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NRC Section XI Report – February 2009

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Presented By:  Mr. Wally Norris, United States Nuclear Regulatory Commission

NRC Report
February 2009

1.    Amendment to 10 CFR 50.55a – ASME Code Edition/Addenda


A final rule was published in the Federal Register [ 73 FR 52730] on September 8, 2008, incorporating Section III and Section XI of the 2004 Edition of the American Society of Mechanicals Engineers (ASME) Boiler and Pressure Vessel Code into Title 10, Part 50.55a, of the Code of Federal Regulations (10 CFR 50.55a).  The effective date of the rule was October 10, 2008.


An amendment to the above rule was published in the Federal Register [73 FR 57235] on October 2, 2008, to correct several paragraph references.


The NRC staff has begun its review of the 2005 Addenda through the 2008 Addenda.  These edition/addenda will be included in the next proposed rulemaking which is scheduled to be published for public comment at the end of 2009.


2.    ASME Code Case Rulemaking/Regulatory Guides


Draft Revision 35 to RG 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III,” draft Revision 16 to RG 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,”and draft Revision 3 to RG 1.193 “ASME Code Cases Not Approved for Use” are in final concurrence.  The guides address Code Cases from Supplement 2 to the 2004 Edition through Supplement 0 to the 2007 Edition (Supplement 0, 2007 Edition also serves as Supplement 12 to the 2004 Edition).  The proposed rulemaking is in final concurrence.


The publication date for the proposed rulemaking is uncertain as a memorandum distributed shortly after the inauguration by the new administration stated that no proposed or final regulation should be sent to the Office of the Federal Register for publication until it has been reviewed by administration representatives.  The process for review has not been finalized.  For final regulations that have been published but have yet to take effect, the administration asked that agency officials consider extending the implementation period for 60 days.  The extension would include reopening the notice and comment period for 30 days.  However, for rules that raise no substantial questions of law or policy, it was also stated that no further action needs to be taken.


The NRC staff has completed its review of Supplements 1 - 6, to the 2007 Edition, and draft revisions 36 to RG 1.84 and 17 to RG 1.147 are under development.


3.    Risk–Informed Activities


Reactor Vessel Weld Inspection


A proposed amendment to 10 CFR 50.61a was published in the Federal Register on October 3, 2007 (72 FR 56275).  The NRC staff has reviewed public comments on the proposed rule and is preparing the final rule.  If the final 10 CFR 50.61a differs from the proposed 10 CFR 50.61a with regard to the augmented inservice inspection (ISI) evaluation requirements, it is expected that the PWROG will review the requirements in the final 10 CFR 50.61a and determine whether a revision to the accepted TR WCAP-16168-NP, Revision 2, is required.  Furthermore, licensees that choose to implement 10 CFR 50.61a must perform the ISI required in Section (e) of the rule, and must submit the required information for review and approval to the Director, Office of Nuclear Reactor Regulation, at least three years before the limiting RTPTS value calculated under 10 CFR 50.61 is projected to exceed the PTS screening criteria in 10 CFR 50.61.  Licensees implementing Section (c) of 10 CFR 50.61a must perform the inspections and analyses required by Section (e) of 10 CFR 50.61a prior to implementing the extended interval.


On August 8, 2008, a revised regulatory analysis for the supplemental proposed rule to amend alternate fracture toughness requirements for protection against pressurized thermal shock events (10 CFR 50.61a) was published [ML081440673].  During the development of the PTS final rule, the NRC determined that several significant changes to the proposed rule language would be needed to adequately address the stakeholders’ comments and their associated implementation concerns.  Two of the modifications are significant changes to the proposed rule language on which external stakeholders did not have an opportunity to comment.  The NRC concluded that obtaining stakeholder feedback on these provisions through the use of a supplemental proposed rule was appropriate.  The two modifications subject to comments from the public do not have a measurable impact in this regulatory analysis.  However, in the supplemental proposed rule, the NRC is considering limiting the applicability and the use of 10 CFR 50.61a to currently-operating plants only.  Therefore, the regulatory analysis was modified to reflect this change.


