 31, 1989.

9.16.11 Information Notice 91-56, Potential Radioactive Leakage to Tank Vented to

Atmosphere, September 19, 1991.

9.16.12 Information Notice 91-74, Changes in Pressurizer Safety Valve Setpoints Before 

Installation, November 25, 1991.

9.16.13 Information Notice 94-08, Potential for Surveillance Testing To Fail To Detect an

Inoperable Main Steam Isolation Valve, February 1, 1994.

9.16.14 Information Notice 94-44, Main Steam Isolation Valve Failure To Close on Demand 

Because of Inadequate Maintenance and Testing, June 16, 1994.

9.16.15 Information Notice 94-56, Inaccuracy of Safety Valve Set Pressure Determination Using 

Assist Devices, August 11, 1994.

9.16.16 Information Notice 96-02, Inoperability of PORVs Masked by Downstream Indications 

During Testing, January 5, 1996.

9.16.17 Information Notice 96-48, Motor-Operated Valve Performance Issues, August 21, 1996.

9.16.18 Information Notice 97-16, Pre-Conditioning of Plant Structures, Systems, and 

Components before ASME Code Inservice Testing or Technical Specification 

Surveillance Testing, April 4, 1997.

9.16.19 Information Notice 97-90, Use of Non-Conservative Acceptance Criteria in Safety 

Related Pump Surveillance Tests, December 30, 1997.
A 9-5

9.16.20 Information Notice 2000-21, Detached Check Valve Disk Not Detected by the Use of

Acoustic and Magnetic Non-Intrusive Test Technique, December 15, 2000.

9.16.21 Information Notice 2003-01, Failure of a Boiling Water Reactor Main Steam 

Safety/Relief Valve, January 15, 2003.

9.17 Regulatory Information Summary 2004-12, Clarification on Use of Later Editions and 

Addenda to the OM Code and Section XI, July 28, 2004.

9.18 RIS 2005-20, “Revision to NRC Inspection Manual Part 9900 Guidance, “Operability 

Determinations & Functionality Assessments for Resolution of Degraded or 

Nonconforming Conditions Adverse to Quality and Safety,” Revision 1, dated April 16, 

2008.

9.19 Inspection Manual (sections published periodically).

9.19.1 Part 9900, “Technical Guidance: Maintenance - Voluntary Entry into Limiting Conditions 

for Operation Action Statements To Perform Preventive Maintenance,” April 18, 1991

9.19.2 Part 9900, “Technical Guidance: Maintenance - Preconditioning of Structures, Systems, 

and Components Before Determining Operability,

9.19.3 Inspection Procedure (IP) 61726, “Surveillance Observations,”

9.19.4 Inspection Procedure (IP) 62707, “Maintenance Observation,”

9.19.5 Inspection Procedure (IP) 62708, “Motor-Operated Valve Capability,”

9.19.6 Attachment 22, “Surveillance Testing,” to NRC Inspection Manual IP 71111, “Reactor 

Safety: Initiating Events, Mitigating Systems, Barrier Integrity,”

9.19.7 Inspection Procedure (IP) 73756, “Inservice Testing of Pumps and Valves,” July 27, 

1995,

9.20 Letters

9.20.1 Letter from W.T. Russell to R.K. Buckles, September 15, 1994.

9.20.2 Memorandum from Frederick J. Hebdon, Director, Project Directorate II-3, Division of

Reactor Projects I/II, NRC Office of Nuclear Reactor Regulation, to Jon R. Johnson, 

Acting Director, Division of Reactor Projects, Region II, in response to Technical 

Assistance Request TIA 96-007: Regulatory Acceptability of Lubricating Valves Prior to 

Surveillance Testing, dated July 2, 1996 [ADAMS Accession No. 9607150019]

9.20.3 Memorandum from E. Beckjord to J. Taylor, Technical Resolution of Generic Issue 105 

(GI-105), Interfacing Systems Loss-of-Coolant Accident (ISLOCA) in LWRs, June 3, 

1993.
A 9-6

9.20.4 Minutes of the Public Meetings on Generic Letter 89-04, October 25, 1989., 9.21

NUREGs:

9.21.1 NUREG-0737, Clarification of TMI Action Plan Requirements, November 1980.

9.21.2 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for

Nuclear Power Plants.

9.21.3 NUREG-1275, Operating Experience Feedback Report: Solenoid-Operated Valve 

Problems, Vol. 6, February 1991.

9.21.4 NUREG-1350, NRC Information Digest, issued annually.

9.21.5 NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, Revision 0 

(April 1995) and Revision 1 (January 2005).

9.22 NUREG/CPs

9.22.1 NUREG/CP-0111, Proceedings of the Symposium on Inservice Testing of Pumps and

Valves, October 1990.

9.22.2 NUREG/CP-0123, Proceedings of the Second NRC/ASME Symposium on Valve and

Pump Testing, July 1992.

9.22.3 NUREG/CP-0123, Proceedings of the Second NRC/ASME Symposium on Valve and

Pump Testing, Supplement 1, November 1992.

9.22.4 NUREG/CP-0137, Proceedings of the Third NRC/ASME Symposium on Valve and

Pump Testing, July 1994.

9.22.5 NUREG/CP-0152, Proceedings of the Seventh NRC/ASME Symposium on Valve and

Pump Testing, July 2002.

9.23 NUREG/CRs

9.23.1 NUREG/CR-5102, Interfacing Systems LOCA: Pressurized-Water Reactors, February

1989.

9.23.2 NUREG/CR-5124, Interfacing Systems LOCA: Boiling-Water Reactors, February 1989.

9.23.3 NUREG/CR-5775, Quantitative Evaluation of Surveillance Test Intervals Including TestCaused Risks, March 1992.

9.23.4 NUREG/CR-6644, Generic Issue 158: Performance of Safety-Related Power-Operated 

Valves Under Operating Conditions.
A 9-7

9.23.5 NUREG/CR-6654, A Study of Air-Operated Valves in U.S. Nuclear Power Plants.

9.24 Regulatory Guides (RGs)

9.24.1 Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-, Steam-, 

and Radioactive-Waste-Containing Components of Nuclear Power Plants, Revision 4, 

March, 2007.

9.24.2 Regulatory Guide 1.58, Qualification of Nuclear Power Plant Inspection, Examination, 

and Testing Personnel, Revision 1, September 1980. (Withdrawn -- See 56 FR 36175, 

07/31/1991)

9.24.3 Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability ASME Section XI 

Division 1, Revision 16, October 2010.

9.24.4 Regulatory Guide 1.174, An Approach for using Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, May 

2011.

9.24.5 Regulatory Guide 1.175, An Approach for Plant-Specific, Risk-Informed DecisionMaking: Inservice Testing, August 1998.

9.24.6 Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability: ASME

OM Code, June 2003.

9.24.6 Regulatory Guide 1.193, Revision 3, ASME Code Cases Not Approved for Use, October, 

2010.

9.25 Safety Evaluation dated September 2006 and its supplement dated September 2008 on 

NRC staff review of JOG Program on MOV Periodic Verification.

9.26 SECY-77-439, Single-Failure Criterion, August 17, 1977.

9.27 SECY-92-223, Resolution of Deviations Identified during the Systematic Evaluation 

Program, June 19, 1992.

9.28 Summary of the NRC’s Public Workshops on the Revision of NRC Inspection Procedure 

73756, dated July 18, 1997.

9.29 Supplement to Minutes of the Public Meetings on Generic Letter 89-04, September 16, 

1991.

9.30 WASH-1400 (NUREG-75/014), Reactor Safety Study: An Assessment of Accident Risk 

in U.S. Commercial Nuclear Power Plants, 1975.

9.31 American Society of Mechanical Engineers Operation and Maintenance Code, Rules for 

Inservice Testing of Light-Water Reactor Power Plants. 2004 Edition with 2005 and 2006 

