United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 95-09: Use of Inappropriate Guidelines and Criteria for Nuclear Piping and Pipe Support Evaluation and Design

                                 UNITED STATES
                          NUCLEAR REGULATORY COMMISSION
                          WASHINGTON, D.C.  20555-0001

January 31, 1995

                               AND DESIGN


All holders of operating licenses or construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to the use of guidelines and criteria for nuclear
piping and pipe support evaluation and design in a manner inconsistent with
published NRC staff guidance.  It is expected that recipients will review the
information for applicability to their facilities and consider actions, as
appropriate, to avoid similar problems.  However, suggestions contained in
this information notice are not NRC requirements; therefore, no specific
action or written response is required.

Description of Circumstances

By letter dated March 28, 1994, the Nuclear Energy Institute (NEI) submitted
several documents prepared by the Electric Power Research Institute (EPRI)
regarding piping and pipe support operability evaluations.  These guidance
documents were intended to clarify the guidance in NRC Generic Letter (GL) 
91-18, "Information to Licensees Regarding Two NRC Inspection Manual Sections
on Resolution of Degraded and Nonconforming Conditions and on Operability." 
Although NEI submitted these documents for information only, a cursory review
by NRC staff indicated that the proposed operability criteria are inconsistent
with the guidance for the NRC staff contained in GL 91-18.  The EPRI documents
recommend the use of criteria that are different than the current endorsed
editions of the American Society of Mechanical Engineers (ASME) Code or other
guidance endorsed by the NRC staff.


On November 7, 1991, the NRC issued GL 91-18 to all nuclear power reactor
licensees and applicants to inform them of guidance issued to ensure
consistency by the NRC staff during the review of licensee operability
determinations and resolution of degraded and nonconforming conditions.
Section 6.13, "Piping and Pipe Support Requirements," of the enclosure to
GL 91-18 gives guidance on operability determinations for piping and pipe
supports.  This section references the criteria in NRC Inspection and
Enforcement (IE) Bulletins 79-02, "Pipe Support Base Plate Designs Using

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                                                            January 31, 1995
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Concrete Expansion Anchor Bolts," and 79-14, "Seismic Analysis for As-Built
Safety-Related Piping Systems," and the applicable revisions and supplements. 

To further clarify the guidance in GL 91-18, EPRI prepared documents for
member utilities to use when evaluating piping and pipe support operability. 
These documents include:

      1.  "Piping System Short-Term Operability Criteria," a report of
          detailed technical criteria and methods for evaluating the
          short-term operability of piping systems;

      2.  "Guideline for Operability Evaluations of Piping Systems," the
          related implementation procedure guideline;

      3.  "Technical Criteria Report Commentary Short-Term Operability
          Criteria for Piping Systems," the companion commentary explaining
          the rationale for the detailed criteria; and

      4.  EPRI Report TR-101968, "Guidelines and Criteria for Nuclear Piping
          and Support Evaluation and Design."

These documents were not submitted for a formal NRC staff review and
endorsement.  However, the staff performed a cursory review and concluded that
they contain operability criteria that are inconsistent with the guidance to
the NRC staff in GL 91-18.  For example, EPRI Report TR-101968, Section
5.2.1., "Piping," recommends the use of the higher allowable stresses for
seismic designs included in a draft ASME Section XI Code Case.  These higher
allowable stresses are also contained in a revision to the criteria in Section
III of the ASME Code.  These increased allowable stresses have not been
endorsed by the staff.  The EPRI report recommends the use of criteria in
Welding Research Council Bulletin 352, which also has not been endorsed by the
staff.  Because the staff has not reviewed the EPRI documents in detail, they
may contain other guidance that is inconsistent with GL 91-18, the ASME Code,
or other staff guidance.  The staff does not discourage the industry from
developing operability criteria and guidelines for situations that may not be
addressed by GL 91-18.  However, the staff cannot, without formally reviewing
the documents, verify that such guidance, methods, and criteria are consistent
with those previously endorsed by the staff.  Therefore, NRC inspectors will
continue to review operability evaluations according to practices, methods and
criteria consistent with GL 91-18 and the current NRC staff-endorsed ASME Code

The staff has learned that some EPRI member utilities are applying the EPRI
guidance discussed above.  However, these operability criteria may be
inconsistent with previously issued guidance to resolve degraded and
nonconforming conditions.  In GL 91-18, the staff discussed the use of the
criteria in Appendix F of Section III of the ASME Code for operability
determinations.  Title 10 of the Code of Federal Regulations specifies
editions and addenda of the ASME Code that are formally endorsed by the NRC.  .                                                            IN 95-09
                                                            January 31, 1995
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Section 50.55a also incorporates by reference NRC Regulatory Guide 1.84,      
"Design and Code Case Acceptability-ASME Section III Division 1," NRC
Regulatory Guide 1.85, "Materials Code Case Acceptability-ASME Section III     
Division 1," and NRC Regulatory Guide 1.147, "Inservice Inspection Code Case
Acceptability-ASME Section XI Division 1."  These regulatory guides list the
ASME Code Cases that have been determined suitable and are endorsed by the
staff for use by licensees.  The Director of the Office of Nuclear Reactor
Regulation (NRR) may authorize the use of other ASME Code Cases upon request
pursuant to 10 CFR 50.55a(a)(3).

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
the technical contact listed below or the appropriate NRR project manager.

                                    /s/'d by BKGrimes

                                    Brian K. Grimes, Director
                                    Division of Project Support
                                    Office of Nuclear Reactor Regulation

Technical contact:  Howard J. Rathbun, NRR
                    (301) 504-2787

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