Phased Approach to Probabilistic Risk Assessment Quality

Regulatory Guide 1.200


The increased use of probabilistic risk assessments (PRAs) in the NRC’s regulatory decisionmaking process requires consistency in the quality, scope, methodology, and data used in such analyses.  A key aspect of implementing a phased approach to PRA quality is the development of PRA standards and related guidance documents.  To achieve that objective, professional societies, the nuclear industry, and the staff have undertaken initiatives to develop national consensus standards and guidance on the use of PRA in regulatory decisionmaking.


On October 6, 2008, the NRC received a letter from Bryan A. Erler, Vice President, ASME Nuclear Codes and Standards, and N. Prasad Kadambi, Chairman, ANS Standards Board, [ML082971081] updating the Commission on the status of efforts by ASME and ANS in producing PRA standards.  In addition to providing the update, the letter requested that the public review and comment period for proposed Revision 2 to Regulatory Guide (RG) 1.200 be extended to the end of December 2008.


In a letter dated October 31, 2008, from Brian W. Sheron, Director, Office of Nuclear Regulatory Research, NRC, to Mr. Erler [ML082910565], the NRC approved extending the public review and comment period to December 31, 2008.  A notice of the granting of request to extend the comment period was published in the Federal Register on November 3, 2008 [73 FR 65414].  The final publication of Revision 2 to RG 1.200 is now scheduled for March 31, 2009.


Certain ASME and ANS standards applying to at-power internal events, internal fire events, and external events were combined into a single joint standard, “Standards for Level 1/Large Early Release Frequency (LERF) Probabilistic Risk Assessment Standard for Nuclear Power Plant Applications” (ASME/ANS RA-S-2008).  Accordingly, the staff had initiated work on a revision of RG 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,” that will endorse the joint standard.


The ASME/ANS letter indicates that a task group is developing an Addendum to ASME/ANS RA-S-2008 that is a substantial improvement in the quality and usability of the standard.  The Addendum was approved on September 12, 2008, by the technical consensus committee, and it is expected that it will receive ANSI approval late this year.




In November 2007, the staff issued draft NUREG-1855, “Treatment of Uncertainties from PRAs in Risk-Informed Decision-Making,” for public review and comment.  It is being developed in collaboration with the Electric Power Research Institute (EPRI) who has issued a draft report on uncertainties, as part of the NRC/EPRI Memorandum of Understanding.  These two documents are meant to be complimentary.  The NRC report along with the EPRI report provides information and guidance on uncertainties associated with PRA.  They are meant to provide guidance on meeting the requirements in the ASME/ANS PRA standard on uncertainties, and provide guidance on how to treat the results from the uncertainty analyses in decision making for risk-informed activities.  Two public meetings have been held, and the staff plans to issue NUREG-1855 by the Summer 2009.


The regulatory guide and NUREG report (including the EPRI report) will assist the staff in establishing the technical acceptability of the PRA results to be used in regulatory decision making.  When used in support of an application, these documents will obviate the need for an in-depth review of the base PRA by NRC reviewers, and provide for a more focused and consistent review process.


NEI 05-04, Rev. 2


On December 1, 2008, the Nuclear Energy Institute submitted NEI 05-04, Rev. 2, “Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard,” and NEI 07-12, Rev. 0, Draft H, “Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines” to the NRC staff for review.  These revisions incorporate lessons learned from internal events and fire PRA peer reviews and also reflect the upcoming release of the ASME/ANS RASa-2008, Addendum A, “Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications.”  the industry has requested that the NRC review and endorse the most recent revisions of both documents in DG-1200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,” when it is issued as RG 1.200, Revision 2.


4.    Generic Activities on Material Degradation/PWR Alloy 600/182/82 PWSCC


In February 2008 NRC issued Temporary Instruction (TI-172) for regional staff to verify that all pressurized water reactors (PWRs) with dissimilar metal (DM) butt welds are implementing Materials Reliability Program (MRP)-139, “Primary System Piping Butt Weld Inspection and Evaluation Guidelines.”


In 2006 ASME started the development of a Code Case for inspection of Alloy 82/182 butt welds.  The Code Case was recently completed.  The staff expects that the Code Case will be acceptable with comments/conditions and will be incorporated by reference directly into the next update to 10 CFR 50.55a.


The NRC staff continues to monitor and evaluate operating experience to ensure that the current inspection schedules are adequate.