Addendas.
B-i

Draft NUREG 1482, Revision 2, 

APPENDIX B 

Guidelines for Inservice Examination and Testing Program for

Dynamic Restraints (Snubbers) at Nuclear Power Plants
B-ii

CONTENTS

APPENDIX B

Page

ABBREVIATIONS B-iii

1 INTRODUCTION B 1-1

1.1 Regulatory Basis B 1-1

1.2 Regulatory History B 1-3

1.3 NRC Recommendations and Guidance B 1-4

1.4 Synopsis of Report B 1-5

2 DEVELOPING AND IMPLEMENTING AN INSERVICE EXAMINATION 

AND TESTING PROGRAM OF SNUBBERS B 2-1

2.1 Compliance Considerations B 2-1

2.1.1 ASME Code Case Applicability B 2-2

2.2 Scope of Inservice Testing Programs B 2-3

2.2.1 Basis for Scope Requirements B 2-4

2.2.2 Snubbers attached to Steam Generator and 

Reactor Coolant Pumps B 2-5

2.2.3 Testing of Non-Code Snubbers B 2-6

2.3 Code Class Systems Containing Safety-Related Pumps and Valves B 2-6

2.4 Snubbers Inservice Examination and Testing Program and their

Documentation B 2-7

2.4.1 Snubber Program while using ASME B&PV Section XI Code B 2-7

2.4.2 Snubber Program while using ASME OM Code B 2-8

2.4.3 Snubber Program while using NRC Authorized Alternative Relief B 2-8

2.4.4 Snubber Programs and their Bases B 2-8

2.4.5 Snubber List or Snubbers Controlled Data Base B 2-10

2.4.6 Snubber Programs Documentation and their Submittal to NRC B 2-11

2.5 Relief Requests and Proposed Alternatives B 2-12

2.6 Snubber Program Documents B 2-13

2.7 Repair and Replacement of Snubbers B 2-14

2.8 ISI of the Integral and Non-Integral Attachments of

Supports Containing Snubbers B 2-14

2.9 Developing Snubber Program for New Nuclear Power Plants B 2-15

3 REFERENCES B 3-1

Tables

2.1 Typical Systems Where Snubbers are Included in the Inservice Examination 

and Testing Snubber Program for a Pressurized-Water Reactor B 2-16

2.2 Typical Systems Where Snubbers are Included in the Inservice Examination 

And Testing Snubber Program for a Boiling-Water Reactor B 2-17

Figures

2.1 Flow Chart – Development of Inservice Examination and Testing Program

for Snubbers B 2-18
B-iii

ABBREVIATIONS

ASME American Society of Mechanical Engineers

B&PV Boiler and Pressure Vessel

BWR boiling-water reactor

CFR Code of Federal Regulations

DBD design-basis document

FR Federal Register

FSAR final safety analysis report

GDC General Design Criterion

GL generic letter

GSI generic safety issue

IN information notice

IP inspection procedure

ISI inservice inspection

IST inservice testing

LBHS Large Bore Hydraulic Snubbers

NEI Nuclear Energy Institute (formerly NUMARC)

NRC U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation (NRC)

NUMARC Nuclear Management and Resources Council (now NEI)

OM Operations and Maintenance

P&ID piping and instrument diagram

PWR Pressurized-water reactor

RCP Reactor Coolant Pump

RG regulatory guide

RIS Regulatory Issue Summary

SAR safety analysis report

SG Steam Generator

STS Standard Technical Specifications

TS Technical Specification(s)

TRM Technical Requirement Manual

UFSAR updated final safety analysis report

WGC Working Group Committee 
B 1-1

1. INTRODUCTION

1.1 Regulatory Basis

Title 10, Section 50.55a, of the Code of Federal Regulations (10 CFR 50.55a) defines the 

requirements for applying industry codes and standards to boiling- or pressurized-water-cooled 

nuclear power facilities. Each of these facilities is subject to the conditions in paragraphs (a), 

(f), and (g) of 10 CFR 50.55a, as they relate to inservice inspection (ISI) and inservice testing 

(IST). By rulemaking effective September 8, 1992 (see Federal Register, Vol. 57, No. 3152, p.

34666, dated August 6, 1992), the U.S. Nuclear Regulatory Commission (NRC) established 

paragraph (f) of 10 CFR 50.55a to separate the IST requirements from the ISI requirements in 

paragraph (g).

The NRC regulations in 10 CFR 50.55a(b)(2) incorporate by reference the 1970 Edition through 

the 2007 Edition with 2008 addenda of the American Society of Mechanical Engineers (ASME) 

Boiler and Pressure Vessel Code (B&PV), Section XI, in which Article IWF-5000 specify the 

requirements of inservice examination and testing of snubbers. The NRC regulations in 10 CFR

50.55a(b)(3) incorporate by reference the 1998 Edition through the 2004 Edition with 2005 and 

2006 addendas of the Code for Operation and Maintenance of Nuclear Power Plants (OM

Code) promulgated by the ASME, in which Subsection ISTA provides general IST and 

examination requirements and Subsections ISTB, ISTC, and ISTD specify the IST requirements 

for pumps, valves, and dynamic restraints, respectively. Based on those requirements, each of 

the NRC’s nuclear power plant licensees must establish IST programs, specify the components 

included in the program as well as the test methods and frequencies for those components, and 

implement the program in accordance with the OM Code.

The regulations in 10 CFR 50.55a(g)(6)(i) specify that if an inservice examination and test 

requirement of the ASME B&PV Code is determined to be impractical for a facility, the NRC’s 

regulations allow the licensee to submit a request for relief from the given requirement, along 

with information to support the determination. Relief requests generally detail the reasons for 

deviating from the Code requirements and propose alternative testing methods or frequencies. 

The Commission is authorized to evaluate licensees’ relief requests, and may grant the 

requested relief or impose alternative requirements, considering the burden that the licensee 

might incur if the Code requirements were enforced for the given facility. Pursuant to 10 CFR

50.55a(a)(3)(i) and (ii), the Commission may also authorize the licensee to implement an 

alternative to the Code requirements, provided that the alternative ensures an acceptable level 

of quality and safety or the Code requirement presents a hardship without a compensating 

increase in the level of quality and safety.

The regulations in 10 CFR 50.55a(g)(4)(iv) specify that inservice examination and testing of 

snubbers may meet the requirements in editions and addenda of the ASME B&PV Section XI or 

OM Code that were published more recently than those that are incorporated by reference in 10

CFR 50.55a(b), subject to Commission approval and the limitations and modifications listed in 

10 CFR 50.55a(b). Requests for approval to use later editions and addenda previously 

incorporated by reference in 10 CFR 50.55a may be made via letter to the NRC. For further 

clarification see NRC Regulatory Issue Summary (RIS) 2004-12, “Clarification on Use of Later 

Editions and Addenda to the ASME OM Code and Section XI.” 
B 1-2

10 CFR 50.55a(g) establishes the ISI requirements, including the effective edition and addenda 

of the American Society of Mechanical Engineer (ASME) Boiler and Pressure Vessel (B&PV) 

Code that licensees must use when performing ISI of components (including supports). 

10 CFR 50.55a(g)(4) states, “Throughout the service life of a boiling or pressurized watercooled nuclear power facility, components (including supports) which are classified as ASME 

Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access 

provisions and preservice examination requirements, set forth in Section XI of editions of the 

ASME B&PV Code and addenda.” ASME B&PV Section XI provides the rules for ISI of nuclear 

power plant components.

10 CFR 50.55a(g)(4)(ii) requires the use of the Code latest edition and addenda that has been 

incorporated by reference 12 months prior to the beginning of each 120-month interval. This 

Code is considered to be the “Code of Record” for the inspection interval.

10 CFR 50.55a(g)(4)(iv) requires that ISI of components (including supports) may meet the 

requirements set forth in subsequent editions to the “Code of Record” and addenda that are 

incorporated by reference in 10 CFR 50.55a(b), subject to limitations and modifications listed in 

10 CFR 50.55a(b) and subject to Commission approval. 

The ISI of ASME Code Class 1, 2, and 3 components including snubbers shall be performed in 

accordance with Section XI, “Rules for Inservice Inspection of Nuclear Power Plant 

Components,” of the ASME B&PV Code and applicable addenda as required by 10 CFR 

50.55a(g), except where specific written relief has been granted by the NRC or the Commission, 

pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states that alternatives to the 

requirements of paragraph (g) may be used, when authorized by the NRC, if: (i) the proposed 

alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the 

specified requirements would result in hardship or unusual difficulty without a compensating 

increase in the level of quality and safety.

10 CFR 50.55a(b)(3)(v) allows the optional use of Subsection ISTD, “Preservice and Inservice 

Examination and Testing of Snubbers in Light-Water Reactor Nuclear Power Plants,” of the 

ASME OM Code-1995 Edition through the latest edition and addenda incorporated by reference 

in 10 CFR 50.55a(b)(3) in lieu of ASME B&PV Section XI, Articles IWF-5200(a) and (b) and 

IWF-5300(a) and (b) by making appropriate changes to technical specification (TS) or licenseecontrolled documents. The regulation at 10 CFR 50.55a(b)(3)(v) also states, “Preservice and 

inservice examination must be performed using the VT-3 visual examination method described 

in IWA-2213.” 
B 1-3

1.2 Regulatory History

Since the start of commercial operation of the nuclear power plants, inservice visual inspection 

and functional testing of snubbers have been regulatory requirements. Originally these 

requirements were imposed by the plant TS surveillance requirements. There are two inservice 

surveillance requirements of snubbers (1) Visual inspection or examination; and (2) Functional 

or operational testing. The visual inspection is the observation of the condition of installed 

snubbers to identify those that are damaged, degraded, or inoperable as caused by physical 

means, leakage, corrosion, or environmental exposure. Functional testing is to verify that a 

snubber can operate within specific performance limits. The snubber functional test is 

performed to check its operational-readiness. The functional test typically involves removing the 

snubber and testing it on a specially-designed test stand. The performance of visual 

examinations is a separate process that complements the functional testing program and 

provides additional confidence in snubber population reliability. The TS specifies a schedule for 

snubber visual inspections and functional testing, which is usually based on refueling outage 

intervals.

