United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 96-32: Implementation of 10 CFR 50.55a(g)(6)(ii)(A), "Augmented Examination of Reactor Vessel"

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C.  20555

                                 June 5, 1996


NRC INFORMATION NOTICE 96-32:  IMPLEMENTATION OF 10 CFR 50.55a(g)(6)(ii)(A), 
                               "AUGMENTED EXAMINATION OF REACTOR VESSEL"


Addressees

All holders of operating licenses or construction permits for nuclear power
reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to certain aspects of scheduling and implementing
the augmented reactor vessel examination required by Section
50.55a(g)(6)(ii)(A) of Title 10 of the Code of Federal Regulations (10 CFR). 
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.

Background

Because of concerns regarding the scope of inspection of reactor vessels, the
NRC issued, in 1992, 10 CFR 50.55a(g)(6)(ii)(A), "Augmented Examination of
Reactor Vessel" [hereinafter referred to as Paragraph (A)], which contains new
requirements for an augmented examination of reactor vessels.  The rule
requires licensees to implement, before the time required by normal updating
of the inservice inspection (ISI) program, provisions in the 1989 Edition of
the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code
(ASME Code), Section XI, to examine "essentially 100%" of the length of all
reactor vessel shell welds.  Licensees with fewer than 40 months remaining in
the ISI interval that was in effect on September 8, 1992, may defer the
augmented reactor vessel examination to the first period of the next ISI
interval [Paragraph (A)(3)].  "Essentially 100%" examination is defined in
Paragraph (A)(2) as "more than 90% of the examination volume of each weld"
[emphasis added].    

Licensees unable to completely satisfy the requirements for the augmented
reactor vessel examination must propose an alternative that would provide an
acceptable level of quality and safety.  The proposed alternative may be used
when authorized by the Director of the Office of Nuclear Reactor Regulation
(NRR) [Paragraph (A)(5)]. 


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The 1989 Edition of the ASME Code, Section XI, incorporated Appendix VIII,
"Performance Demonstration for Ultrasonic Examination Systems."  Appendix VIII
was developed to ensure the effectiveness of ultrasonic examinations through a
performance demonstration to evaluate the adequacy of procedures, equipment,
and personnel for detecting and sizing flaws during examinations.  Licensees
are not currently required to implement Appendix VIII.

Description of Circumstances

It became evident to the staff while it was conducting ISI reviews that some
licensees were unaware of or uncertain about some aspects of the augmented
reactor vessel examination rule.  

The staff learned that a small number of licensees were unaware of the rule
and its requirements for some time after it was published.  Licensees need to
be aware of the schedular requirements of the rule to ensure timely
implementation of its provisions.  Because of the scope and extent of the
examination, significant planning is necessary to address the technical,
schedular, and regulatory issues associated with a comprehensive examination
of the reactor pressure vessel.

This information notice contains a discussion of certain areas of
misinterpretation that the staff has dealt with in the implementation of the
augmented reactor vessel examination rule.

Discussion

Schedular Requirements of the Rule

In one instance, a licensee original 10-year ISI interval end date allowed
deferral to the first period of the next interval.  However, this licensee
experienced an extended shutdown and, as permitted by Section XI, extended the
ISI interval to complete the examinations required for the interval.  As a
result, more than 40 months remained in the interval in effect on September 8,
1992, and the licensee would have been required to do the examination sooner
than expected.  The licensee requested and was granted approval by NRR to
schedule the examination in accordance with the original 10-year ISI interval
end date to allow for proper scheduling and to ensure the availability of
examination equipment.

"Essentially 100%" Examination Standard

Most licensees are finding that while the overall average examination coverage
for reactor vessel shell welds may be more than 90%, examination coverage for
individual welds may be substantially less than 90%.  When a licensee is
unable to examine "essentially 100%" of each shell weld, it must seek NRC
authorization of an alternative in accordance with Paragraph (A)(5).

During discussions with the NRC staff regarding the review of the 10-year ISI
program plan, a licensee stated that it had obtained "essentially 100%" .                                                            IN 96-32
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                                                            Page 3 of 4


coverage of the total volume of the reactor vessel shell welds but coverage of
less that 90% of several individual welds.  Contrary to the requirements of   
the rule, the licensee did not submit a request for authorization of an
alternative to the NRC as required by the rule, until asked to do so by the
NRC.

"Spirit of Appendix VIII" Examination
 
Section XI contains rules for evaluating the significance of flaws identified
through non-destructive examination.  Flaws that are of such size that they
cannot be dispositioned through comparison with code tables must be analyzed
in accordance with Section XI, Paragraph IWB-3600, "Analytical Evaluation of
Flaws."  Furthermore, Section XI, Paragraph IWB-3134(b), "Review by
Authorities," requires that analytical evaluations performed in accordance
with Paragraph IWB-3600 be submitted to the regulatory authority having
jurisdiction at the plant site (i.e., NRC).

One licensee administered a "Spirit of Appendix VIII" performance
demonstration for the procedures, personnel, and equipment to be used for the
augmented reactor vessel examination.  This type of examination essentially
satisfies the technical requirements of Appendix VIII and would be expected to
yield more accurate and reliable inspection results.  The licensee concluded
that the performance demonstration resulted in examination and evaluation
techniques that surpassed the conventional techniques of Section XI of the
ASME Code and Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel
Welds During Preservice and Inservice Examinations."  During the augmented
reactor vessel examination, the licensee identified 15 flaws in the shell
welds and in the shell-to-flange weld outside the scope of the augmented
reactor vessel examination, which required analytical evaluation in accordance
with Section XI, Paragraph IWB-3600.  The licensee stated that if the
conventional techniques of Section XI and Regulatory Guide 1.150 had been
used, 12 of these 15 flaws would not have even been recordable and only 2 of
the remaining 3 flaws would have required analytical evaluation in accordance
with Paragraph IWB-3600.  This licensee experience indicates that flaws of
sufficient size to require analytical evaluation may not be detected when
using conventional techniques for the augmented reactor vessel examination.

Although the licensee in the above example submitted a request for
authorization of an alternative as the examination coverage ed by the
ASME Code, until asked to do so by the NRC.  

Need for NRC Authorization of Alternatives      

A licensee unable to obtain the required examination coverage quoted 10 CFR
50.55a(g)(4) as a basis for not seeking NRC authorization of an alternative as
required by Paragraph (A)(5).  However, 10 CFR 50.55a(g)(4) states, in part,
that "components. . . must meet the requirements. . . to the extent practical
within the limitations of design, geometry and materials of construction of
the components."  As with relief requests for other Code components for .                                                            
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incomplete or partial ASME Code-required ISI examinations, NRC authorization
is required when all the examination requirements of Paragraph (A) are not
met.

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


                                          signed by

                                       Brian K. Grimes, Acting Director       
                                       Division of Reactor Program Management
                                       Office of Nuclear Reactor Regulation

Technical contacts:  Edmund J. Sullivan, NRR          
                     (301) 415-3266                   
                     Internet:ejs@nrc.gov             

                     Eric J. Benner, NRR
                     (301) 415-1171
                     Internet:ejb1@nrc.gov

Attachments:
1.  Referenced Codes and Standards
2.  List of Recently Issued NRC Information Notices

                                                            Attachment 1
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                                                            June 5, 1996
                                                            Page 1 of 1

                     Referenced Codes and Standards

1.  Title 10 of the Code of Federal Regulations (10 CFR), Section
    50.55(g)(6)(ii)(A), "Augmented Examination of Reactor Vessel"

2.  American Society of Mechanical Engineers, Boiler and Pressure Vessel
    Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant
    Components," 1989 Edition.


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