On October 22, 2008, the NRC issued NRC Regulatory Issue Summary 2008-25, “Regulatory Approach for Primary Water Stress Corrosion Cracking of Dissimilar Metal Butt Welds in Pressurized Water Reactor Primary Coolant System Piping.”  The intent in issuing this regulatory issue summary (RIS) is to inform addressees of the regulatory approach for ensuring the integrity of primary coolant system DM butt welds containing Alloy 82/182 in PWR power plants.


In late October 2008, a small inside surface connected circumferential indication was identified in a hot leg nozzle to safe-end Alloy 82/182 weld.  As far as the NRC staff is aware, this is the first circumferential indication that has been identified in the U.S. in an Alloy 82/182 reactor vessel to hot leg piping weld.  The staff completed a preliminary evaluation of this flaw and, based on this evaluation, does not believe that this inspection finding has an effect of the current inspection schedules in MRP-139 or Code Case N-770, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 UNS W86182 Weld Filler Material With or Without the Application of Listed Activities.”


5.    New Reactor Licensing Activities


The New Reactor Licensing public web-site [http://www.nrc.gov/reactors/new-reactors.html] has a list of expected new nuclear power plant applications, and an estimated schedule by fiscal year for new reactor licensing applications.


New Reactor Licensing Status


As of January 27, 2009, the status of new reactors licensing under10 CFR Part 52 is as follows:


Design Certification


NRC has issued four design certifications to date (ABWR, System 80+, AP600, and AP1000).  These are certified in 10 CFR Part 52, Appendices A, B, C, and D, respectively.  The NRC is currently reviewing four design certifications:


  • General Electric-Hitachi’s ESBWR (first passive BWR)
  • AREVA’s EPR (evolutionary pressurized-water reactor)
  • Mitsubishi Heavy Industries’ US-APWR (advanced pressurized water reactor)
  • AP1000 Revision 16 (first amended design certification)


Early Site Permits (ESP)


NRC has issued three ESPs to date (Clinton, Grand Gulf, and North Anna).  The NRC is currently reviewing one ESP (Vogtle).  To date, there have been no ESPs submitted for greenfield sites.


Combined License (COL) Applications


NRC is currently reviewing 17 COL applications (26 new reactor units):

  • 1 ABWR         South Texas Project 3 and 4

·        6 AP1000      Bellefonte 3 and 4, William S. Lee Station 1 and 2, Shearon Harris 2 and 3, Vogtle 3 and 4, V.C. Summer 2 and 3, and Levy County 1 and 2

·        5 ESBWR      North Anna 3 and Grand Gulf 3*, River Bend 3*, Victoria County 1 and 2*, Fermi 3

  • 4 EPR             Calvert Cliffs 3 , Callaway 2, Nine Mile Point 3, Bell Bend
  • 1 US-APWR Comanche Peak Units 3 and 4

*  The reviews of the FSARs for these COL applications are on hold pending possible selection of another standard design.


NRO Vendor Inspection:


The NRO vendor inspection program is described in Inspection Manual chapter (IMC) 2507, “Construction Inspection Program, Vendor Inspection.”  This IMC will be implemented by various Inspection procedures (IPs) including:


·        IP 43002:  Routine Inspections of Nuclear Vendors;

·        IP 43003:  Reactive Inspections of Nuclear Vendors;

·        IP 43004:  Inspection of Commercial-Grade Dedication Programs;

·        IP 43005:  NRC Oversight of Third Party Organizations Implementing Quality Assurance Requirements; and

·        IP 36100:  Inspection of 10 CFR Parts 21 and 50.55(e) Programs for Reporting Defects and Noncompliance.


NRO Vendor Inspection Reports recently issued and planned inspections


  • Enertech, Brea, CA, September 15-19, 2008 – issued
  • Westinghouse, Monroeville, PA, AP1000 Design Control Document (DCD) and pilot Engineering Design Verification Inspection, October 27 – 31, 2008 - issued
  • General Electric, Wilmington, NC, ESBWR DCD inspection, December 15 - 19, 2008 - completed
  • Dresser Valves, Alexandria, LA, March 9 - 13, 2009 - announced
  • Doosan Heavy Industries, Changwon, Korea, March 30 - April 3, 2009 - announced.