Improved TS for various boiling and pressurized water-cooled nuclear power plants (NUREG1430 thru 1434, Revision 3) allow relocating inservice examination and testing requirements of 

snubbers from the TS to a plant’s Technical Requirements Manual (TRM). Relocating snubber 

ISI and testing requirements from the TS to TRM however, does not eliminate the need to 

comply with the 10 CFR 50.55a requirements.

In 1990, the NRC issued Generic Letter 90-09, “Alternate Requirements for Snubber Visual 

Inspection Intervals and Corrective actions.” The alternative visual inspection schedule is based 

on the number of unacceptable snubbers found during the previous inspection interval in 

proportion to the sizes of the various snubber populations or categories. This reduces future 

occupational radiation exposure and is highly cost effective. The alternative inspection interval 

is based on a fuel cycle of up to 24 months.

At that time most of the licensees revised their snubbers examination and testing documents 

such as TS, TRM or licensee-controlled documents, based on GL 90-09. GL 90-09, provides an 

alternative to requirements for snubber visual inspection intervals only. .

Until 1990, the ASME Code requirements addressing inservice examination and testing of 

snubbers were only contained in ASME Section XI, Article IWF-5000. IWF-5000 referenced 

ASME/American National Standards Institute Standard for Operation and Maintenance of 

Nuclear Power Plants, Part 4 (OM-4), 1987 Edition with OMa-1988 Addenda for the preservice 

and inservice examinations and testing of snubbers. In 1990, the ASME published the initial 

edition of the ASME OM Code which provides rules for inservice examination and testing of 

snubbers by incorporating most of the requirements of OM-4. In the 1995 Edition of the ASME 

OM Code incorporated all the visual inspection requirements provided in GL 90-09.

The 1990 Edition of the ASME OM Code consisted of one section (Section IST) entitled “Rules 

for Inservice Testing of Light-Water Reactor Power Plants.”' This section is divided into four 

subsections: ISTA, “General Requirements,”' ISTB, “Inservice Testing of Pumps in Light-Water 

Reactor Power Plants,” ISTC, “Inservice Testing of Valves in Light-Water Reactor Power 

Plants,” and ISTD, “Examination and Performance Testing of Nuclear Power Plant Dynamic 

Restraints (Snubbers).” At that time, the inservice examination and testing of snubbers was
B 1-4

governed (under rule making) by the ISI requirements of Section XI of the ASME B&PV Code1989 Edition. Therefore, the Subsection ISTD was not incorporated by reference in 10 CFR 

50.55a in rule making at that time.

In 2000, for the first time, the proposed rule 10 CFR 50.55a(b)(3)(v)) stated that licensees may 

use the guidance in Subsection ISTD, ASME OM Code, 1995 Edition with the 1996 Addenda, 

for examination and testing of snubbers. The current regulation at 10 CFR 50.55a(b)(3)(v) 

allows the optional use of Subsection ISTD, “Preservice and Inservice Examination and Testing 

of Dynamic Restraint (Snubbers) in Light-Water Reactor Nuclear Power Plants,” of the ASME 

OM Code-1995 Edition through the latest edition and addenda incorporated by reference in 

10 CFR 50.55a(b)(3) in lieu of ASME B&PV Section XI, Articles IWF-5200(a) and (b) and IWF5300(a) and (b) by making appropriate changes to TS or licensee-controlled documents. The 

regulation at 10 CFR 50.55a(b)(3)(v) also states, “Preservice and inservice examination must be 

performed using the VT-3 visual examination method described in IWA-2213.”

Some commenters proposed Subsection ISTD as an acceptable alternative to all preservice 

and inservice examination requirements in IWF-5000, Section XI. The NRC has not accepted 

this suggestion because some preservice and inservice examinations for snubbers are not 

included in the OM Code. For example, Subsection ISTD does not address the scope of IWF5200(c) and IWF-5300(c), inspection of integral and non-integral attachments, such as lugs, 

bolting, pins, and clamps. Further, Subsection ISTD does not address snubbers in systems 

required to maintain the integrity of reactor coolant pressure boundary. 

The existing “Codes of Record” for this NUREG revision are 2004 Edition with 2005 and 2006 

addendas of the ASME Code for Operation and Maintenance of Nuclear Power Plants and 2007 

Edition with 2008 addenda of the ASME B&PV Code, Section XI. 

Snubber inservice inspection provisions are specified in the editions and addenda of Section XI 

up through the 2005 Addenda. Snubber inservice inspection provisions were removed from 

Section XI in the 2006 Addenda. Snubber inservice inspection provisions are also located in 

Subsection ISTD of the ASME OM Code, and 10 CFR 50.55a(b)(3)(v) allows licensees the 

option of using the inservice inspection provisions for snubbers in Section XI or the ASME OM 

Code. However, ASME B&PV Section XI option will no longer exist when using the 2006 

addenda and later editions and addenda of Section XI because these editions and addenda of 

Section XI do not provide inservice inspection provisions for snubbers. 

1.3 NRC Recommendations and Guidance

The recommendations herein supplement the guidance provided in the ASME B&PV Section XI 

and ASME OM Code for inservice inspection and testing of snubbers. This document is written 

for the latest edition of the ASME OM Code and ASME B&PV Section XI incorporated into 

Paragraph (b) of 10 CFR 50.55a. To the extent practical, this document reflects the applicable 

section, subsection, or paragraph of the applicable documents (subsection of 10 CFR 50.55a, 

OM Code ASME B&PV Section XI, regulatory guides, etc.).

The guidance presented herein may be applied when requesting to use alternative (relief) in lieu 

of the Code requirements. However, licenses may also request relief that is not in conformance
B 1-5

with the guidance. The NRC may reference a recommendation contained in this document in

future safety evaluations and may grant relief or authorized an alternative if the licensee has 

addressed all the aspects included in the respective section, where applicable.

This document specifically discusses applicable portions of Article IWA-1000 and IWA-2000, 

IWA-4000, IWF-5000, and IWF-6000 of the ASME B&PV Code, Section XI and Subsections 

ISTA and ISTD, Nonmandatory Appendix A with Supplement 3, B, C, D, E, F, G, and H of the 

ASME OM Code, which licensees may implement pursuant to 10 CFR 50.55a(g)(4)(iv). It also 

gives the requisite approval for licensees to use 10 CFR 50.55a(g)(4)(iv) in updating their 

snubber ISI and testing program to the requirements of the ASME B&PV Code, Section XI or 

OM Code as applicable.

If the licensee chooses to implement the guidance contained herein for issues approved under 

10 CFR 50.55a(g)(4)(iv), any deviation from the guidance requires Commission approval.

1.4 Synopsis of Report

This appendix follows the format of a typical ISI and testing snubber program plan, including 

Development and Implementation, General Guidance, and , Code Noncompliance. 

Section 2, “Developing and Implementing an Inservice Examination and Testing Program of 

Snubbers,” describes existing inservice examination and testing requirements, discusses the 

scope of the snubber program, and describes guidance for presenting information in snubber 

programs. This also provides specific recommendations on snubber and large bore (SG and 

RCP) snubber related issues. 

Rather, these discussions are intended to clarify the existing requirements of the Code or the

regulations and, as such, they may provide recommendations to ensure that licensees continue 

to meet the Code and other regulatory requirements.
B 2-1

2. DEVELOPING AND IMPLEMENTING AN INSERVICE EXAMINATION AND TESTING 

PROGRAM OF SNUBBERS

Licensees may use the following guidance for developing and implementing snubber inservice 

examination and testing programs. This guidance supplements existing requirements and 

previously approved guidance on inservice examination and testing. 

2.1 Compliance Considerations

The NRC regulations in Section 50.55a states that ISI of components (including supports) which 

are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except 

design and access provisions and preservice examination requirements, set forth in Section XI 

of editions of ASME B&PV Code and addenda. ASME B&PV Code, Section XI provides the 

rules for ISI of nuclear power plant components. 10 CFR 50.55a requires ISI of components 

(including supports) without specifically mentioning “snubbers.” 10 CFR 50.55a also allows the 

optional use of Subsection ISTD of the OM Code promulgated by the ASME for inservice 

inspection and testing of snubbers. This ISI along with testing is intended to assess the 

reliability and operational readiness of the snubbers. 

10 CFR 50.55a(g)(4)(ii) requires the use of the latest edition and addenda of the Code that has 

been incorporated by reference 12 months prior to the beginning of each 120-month interval. 

This Code is considered to be the “Code of Record” for the inspection interval. 

10 CFR 50.55a(g)(4)(iv) allows that ISI of components (including supports) to meet the 

requirements set forth in subsequent editions to the “Code of Record” and addenda that are 

incorporated by reference in 10 CFR 50.55a(b), subject to limitations and modifications listed in 

10 CFR 50.55a(b) and subject to Commission approval. 

Request for approval to use later editions and addenda previously incorporated by reference in 

10 CFR 50.55a may be via letter to the NRC. See RIS 2004-12 for further clarification. 

However, pursuant to 10 CFR 50.55a(g)(4)(iv), licensees’ ISI and testing programs may meet 

the requirements of editions and addenda of the Code (or portions thereof) that are more recent 

than those incorporated in 10 CFR 50.55(b). When requesting to use editions and addenda of 

the ASME Code that have not been incorporated by reference, licensees must request 

authorization to use these later editions and addenda as an alternative to the regulations 

pursuant to 10 CFR 50.55a(a)(3). When licensees choose to use any or all portions of a revised 

edition, they must meet all related requirements of the respective editions or addenda, and such 

exceptions are subject to Commission approval in accordance with 10 CFR 50.55a(g)(4)(iv).

The NRC may authorize alternatives to Code testing requirements submitted as relief requests 

or in a similar format that includes a discussion of the requirements, a description of the

proposed alternative, and the justification for approval of the alternative. 10 CFR 50.55a 

includes the following provisions for accepting alternatives or granting relief:

• 10 CFR 50.55a(a)(3)(i) allows the NRC to authorize alternatives if the proposed 

alternatives would provide an acceptable level of quality and safety. The NRC will 

normally approve an alternative pursuant to this provision only if the licensee proposes a 

method of testing that is equivalent to, or an improvement of, the method specified by 

the code, or if the testing will comply or is consistent with later Code editions approved 

by the NRC in 10 CFR 50.55a(b).
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• 10 CFR 50.55a(a)(3)(ii) allows the NRC to authorize an alternative if compliance [with 

the Code requirement] would result in hardship or unusual difficulty without a 

compensating increase in the level of quality and safety. The NRC may approve an 

alternative pursuant to this provision if, although the proposed alternative testing does 

not comply with the Code, the increase in overall plant safety and quality attained by 

complying with the Code requirement is not justified in light of the difficulty associated 

with compliance.

• 10 CFR 50.55a(f)(6)(i) includes the following provision:

The Commission will evaluate determinations that Code requirements are impractical. 

The Commission may grant relief and may impose such alternative requirements as it 

determines is authorized by law, giving due consideration to the burden upon the 

licensee that could result if the requirements were imposed on the facility.

The NRC may grant relief pursuant to this provision or may impose alternatives if the licensee 

demonstrates that the design or access limitations make the Code requirement impractical. 

Thus, the staff’s evaluation considers the burden created by imposing the Code requirements 

on the licensee.

For plants using their TS to govern ISI and testing of snubbers, 10 CFR 50.55a (g)(5)(ii) 

requires that if a revised ISI program for a facility conflicts with the TS, the licensee shall apply 

to the Commission for amendment of the TS to conform the TS to the revised program. 

Therefore, when performing 120-month ISI program updates in accordance with 10 CFR 50.55a 

(g)(4), licensees must submit any required amendments to ensure their TS remain consistent 

with the new code of record or NRC-approved alternative used in lieu of the Code requirements. 

The TS governing the snubber ISI and test program do not eliminate the 10 CFR50.55a 

requirements to update the program at 120-month intervals or to request and receive NRC 

authorization for alternatives to the Code requirements when appropriate.

2.1.1 ASME Code Case Applicability

Code Cases are typically formatted in terms of an Inquiry [as from a licensee] and Reply [by the 

applicable ASME Code Committee]. Oftentimes the Inquiry and Reply are accompanied by an 

Applicability statement that identifies the specific Editions an Addenda of the ASME Code to 

which the Reply applies. However, the ASME Code Committees do not always 1) include an 

Applicability statement or 2) identify all of the editions and addenda to which the Reply applies. 

As a result, several licensees have questioned whether they could use certain ASME Code 

Cases without additional interaction with the regulatory authority having jurisdiction at the plant 

site.

If a licensee would like to use an ASME Code Case with an Edition or Addendum of the ASME 

Code to which it is not applicable, the licensee has the following options:

a. Have the alternative to use the Code Case, beyond its stated applicability, authorized by 

the NRC pursuant to 10 CFR 50.55a(a)(3), or

b. If the Code Case is applicable to an Edition or Addendum of the ASME Code later than 

the version of the Code being used by the licensee, the licensee could update to the 

later version of the Code pursuant to 10 CFR 50.55a(f)(4)(iv) or (g)(4)(iv) and then use
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the Code Case, provided the Code Case has been approved for use in the appropriate 

Regulatory Guide and incorporated by reference into 10 CFR 50.55a. Note that the later 

version of the ASME Code must also have been incorporated by reference into 10 CFR 

50.55a, the licensee must update all related requirements of the respective Edition or 

Addenda, and the update must be specifically approved by the Commission.

Licensee should not use ASME Code Cases with Editions and Addenda of the ASME Code to 

which they do not apply and that are not specifically approved for use by the NRC. More 

specifically, licensees should not reconcile the Applicability of Code Cases without consulting 

with the applicable ASME Code Committee.

In incorporating the OM Code by reference in 10 CFR 50.55a, the NRC staff recognized the

need for a new regulatory guide that would approve OM Code cases. Such a regulatory guide 

would perform a function similar to that of existing Regulatory Guide (RG) 1.147, “Inservice 

Inspection Code Case Acceptability, ASME B&PV Code, Section XI, Division 1,” which 

approves ASME Code cases applicable to Section XI of the ASME B&PV Code. Accordingly, 

the NRC staff developed RG 1.192, “Operation and Maintenance Code Case Acceptability, 

ASME OM Code,” as well as RG 1.193, “ASME Code Cases Not Approved for Use.” 

If RG 1.193 identifies a Code case as being unacceptable, licensees may not implement the 

specified Code case without first obtaining NRC approval. Licensees may request the NRC’s 

approval to implement a Code case listed in the guide under the provisions of 10 CFR

50.55a(a)(3), which permits the use of alternatives to the Code requirements referenced in 10

CFR 50.55a, provided that the proposed alternative results in an acceptable level of quality and 

safety, by addressing the NRC’s concern and submitting a plant-specific relief request.

An ISI and testing snubber program, including implementing procedures, is subject to the 

requirements of 10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power 

Plants and Fuel Reprocessing Plants,” and ASME OM Code, Subsection ISTA or ASME B&PV 

Code, Section XI, Section IWA. Changes to the scope, test methods, or acceptance criteria 

should be reviewed to the requirements of 10 CFR 50.59, 10 CFR 50.55a and 10 CFR 50.65 as 

appropriate.

2.2 Scope of Inservice Examination and Testing Programs

The intent of General Design Criterion (GDC) 1, “Quality Standards and Records,” of Appendix 

A, “General Design Criteria for Nuclear Plants,” to the 10 CFR Part 50, “Licensing of Production 

and Utilization Facilities,” requires that all structures, systems, and components that are 

necessary for safe operation must be tested to demonstrate that they will perform satisfactorily 

in service. Among other things, GDC 1 requires that components that are important to safety 

must be tested to quality standards that are commensurate with the importance of the safety 

function(s) to be performed. Appendix B to 10 CFR Part 50 describes the requisite quality 

assurance program, which includes testing, for safety-related components. In addition, 

10 CFR 50.55a(g) requires that licensees must use the ASME B&PV Code, Section XI or the 

optional ASME OM Code for inservice inspection and testing of components that are covered by 

the Code. Each licensee has the responsibility to demonstrate the continued operability of all 

components within the scope of their ISI and testing program. The regulatory guides augment 

those requirements by providing additional NRC guidance regarding scope and classification. 

In short, the ASME Code defines the scope, 10 CFR 50.55a endorses the Code with 

clarifications, and regulatory guides provide additional guidance.
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2.2.1 Basis for Scope Requirements

The regulations at 10 CFR 50.55a(g) establish the ISI requirements that licensees must satisfy 

when performing ISI of components (including supports). Specifically, 10 CFR 50.55a(g)(4) 

states, “Throughout the service life of a boiling or pressurized water-cooled nuclear power 

facility, components (including supports) which are classified as ASME Code Class 1, Class 2, 

and Class 3 must meet the requirements, except design and access provisions and preservice 

examination requirements, set forth in Section XI of editions of the ASME B&PV Code and 

addenda.” ASME B&PV Code, Section XI provides the rules for ISI of nuclear power plant 

components.

The regulation at 10 CFR 50.55a(g)(4)(ii) requires the use of the latest edition and addenda of 

the Code that has been incorporated by reference 12 months prior to the beginning of each 

120 month inspection interval. This Code is considered the “Code of Record” for the inspection 

interval.

The regulation at 10 CFR 50.55a(g)(4)(iv) states that ISI of components (including supports) 

may meet the requirements set forth in subsequent editions to the “Code of Record” and 

addenda that are incorporated by reference in 10 CFR 50.55a(b), subject to limitations and 

modifications listed in 10 CFR 50.55a(b) and subject to Commission approval.

ASME Code Class 1, components includes all snubbers within the reactor coolant pressure 

boundary. Regulatory Guide (RG) 1.26, “Quality Group Classifications and Standards for 

Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants,” 

Revision 4, dated March 2007, provides guidelines for establishing quality standards for Quality 

Group B, C, and D (and ASME Code classification) for water-, steam-, and radioactive-wastecontaining components important to safety of water-cooled nuclear power plants, other than 

those in the reactor coolant pressure boundary (i.e., ASME Code Class 2 and 3 components). 