Previously issued NRC inspection and trip reports can be located at



6.    NRC Regulatory Issue Summary 2008-27


On December 8, 2008, the NRC issued Regulatory Issue Summary 2008-27, Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50.  This regulatory issue summary (RIS) was issued to clarify the NRC position concerning licensee requests to extend Type A test (also known as integrated leak rate test or ILRT) intervals beyond the currently approved 15 years under Option B, “Performance-Based Requirements,” of Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,” to Title 10, Part 50, “Domestic Licensing of Production and Utilization Facilities,” of the Code of Federal Regulations (10 CFR Part 50).


Several licensees with approved one-time 15-year Type A test intervals have submitted license amendment requests to the NRC seeking further extensions of the interval for periods ranging from 3 months to 15 months beyond the currently approved 15 years.  Many of these requests did not have proper justification.  This RIS provides the NRC’s expectations concerning Type A test interval extension requests beyond the currently approved performance-based interval of 15 years.


The containment is the final barrier against radioactive release in the event of an accident.  The RIS emphasizes the importance of the discipline licensees should follow in performing the periodic verification of the structural and leakage integrity of the containment within the specified interval.  Accordingly, except for compelling reasons, licensees are expected to conduct the Type A tests within the approved 15-year interval without seeking extensions.


7.    NRC Regulatory Issue Summary 2008-30


On, December 16, 2008, the NRC issued RIS 2008-30, Fatigue Analysis of Nuclear Power Plant Components.  The purpose of the RIS is to inform licensees of an analysis methodology used to demonstrate compliance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) fatigue acceptance criteria that could be nonconservative if not correctly applied.  Title 10 of the Code of Federal Regulations (10 CFR) Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” requires that applicants for license renewal perform an evaluation of time-limited aging analyses relevant to structures, systems, and components within the scope of license renewal.  In addition, the staff has provided guidance in NUREG-1800, Rev. 1, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants,” issued September 2005.  NUREG-1800, Rev. 1, specifies that the effects of the reactor water environment on fatigue life be evaluated for a sample of components to provide assurance that cracking because of fatigue will not occur during the period of extended operation.


The staff identified a concern regarding the methodology used by some license renewal applicants to demonstrate the ability of nuclear power plant components to withstand the cyclic loads associated with plant transient operations for the period of extended operation.  This particular analysis methodology involves the use of the Green’s (or influence) function to calculate the fatigue usage during plant transient operations such as startups and shutdowns. The Green’s (or influence) function methodology is not in question.  The concern involves an input in which only one value of stress is used for the evaluation of the actual plant transients. The detailed stress analysis requires consideration of six stress components, as discussed in ASME Code, Section III, Subsection NB, Subarticle NB-3200.  Simplification of the analysis to consider only one value of the stress may provide acceptable results for some applications; however, it also requires a great deal of judgment by the analyst to ensure that the simplification still provides a conservative result.  The staff has requested that recent license renewal applicants that have used this simplified methodology perform confirmatory analyses to demonstrate that the simplified analyses provide acceptable results.  To date, the confirmatory analysis of one component, a boiling-water reactor feedwater nozzle, indicated that the simplified input for the influence function did not produce conservative results in the nozzle bore area when compared to the detailed analysis.  However, the confirmatory analysis still demonstrated that the nozzle had acceptable fatigue usage.


Licensees may have also used the simplified methodology in operating plant fatigue evaluations for the current license term.  For plants with renewed licenses, the staff is considering additional regulatory actions if the simplified methodology was used.


8.    ASME Letter to NRC – Proposed Interim Staff Guidance for License Renewal


In a letter dated September 26, 2008, from Bryan Erler, Vice President, Nuclear Codes and Standards, ASME, to Brian Holian, Director, Division of License Renewal, Office of Nuclear Reactor Regulation, the ASME requested that the NRC consider issuing interim staff guidance (ISG) on the use of Section XI editions and addenda, relief requests, and Code Cases for license renewal.  The three issues for which guidance was requested are:


  • Use of editions and addenda earlier than those referenced in the Generic Aging Lessons Learned (GALL) Report.


  • Relief request time limits when an inspection interval extends into the license renewal period


  • Use of Code Cases when an inspection interval extends into the license renewal period


In a letter dated January 15, 2009, the NRC responded that a revision to the GALL Report is scheduled to be published in December 2010.  The staff indicated in the letter that it will consider the above issues as a potential LR-ISG and agrees that it would be useful to hold additional discussions.

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