There are also systems of light-water-cooled reactors important to safety that are not identified 

in RG 1.26 for which there are established staff positions regarding quality group classification. 

These systems, and reference establishing their acceptable classification, are identified in 

Appendix A of Section 3.2.2, “System Quality Group Classification” of NUREG-800, “Standard 

Review Plan.” 

The ASME B&PV Code, Section XI and ASME OM Code are incorporated by reference in 

10 CFR 50.55a(b)(3). The ASME B&PV Code, Section XI as well as the optional ASME OM 

Code defines the scope by stating that ISI and testing programs shall include components 

(including snubbers) in systems that are required to perform a specific function in (1) shutting 

down the reactor to a safe shutdown condition, (2) maintaining the safe shutdown condition, or 

(3) mitigating the consequences of an accident. 

Subsection ISTA-1100 of the OM Code refers to components that are “needed to mitigate the 

consequences of an accident.” This statement is intended to provide confidence that the health 

and safety of the public will be protected in the event of certain accidents and anticipated 

transients at a nuclear power plant. The term “accident” is also used throughout the 

Commission’s regulations. For example, Appendix B to 10 CFR Part 50 establishes quality

assurance requirements for the design, construction, and operation of “structures, systems, and 

components that prevent or mitigate the consequences of postulated accidents that could cause 

undue risk to the health and safety of the public.” Similarly, 10 CFR Part 100 describes 
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structures, systems, and components that must be designed to remain functional during and 

following a “safe shutdown earthquake” as those necessary to ensure (1) the integrity of the 

reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in 

a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of 

accidents that could result in potential offsite exposures comparable to the guideline exposures.

In establishing such requirements, the Commission uses the term “accident” to describe a broad

range of possible adverse events at a nuclear power plant. Therefore, although most of the 

accidents of concern to inservice testing are addressed in the accident analyses chapter, 

licensees should be aware that the plant’s final safety analysis report (FSAR) may address 

other accident analyses that need to be considered within the context of inservice testing.

Thus, an introductory section of the inservice inspection and testing program document

submitted to the NRC for each plant must state the plant’s safe-shutdown condition (i.e., hot 

standby, hot shutdown, cold shutdown, etc.). If the scope in Section ISTA appears to be 

broader than that specified in 10 CFR 50.55a, the more narrow scope applies.

Components within the scope of 10 CFR 50.55a are included in the scope of 10 CFR 50.65, 

“Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants” (the 

Maintenance Rule). Licensees may elect to consolidate inspection and testing for snubbers, 

designating any non-Code components as such in the ISI and testing program.

The plant’s FSAR (or equivalent) defines the equipment that is necessary to meet specific 

functions. If the FSAR indicates that a system or component is Code Class 1, 2, or 3, that 

system or component is within the scope of 10 CFR 50.55a. By contrast, if the FSAR states 

that a system or component is designed, fabricated, and maintained as Code class at the option 

of the Owner as permitted by Subsection ISTA-1320, the application of the related OM Code

requirements is also optional.

Tables 2.1 and 2.2 (which appear at the end of this chapter) provide examples of systems with 

snubbers that licensees typically include in their ISI and testing snubber programs. These 

tables are not intended to be all-inclusive, but they may form the basis for the initial review of a 

licensee’s snubber program scope.

Figure 2.1, Flow Chart (which appears at the end of this chapter) provides a quick reference to 

regulatory requirements for development of the inservice examination and testing program for 

snubbers. For complete details, see 10 CFR 50.55a.

2.2.2 Snubber Attached to Steam Generator and Reactor Coolant Pumps

There are special requirements for PWR plants with regard to the testing of snubbers attached 

to the Steam Generators (SG) and Reactor Coolant pumps (RCP). These are generally large 

bore hydraulic snubbers (LBHS). Large bore hydraulic snubbers are defined as those units with 

rated capacities of 50 kips or greater. Unlike smaller hydraulic snubbers, LBHSs were exempt 

from inservice functional testing prior to 1980, primarily due to a lack of available test equipment 

of sufficient size. In 1980 and 1984, the NRC issued Generic Letters to all licensees requesting 

modification of plant TS to include LBHSs testing provisions. The results of initial tests revealed 
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numerous cases where LBHSs were either out of specified tolerance or completely inoperable. 

Subsequently, the NRC developed Generic Issue (GI-113), “Dynamic Qualification and Testing 

of Large Bore Hydraulic Snubbers (LBHSs),” with the objective of evaluating the reliability of 

LBHSs in operating commercial nuclear power plants. 

NUREG/CR-5416, “Technical Evaluation of Generic Issue 113: Dynamic Qualification and 

Testing of Large Bore Hydraulic Snubbers,” dated August 1992 provided major 

recommendations for LBHSs. This effort was in coordination with the industry, vendors, and 

snubber manufacturers. As a result, the NRC established specific inservice testing 

recommendations for LBHSs installed on PWR Steam Generators or Reactor Coolant Pumps. 

This recommendation was that these snubbers be tested as a separate test population, and it 

was eventually incorporated into the ASME OM Code. Although the generic Issue focused on 

hydraulic snubbers, the resultant requirement does not specify a snubber size or type to which it 

applies. Therefore, all the licensees are reminded that 10-year inservice examination and test 

snubber programs should include their steam generator snubbers and reactor coolant pumps 

snubbers, regardless of size or type. The Code requirement is that snubbers attached to steam 

generators and those attached to reactor coolant pumps be designated as at least one, 

separate Defined Test Plan Group (DTPG) for testing purposes as specified in ISTD-5353. 

Large bore snubbers (greater than 50 kips) located on piping or other may be included in the 

general snubber population for testing and examination purposes.

2.2.3 Testing of Non-Code Snubbers

An ISI and testing program is also a reasonable vehicle to periodically demonstrate the 

operability of non-safety related snubbers that are not covered by the Code. Thus, if a licensee 

chooses to include non-Code components in its ISI and testing snubber (or some other 

licensee-developed inspection and testing program) and, as a result, is unable to meet certain 

Code provisions, the regulations (10 CFR 50.55a) do not require the licensee to submit a relief 

request to the NRC. Nonetheless, the licensee should maintain documentation that provides 

assurance of the continued operability of the non-Code components through the performed 

tests, and such documentation should be available for staff inspection at the plant site.

Therefore, while 10 CFR 50.55a delineates the inspection and testing requirements for ASME 

Code Class 1, 2, and 3 snubbers, licensees should not limit their inservice inspection and 

testing to only those snubbers that are covered by 10 CFR 50.55a. 

Licensees may implement deviations from the Code for non-Code components without NRC

review and approval, and need not document such deviations as “relief requests.” Nonetheless, 

a notation in the licensee’s ISI and testing program document would help to identify the

deviations and clarify that they relate to non-Code components. If it is not clear that the

deviations relate to non-Code components, the staff might assume that the licensee is not

meeting the requirements of 10 CFR 50.55a. Some licensees use the relief request format to

document such deviations, while other licensees place notes, footnotes, or brief descriptions in

their program documents.

2.3 Code Class Systems Containing Safety-Related Snubbers

The plant safety analysis report (SAR), technical specifications (TSs), and other documents list 

the systems and components (i.e. snubbers) that must function to support the safe operation
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and shutdown of the plant. Tables 2.1 and 2.2 (which appear at the end of this chapter) list 

systems and components that are typically included in the ISI and testing of snubber programs 

for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). These tables are 

not intended to apply to all plants; the listed systems and components are not considered 

safety-related at every plant, and are not necessarily classified as Code Class 1, 2, or 3. (For 

information on quality group and Code classifications, see RG 1.26 and Section 3.9.6 

of.NUREG-0800.) The licensee’s safety analysis generally contains a section describing the 

Code classification of components. The snubber inservice examination and testing program 

scope must be consistent with the SAR.

2.4 Snubber Inservice Examination and Testing Programs and their Documentation 

10 CFR 50.55a(g)(4) states, in part, that throughout the service life of a boiling or pressurized 

water-cooled nuclear power facility, ASME Code Class 1, 2, and 3 components (including 

supports) meet the ISI and testing requirements of the ASME B&PV Code, Section XI or ASME 

OM Code as incorporated by reference in 10 CFR 50.55a(b). 

2.4.1 Snubber Program while using ASME B&PV Code, Section XI Code

Licensees using ASME B&PV Code, Section XI for its snubber inservice examination and 

testing program shall consider the following rules for inservice inspection and testing of 

snubbers:

• Subsection IWA addresses the general requirements for inservice examination and 

testing of snubbers.

• Article IWF-5000 addresses the inservice inspection and testing requirements for 

snubbers in addition to those required per Article IWF-1000. 

• Subarticle IWA-1400(c) requires that owner shall submit these plans and reports to the 

enforcement and regulatory authorities.

• Table IWA-1600-1 specifies the Edition 1987 with OMa-1988 for ASME/ANSI OM, Part 4 

(OM-4) to be use while using Article IWF-5000 for snubber inservice inspection and 

testing.

• Article IWA-4000 provides the requirements of Repair/Replacement activities including 

snubbers.

• Article IWA-6000 addresses the records and reports that are required for these 

inspection and testing programs of snubbers.

• Subarticle IWA-6210 states that the owner shall prepare plans for preservice and 

inservice examinations and testis to meet the requirements of the ASME B&PV Code, 

Section XI requirements. Article IWF-1000 addresses the scope and responsibility for 

inservice examination and testing of snubbers as a subset of components support 

examination.
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• ASME OM Code Subsections ISTA and ISTD may be used in lieu of the ASME Section 

XI, Article IWF-5000 as allowed by the regulation 10 CFR 50.55a(b)(v).

2.4.2 Snubber Program while using ASME OM Code

Licensees using ASME OM Code for its snubber inservice examination and testing program 

shall consider the following rules for inservice inspection and testing of snubbers:

• Subsection ISTA includes general requirements for inservice examination and testing of 

snubbers.

• Subsection ISTD addresses the “Preservice and Inservice Examination and Testing of 

Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants.

• Subsection ISTA-3200 states that ISI and testing plan shall be filed with the regulatory 

authorities.

• Subsection ISTA-9000 addresses the records and reports that are required for these 

inspection and testing programs.

• Subsection ISTA-9210 states that the owner shall prepare plans for preservice and 

inservice examinations and tests to meet the requirements of the OM Code.

• Subsection ISTA-9220 states that licensees shall prepare examination, test, 

replacement, and repair records in accordance with the requirements of the respective 

articles of the OM Code.

• Appendix A and its supplements to the OM Code describe voluntary guidance for 

licensees to use in preparing their inspection and test plans.

2.4.3 Snubber Program while using NRC Authorized Alternative Relief Request.

Licensees not using the ASME Section XI or ASME OM requirements for their snubber 

program, must submit relief request to NRC for approval to use alternative methods in lieu of the 

ASME Section XI and ASME OM requirements. NRC authorized relief to use TRMs or otherlicensee-controlled documents, in lieu of the ASME B&PV Code, Section XI or ASME OM Code 

requirements for inservice examination and testing of snubbers, do not provide relief from 

submitting snubber programs to the regulatory authorities. 

2.4.4 Snubber Programs and their Bases

Licensees are using TRMs or other licensee-controlled documents for snubber inservice 

examination and testing in lieu of the ASME B&PV Code, Section XI requirements. TRMs or 

other licensee-controlled documents serve as bases for snubber programs and most of the 

snubber programs have similarities across the industry. Many licensees are in the process of 

updating their snubber programs as required by the ASME OM Code. Some licensees have 

already updated their programs to use ASME OM Code. The updated snubber programs
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should contain at least the details and bases as documented in the TRM, or other licenseecontrolled documents in alignment with the ASME OM Code. Bases documents have typically 

included a description of the methodology used in preparing the snubber programs. The bases 

document should clearly state where a list of each snubbers is kept and how it is being 

maintained. Although not required by the regulation, the bases documents will help licensees 

ensure the continuity of their snubber programs when the responsibilities of personnel or groups 

change. A good bases document will also enable the plant staff to clearly understand the 

reasons that the snubbers are either in the program or not, as well as the basis for examination

and testing. The bases document can also serve as a useful reference for reviews performed 

under 10 CFR 50.59 when changes are made to a facility.

As a minimum the following details (three elements) with appropriate section numbers of the 

ASME Code should be included in a base snubber program bases document:

2.4.4.1 Inservice Visual Examination (description, definitions, and basis of each item)

1. Integral and nonintegral attachment to snubbers (see Section 2.8)

2. Initial Examination Intervals

3. Subsequent Examination Intervals

4. Method of Visual Examination

5. Inservice Examination Failure Evaluation

6. Code Case used for visual examination

2.4.4.2 Inservice Operability or Functional Test (description, definitions, and basis of 

each item)

1. Functional Test Frequency (every Refueling Outage)

2. Test Plan Groups [Defined Test Plan Group (DTPG)]

3. Sample Plans (10% Testing sample, 37 Testing sample, or 55 testing sample) 

used and Initial Snubber Sample size(s) for each DTPG 

4. Additional Snubber Sampling Method for each plan

5. Failure Evaluation requirements and methods

6. Test Failure Mode Groupings methodology 

7. Corrective Actions for each sample plan and FMG identified
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2.4.4.3 Service Life Monitoring Program

The licensees must develop the service-life monitoring (SLM) program as defined in the ASME 

OM Code, Subsection ISTD-6000 or TS or TRM or in accordance with an approved Relief 

Request. Nonmandatory Appendix F of ISTD provides additional guidance in developing a

Service Life Monitoring Program. The SLM program is the primary instrument for assuring 

continued reliability of a snubber population at a plant. The statistical method sample testing is 

one measure of program effectiveness but in general does not serve as an efficient tool to 

actually maintain or improve reliability. This is due to the fact that such testing is based on small 

samples (10% or 37 snubbers) every refueling outage, and in not predictive in nature. Based on 

snubber aging study process, in the NUREG/CR-5870, “Results of LWR Snubber Aging 

Research,” dated May 1992, the NRC recommended the Service Life Monitoring (SLM) of 

snubbers in addition to the statistical testing process. Most of the plants have included SLM in 

their TS, TRM etc. in the past. The ASME OM Code, Subsection ISTD also included SLM along 

with snubbers examination and inservice testing requirements. 

Some of the licensees have updated their snubber programs by using ASME OM Code 

requirements. The updated snubber programs simply reference plant procedures for snubber 

examinations and testing without providing any details about sections, subsection(s) and/or 

paragraphs, SLM requirements of the applicable ASME OM Code. Licensees should use the 

above guidelines as a minimum and consult with the Snubber User Group (SNUG), when 

developing guidance for snubber programs and their bases, to help ensure consistency 

throughout the industry. 

2.4.5 Snubber List or Snubber Controlled Data Bases

In preparing and maintaining a snubber list, licensees should consider the following information:

Reports generated to provide a snubber listing should include the suggested headings and a 

description of the text that licensees could include under each heading.

Title: List the applicable plant and unit.

Reference information: Include reference to method and location of controlled snubber data. 

This may be a controlled station data base from which reports and lists are generated as 

needed.

Program/Report revision or revision date: List the revision number and date/or date (on each 

page). 

System, Code class, and group: List the plant system, Code, and type of snubber (hydraulic or 

mechanical).

Snubber identification: List a unique identifier for each snubber; this identifier should be used 

consistently in all ISI and testing program documentation and design information such as 

system piping and instrument diagrams (P&IDs), isometric, test procedures, and relief requests.

Drawings number: List the applicable isometric, support drawing or figure that depicts the 

snubber.
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Test frequency: List the actual frequency for each inspection and test to be performed. 

Relief request(s): List any applicable relief requests in the snubber list. 

2.4.6 Snubber Programs Documentation and their Submittal to NRC

10 CFR 50.55a(g)(4) requires that, throughout the service life of a boiling or pressurized watercooled nuclear power facility, ASME Code Class 1, 2, and 3 components (including supports) 

meet the ISI and testing requirements of the ASME B&PV Code, Section XI or ASME OM Code 

as incorporated by reference in 10 CFR 50.55a(b). The applicable ASME B&PV Code, Section 

XI, Article IWA-1000, “General Requirements,” and ASME OM Code, Subsection ISTA, 

“General Requirements,” provide the documentation and submittal requirements for inservice 

testing and examination of certain components in light-water nuclear power plants. Therefore, 

based on these requirements, licensees are required to submit their snubber examination and 

testing programs and their updates every 120 months. 

(a) Documentation requirements for snubber programs when using the ASME B&PV 

Code, Section XI

IWA-1400(c) notes that owners, have the responsibility to prepare plans, schedules, and 

inservice inspection summary reports, and submit of these plans and reports to the 

enforcement and regulatory authorizes having jurisdiction at the plant site.

Article IWA-6000, Record and Reports, provides the requirements for preparation, 

submittal, and retention of records and reports.

(b) Documentation requirements for snubber programs when using the ASME OM 

Code

ISTA-3200(a) requires that plans for inservice examination and testing of snubbers shall 

be filed with the regulatory authorities having jurisdiction at the plant site.

ISTA-9000, Records and Reports, provides the requirements for preparation, submittal, 

and retention of records and reports.

Nonmandatory Appendix-A, and the Supplement to Nonmandatory Appendix-A 

describes voluntary guidance for licensees to develop snubber inservice examination 

and testing plans. 

(c) Documentation requirements for snubber programs when using NRC authorized 

alternative TS, TRM or other-licensee-controlled documents in lieu of the ASME 

B&PV Code, Section XI, or ASME OM Code

NRC authorized relief to use TRMs or other-licensee-controlled documents, in lieu of the 

ASME B&PV Code, Section XI or ASME OM Code requirements for inservice 

examination and testing of snubbers, do not provide relief from submitting snubber 

programs to the regulatory authorities. Submittal is required by the applicable ASME 

B&PV Code, Section XI or ASME OM Code as noted in (a) and (b) above. 
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Licensees not meeting the requirements of IWA-1400(c) or ISTA-3200(a) must submit 

appropriate documents containing snubber inservice examination and testing plans and submit 

a request for relief to the NRC pursuant to 10 CFR 50.55a(a)(3). NRC staff will not perform a 

review of submitted snubber inservice examination and testing programs unless requesting 

alternatives or reliefs to Code requirements.

2.5 Relief Requests and Proposed Alternatives

Licensees are required to perform the ISI and testing of snubbers in accordance with ASME 

BPV Code, Section XI or the OM Code and the applicable addenda as required by 

10 CFR 50.55a(g) or 10 CFR 50.55a(b)(3)(v), except where the NRC has granted specific 

written relief, pursuant to 10 CFR 50.55a(g)(6)(i), or authorized alternatives pursuant to 

10 CFR 50.55a(3). 10 CFR 50.55a(a)(3) states that licensees may use alternatives to the 

requirements of 10 CFR 50.55a(g) when authorized by the NRC if (1) the proposed alternatives 

would provide an acceptable level of quality and safety, or (2) compliance with the specified

requirements would result in hardship or unusual difficulty without a compensating increase in 

the level of quality and safety A licensee may submit a request for the NRC to review and 

approve relief from requirements of the Code, or to authorize the use of proposed alternatives. 

The staff recommends that the basis for relief should address the following considerations:

(1) Does the proposed alternative provide an acceptable level of quality and safety? 

(10 CFR 55a(a)(3)(i))

(2) Would compliance with the specified requirement (for which relief is sought) result in a 

hardship or unusual difficulty without a compensating increase in the level of quality and 

safety? 

(10 CFR 55a(a)(3)(ii))

(3) Would it be impractical to comply with the Code requirements? (10 CFR 55a(f)(6)(i))

The justification must include adequate information for the staff to determine if the relief can be

granted or the alternative can be authorized. NRC approval is required before a licensee may 

implement proposed alternatives that must be authorized pursuant to 10 CFR 50.55a(3). By 

contrast, a licensee may implement proposed alternative testing while the NRC is reviewing 

requests for relief from Code requirements made pursuant to 10 CFR 50.55a(f)(6)(i), if the 

licensee has determined that the requirements are impractical.

The staff performs a detailed review of each relief request, grants relief from the requirements or

authorizes an alternative to those requirements, and may impose alternative requirements. 

When granting relief, the NRC considers the burden that would be imposed upon the licensee if

the agency enforced the specified requirements.

For more details contents and format etc. see NEI issued document, “Standard Format for 

Requests from Commercial Reactor Licensees Pursuant to 10 CFR 50.55a, Revision 1, dated 

2004.

Currently, few licensees are using the ASME OM Code to meet the requirements of 10 CFR 

50.55a for snubber inservice examination and testing, whereas most of the licenses are using a
B 2-13

variety of licensee-controlled documents or procedures in lieu of the applicable ASME Code 

requirements. These licensee-controlled documents or procedures include the following:

1. Technical Specification (TS)

2. Technical Requirement Manual (TRM)

3. Final Safety Analysis Report (FSAR)

4. Updated Final Analysis Report (UFSAR)

5. Selected Licensee Commitment (SLC)

6. Licensee-Controlled Specification (LCS)

7. Equipment Control Guidelines (ECG)

8. Other Licensee-Controlled Procedures

Recently, the NRC staff has identified several instances in which nuclear power plants licensees 

have used a TRM, or other licensee-controlled documents and procedures, which do not meet 

requirements of their “Code of Record” for the ISI and testing of snubbers. These licensees 

have not requested approval to use these alternatives from the Commission. The NRC issued 

Regulatory Issue Summary (RIS) 2010-06, “Inservice Inspection and Testing Requirements of 

Dynamic Restraints (Snubbers)” on June 1, 2010 to remind all the licensees of the NRC’s rules 

and regulations regarding snubber ISI and testing, in accordance with 10 CFR 50.55a(g), at 

nuclear power plants.

The NRC expects licensees to ensure that their snubber ISI and testing programs are in 

compliance with 10 CFR 55.55a(g) or authorized alternatives. If licensees discover that their 

programs are not meeting 10 CFR 50.55a(g) requirements or authorized alternatives, they 

should take appropriate actions to bring their programs back into compliance and ensure that 

non-compliant systems, structures and components are operable. In certain circumstances 

involving snubber programs at nuclear power plants that are not in compliance with NRC 

requirements, enforcement discretion has been provided by the NRC. The NRC’s Office of 

Enforcement issued Enforcement Guidance Memorandum (EGM)-10-001, “Dispositioning 

Violation of Inservice Examination and Testing Requirements for Dynamic Restraints 

(Snubbers),” on June 1, 2010 to provide NRC staff guidance for the disposition of certain 10 

CFR 50.55a violations and the potential of granting enforcement discretion for the affected 

requirements. NRC expects that licensees of nuclear power plants, who were not meeting the 

10 CFR 50.55a requirements for snubber inservice examination and testing as described in RIS 

2010-06, should have entered any noncompliance into the corrective action and corrected the 

noncompliance by meeting the applicable ASME Code requirements or by submitting for relief 

to NRC.

2.6 Snubber Program Documents

Snubber ISI and testing program documents submitted to the NRC are used to prepare for ISI

inspections and to address other licensing actions that may arise. Between a licensee’s 10-year 

interval program submittals, the NRC would like to receive up-to-date program documents when 

the licensee makes significant changes to the snubber ISI and testing program to facilitate these 

regulatory activities.
B 2-14

As long as the snubber ISI and testing program is consistent with the regulations, ASME Code 

relief is not required. That is, deletions from or additions to the snubber program do not 

necessarily require NRC approval, unless commitments to do exist as a result of approved relief 

requests. The burden is on each licensee to verify that its snubber program is complete and 

includes all snubbers that require inservice examination and testing. If a licensee deletes a 

particular snubber from its snubber program, the staff recommends that the licensee should 

document the basis in an appropriate manner.

The staff expects each licensee to maintain its snubber examination and testing program up-todate and ensure that it remains consistent with changes in plant configuration. Conversely, if a

system modification results in the addition of a snubber to the snubber program, the licensee

should ensure that it is incorporated into the program to satisfy the Code and licensing 

requirements or that a relief request is submitted for NRC review and approval, as appropriate.

2.7 Repair and Replacement of Snubbers:

The repair and replacement of snubbers is to be performed using applicable edition and 

addenda of the ASME B&PV Code, Section XI, Article IWA-4000. The NRC approved Inservice 

Inspection Code Cases in RG 1.147, such as N-508-3, “Rotation of Serviced Snubbers and 

Pressure Relief Valves for the purpose of Testing, Section XI, Division 1,” may be used.

2.8 ISI of the Integral and Non-Integral Attachments of Supports Containing Snubbers

Inservice inspection of integral and non-integral attachments, such as lugs, bolts, pins, and 

clamps must be performed by use of ASME B&PV Code, Section XI, Subsection IWF. ASME 

OM Code, Subsection ISTD covers inservice examination and testing of snubbers (pin to-pin) 

inclusive and does not address the inspection of integral and non-integral attachments, such as 

lugs, bolting, pins, and clamps. 
B 2-15

2.9 Developing Snubber Program for New Nuclear Power Plants

Under 10 CFR Part 52, the development of a plant-specific IST program is the responsibility of 

the applicant for a combined license (COL) to construct and operate a nuclear power plant. The 

Commission’s Staff Requirements Memorandum (SRM), dated September 11, 2002, for 

Commission Paper SECY-02-0067, “Inspections, Tests, Analyses, and Acceptance Criteria 

(ITAAC) for Operational Programs (Programmatic ITAAC),” stated that ITAAC for an operational 

program are unnecessary if the program and its implementation are fully described in the COL 

application and found to be acceptable by the NRC. The Commission also stated that the 

burden is on the COL applicant to provide the necessary and sufficient programmatic 

information for approval of the COL without ITAAC. 

In its May 14, 2004, SRM for SECY-04-0032, “Programmatic Information Needed for Approval 

of a Combined License Without Inspections, Tests, Analyses and Acceptance Criteria,” the 

Commission defined “fully described” as meaning that the program is clearly and sufficiently 

described in terms of the scope and level of detail to allow a reasonable assurance finding of 

acceptability. The Commission also noted that required programs should always be described 

at a functional level and at an increasing level of detail where implementation choices could 

materially and negatively affect the program effectiveness and acceptability. SECY-05-0197, 

“Review of Operational Programs in a Combined License Application and Generic Emergency 

Planning Inspections, Tests, Analyses, and Acceptance Criteria,” summarizes the NRC position 

regarding the full description of operational programs to be provided by COL applicants. RG 

1.206, “Combined License Applications for Nuclear Power Plants (LWR Edition),” provides 

guidance for COL applicants with respect to fully describing plant operational programs. 

COL applicants provide a description of the snubber examination and testing program for NRC 

review as part of their COL applications. In some cases, the COL applicant incorporates by 

reference in its Final Safety Analysis Report (FSAR) the description of the snubber examination 

and testing program provided in the Design Certification documentation, such as the Design 

Control Document (DCD) or Design Certification FSAR. RG 1.206 provides guidance regarding 

information to be provided in the COL application as part of the description of the IST program 

for snubbers. To date, COL applicants have described their snubber examination and testing 

program based on the requirements in ASME OM Code, Subsection ISTD. The guidance in this 

NUREG may be used in developing and implementing the IST program for new nuclear power 

plants. 

The NRC staff is reviewing the descriptions of IST programs, including snubber examination 

and testing programs, provided by COL applicants as part of their COL applications. The staff 

will reach a conclusion regarding the acceptability of the description of the IST program in its 

safety evaluation on the COL application. The NRC staff will conduct inspections of the

development and implementation of IST programs during the construction of new nuclear power 

plants following COL issuance.
B 2-16

Table 2.1 Typical Systems Where Snubbers are included in the 

Inservice Examination and Testing Snubber Program

for a Pressurized-Water Reactor (non-inclusive)

Typical safety-related, Code-class system in pressurized-water reactors

Reactor coolant system and flowpaths for establishing natural circulation

Main steam system

High-pressure safety injection system

Chemical and volume control or makeup system

Low-pressure safety injection system

Shutdown cooling, residual heat removal, or decay heat removal systems

Containment spray system

Main feedwater system

Auxiliary feedwater system

Primary containment system

Component cooling water system

Spent fuel pool/pit cooling system

Service water system

Emergency diesel generator system (within scope of 10 CFR 50.55a)

Ventilation systems

Instrument air system (if within the scope of 10 CFR 50.55a)
Note: Flow chart provided for guidance only. For complete details see 10 CFR 50.55a

B 2-17

Table 2.2 Typical Systems Where Snubbers are included in the 

Inservice Examination and Testing Snubber Program

for a Boiling-Water Reactor (non-inclusive)

Typical safety-related, Code-class system in boiling-water reactors

Nuclear boiler and reactor recirculation system

Main steam system

High-pressure core coolant injection (HPCI) system

High-pressure core spray system

Reactor core isolation cooling (RCIC) system (if safety-related)

Reactor water cleanup system

Residual heat removal (RHR) system

Spent fuel pool cooling system

Feedwater coolant injection and isolation condenser system (if applicable)

Standby liquid control (SBLC) system

Main feedwater system

Primary containment system

Closed cooling or component cooling water system

Service water system

Control rod drive system (portions within the scope of 10 CFR 50.55a)

Emergency diesel generator systems (if within the scope of 10 CFR 50.55a)

Ventilation systems

Instrument air system (if within the scope of 10 CFR 50.55a)

Traversing incore probe system (if within the scope of 10 CFR 50.55a)
*Note: Flow chart provided for guidance only. For complete details see 10 CFR 50.55a

B 2-18

Figure 1 – Appendix-B

Flowchart - Development of Inservice Inservice Examination and Testing

Program for Snubbers*

Development of 

(1) Snubber Inservice Examination and Testing 

Program (10 CFR 50.55a), or

(2) Mandatory update of Snubber Program Every 

10-Year Interval 

[10 CFR 50.55a(f)(4) or 10 CFR 50.55a(g)(4)], 

[Regulatory Issue Summary (RIS) 2010-06]

Licensee must Submit Relief or Alternative 

via Relief Request (RR) to use Alternative 

Requirements such as TS, TRM or other 

licensee-controlled document for a 10-Year 

Interval

[NEI Guidance for Standard Format of RR 

dated June 7, 2004 

 

ASME B&PV Section XI, 

Article IWF-5000

[10 CFR 50.55a(b)]

[Note: ASME Section XI, 

2006 addenda deleted 

Snubber ISI and Testing 

requirements; then, licensees 

must use ASME OM Code]

ASME OM Code, 

Subsection ISTD

[10 CFR 50.55a(b)(3) (v)]

IWF-1000, General Requirements

IWF-5200 and IWF-5300

(ASME/ANSI OM-1987, Part 4 

(OM-4) with OMa-1988 addenda)

Use Subsection IWF of ASME 

B&PV Code, Section XI for Integral 

and nonintegral attachments etc. of 

snubbers

ISTA, General Requirements

ISTD, Snubber (pin-to-pin) Inservice 

Examination & Testing Requirements

Use Subsection IWF of ASME B&PV 

Code Section XI for Integral and 

nonintegral attachments etc. of snubbers

IWA-1400(c), Requires Licensee to 

Submit Snubber Program to Regulatory 

Authority (NRC)

IWA-6000 provides “Record and Reports” 

requirements.

Note: NRC does not review and approve 

submitted snubber program..

ISTA-3200, Requires Licensee to Submit 

Snubber Program to Regulatory Authority

(NRC) 

ISTA-9000, provides “Record and 

Reports” requirements

Note: NRC does not review and approve 

submitted snubber program. 

NRC authorized relief or alternative 

to use TS, TRM or other licenseecontrolled document in lieu of the 

ASME Code requirements

10 CFR50.55a(a)(3)(i)

Applicable ASME Code 

IWA-1400(c) or ISTA-3200, Requires 

Licensee to Submit Snubber Program to 

NRC 

IWA-6000 or ISTA-9000 provide “Reports 

and Records” requirements

Note: NRC does not review and approve 

submitted snubber program.

Sunbber Examination and 

Testing Program based on 

authorized alternative TS, 

TRM or licensee-controlled 

document

Not using or 

meeting 

ASME Code 

Requirement

May use Code Cases,

listed in Regulatory Guide 

(RG) 1.192 as conditioned 

[10 CFR 50.55a(b)(6)]

Yes, meeting 

ASME Code 

requirements

Use of

Either Code

Meet the Snubber ISI & IST 

requirements as specified in 

(1)ASME OM Code with 

Edition and addenda, or

(2) ASME B&PV, Section XI 

with Edition and Addenda

 [10 CFR 50.55a(b)]

May submit RR to use 

Code Case not listed in 

RG 1.192

[10 CFR 50.55a(a)(3)(i)]
B 3-1

3. REFERENCES

3.1 U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization 

Facilities, Title 10, “Energy,” Chapter I – Nuclear Regulatory Commission, Part 50, 

Section 50.55a, Codes and standards.

3.2 American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section 

XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2004 Edition thru 

2008 Addenda.

3.3 American Society of Mechanical Engineers Operation and Maintenance Code, Rules for 

Inservice Testing of Light-Water Reactor Power Plants. 2004 Edition with 2005 and 2006 

Addendas

3.4 American Society of Mechanical Engineers Operation and Maintenance Code, Operation 

and Maintenance of Nuclear Power Plants, Part 4 (OM-4) 1987 Edition with 1988 

Addenda.

3.5 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.192, “Operation and 

Maintenance Code Case Acceptability: ASME OM Code,” Washington, DC, June 2003.

3.6 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.193, “ASME Code Cases Not 

Approved for Use,” Washington, DC, June 2003.

3.7 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.147, “Inservice Inspection 

Code Case Acceptability, ASME Section XI, Division 1,” Washington, DC, October 2010.

3.8 Generic Letter (GL) 90-09, “Alternative Requirements for Snubber Visual Inspection 

Intervals and Corrective Actions,” dated December 11, 1990.

3.9 Nuclear Energy Institute (NEI), White Paper, entitled Standard Format for Requests from 

Commercial Reactor Licensees Pursuant to 10 CFR 50.55a, Rev 1 June 2004.

3.10 Regulatory Issue Summary (RIS) 2010-06, “Inservice Inspection and Testing 

Requirements of Dynamic Restraints (Snubbers)” dated June 1, 2010.

3.11 Enforcement Guidance Memorandum (EGM)-10-001, “Dispositioning Violation of 

Inservice Examination and Testing Requirements for Dynamic Restraints (Snubbers),” 

on June 1, 2010.

3.12 NRC Generic Issue (GI-113), “Dynamic Qualification and Testing of Large Bore 

Hydraulic Snubbers (LBHSs),”

3.13 Technical Evaluation of Generic Issue 113: Dynamic Qualification and Testing of Large 

Bore Hydraulic Snubbers, NUREG/CR-5416.

3.14 Generic Letter from D. Eisenhut (U.S. NRC) to all Power Reactor Licensees, “Technical 

Specification Revision for Snubber Surveillance,” November 20, 1980.
B 3-2

3.15 Letter from B. D. Johnson (Consumer Power Company) to J. G. Keppler (USNRC), 

“Palisades Plant- 10 CFR 21 Report- Defective ITT Grinnell Hydraulic Snubbers,” date 

September 30, 1983.

3.16 NUREG/CR-5870, Results of LWR Snubber Aging Research, Date Published May 1992