United States Nuclear Power

Federal Regulations, Codes, & Standards

Users Group ©

NRC Section XI Report – November 2008

Site Updates

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

NRC Report
November 2008

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).  Code Cases N–722, ‘‘Additional Examinations for PWR [pressurized water reactor (PWR)] Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1,’’ and N–729–1, ‘‘Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section XI, Division 1,’’ were also incorporated by reference but with conditions.  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.


It should be noted that the original version of the rule published on the NRC’s public website (Electronic Reading Room / Document Collections / FRNs / Regulations) was incorrect.  The correct version is now available.  In addition, a correction to paragraph10 CFR 50.55a(g)(6)(ii)(D)(4)(ii) will be published soon.  This paragraph addresses the NDE qualification for the reactor vessel head penetration inspection.  Specifically it requires UT specimens to contain certain number of flaw and flaw sizes.  The last sentence of the paragraph states that at least 20 % and no more than 40% of the flaws shall be oriented axially.  The sentence should have read "at least 20% and no more than 60% of the flaws shall be oriented axially".


The U.S. Nuclear Regulatory Commission (NRC) began a Lean Six Sigma Project on May 27, 2008, for the purpose of reducing the rulemaking process cycle time to incorporate edition and addenda of the ASME Code and OM Code into 10 CFR 50.55a.  A briefing providing recommendations for improving the 10 CFR 50.55a rulemaking process was given to the Deputy Executive Director (DEDO) for Reactor and Preparedness Programs on October 28, 2008.  There will be a presentation by NRC staff summarizing the scope of the project and the recommendations provided to the DEDO at the ASME Boiler Code meetings on Tuesday, November 11, 2008, at 5 p.m.


2.    ASME Code Case Rulemaking/Regulatory Guides


Proposed Revision 35 to RG 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III,” proposed Revision 16 to RG 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,”and proposed Revision 3 to RG 1.193 “ASME Code Cases Not Approved for Use” are currently in review.  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 draft guides are expected to be published for public comment in December 2008.


The NRC staff has completed its review of Supplements 1, 2, and 4 to the 2007 Edition (there were no nuclear Code Cases in Supplement 3).  Supplements 5 and 6 are under review.


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.


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 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.




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.  The staff held two public meetings and plans to issue a final version in late 2008.


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.


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


Prior to 2005, inspection of dissimilar metal (DM) butt welds was performed in accordance with ASME Code, Section XI, requirements.  In late 2005, the industry implemented an initiative for more aggressive inspection schedules than required by Code.  This initiative is documented in MRP-139, “Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline.”  The base line inspection schedules in MRP-139 are based on temperature and size; for example, pressurizer inspections were required to be completed first.  The on-going inspection frequency is based on the mitigation method applied to a weld (for example, the most frequent inspections would occur in unmitigated pressurizer welds).  The staff is monitoring the implementation of the MRP-139 program for managing PWSCC in DM butt welds.  NRC recently issued Temporary Instruction (TI-172) for regional staff to verify that all PWRs with DM butt welds are implementing MRP-139.


For the longer term, the NRC staff requested that ASME develop a Code Case for inspection of Alloy 82/182 butt welds.  ASME has been actively working on the Code Case and has been responsive to NRC input on the Code Case.  The Code Case is nearing completion.  Based on progress to date, the staff expects that the Code Case will be acceptable, possibly with comments/conditions, and that the Code Case will be incorporated by reference directly into 10 CFR 50.55a (as opposed to through Regulatory Guide 1.147).


In Oct 2006 inspections were performed at Wolf Creek prior to the pressurizer DM welds being mitigated by application of weld overlays.  Large circumferential flaw indications were found in the Wolf Creek pressurizer safety, relief, and surge nozzle welds.  Based on industry and NRC advanced finite element (AFE) fracture mechanics analyses, NRC staff agreed to industry’s original planned inspection schedules, i.e., some plants inspected their welds a few months later than called for in MRP-139.


Last spring, two B&W plants experienced PWSCC in decay heat removal to hot leg welds.  The staff evaluated the experience from these plants and concluded that these experiences are within the assumptions upon which the current inspection schedules are based.


In February 2008, a potential safety issue was identified related to inspections of nozzles from a retired pressurizer.  These inspections caused the staff to question whether the initial flaw characterization was inconsistent with the advanced finite element analysis basis for pressurizer inspections scheduled in spring 2008.  Based on quick turn around but more advanced inspection techniques used by industry and witnessed by the NRC staff, the staff concluded that the flaws were fabrication-induced and there was no structurally significant PWSCC in the welds.  This was confirmed via destructive analysis.


The NRC staff continues to monitor and evaluate operating experience to ensure 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 dissimilar metal (DM) butt welds containing Alloy 82/182 in pressurized-water reactor (PWR) power plants.


As discussed in the RIS, the NRC staff has concluded that MRP-139 and the MRP interim guidance letters, with the exception of the reinspection interval for unmitigated pressurizer DM butt welds as addressed by the NRC staff issued Confirmatory Action Letters (CALs), provide adequate protection of public health and safety for addressing PWSCC in butt welds for the near term.  The incorporation of Code Case N-722 (see the first item on page 1 of this NRC Report) contains visual inspection requirements that apply to butt welds and partial penetration welds made with Alloy 82/182 weldment.  The NRC staff indicated in the RIS that it would continue to monitor the industry’s MRP-139 inspections and operating experience and will use this information to determine if any additional regulatory actions are necessary.


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 is evaluating the significance of this inspection finding.


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 November 10, 2008, 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 (25 new reactor units):

  • 1 ABWR          South Texas Project 3 and 4

·         6 AP1000        Bellefonte 3 and 4, William S. Lee Station, 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 (AR), Victoria County 1 and 2 (AR), Fermi 3 (AR)

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

(AR) = In acceptance review.


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


  • Energy Steel & Supply, Lapeer, MI, July 21-25, 2008 -- issued
  • B&W Power Generation Group, Inc., Mount Vernon, IN, August 25-29, 2008 - issued
  • Enertech, Brea, CA, September 15-19, 2008 – completed
  • Westinghouse, Monroeville, PA, October 27 – 31, 2008 - completed


Previously issued NRC inspection and trip reports can be located at



6.    High Density Polyethylene Piping (HDPE)


By letter dated October 31, 2008 (ML082640007), from Michael T. Markley, Chief Plant Licensing Branch IV, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation, to Adam C. Heflin, Senior Vice President and Chief Nuclear Officer, Union Electric Company, the NRC approved the use of high density polyethylene pipe (HDPE) in lieu of carbon steel pipe in buried essential service water (ESW) system at the Callaway Plant, Unit 1.  The relief request authorizes the use of HDPE pipe for the buried section of the ESW system for Callaway's third 10-year lSI interval, scheduled to end on December 18, 2014.  All other requirements of the ASME Code, Section XI for which relief has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

By letter dated September 12, 2008 (ML082240386), from Melanie C. Wong, Chief, Plant Licensing Branch II-1, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation, to J. R. Morris, Site Vice President, Catawba Nuclear Station, Duke Energy Carolinas, LLC, the NRC approved the use of HDPE for the Catawba Nuclear Station, Units 1 and 2.  Specifically, the relief approves the use of HDPE pipe in lieu of carbon steel piping in the Class 3, 12-inch nominal diameter supply and return buried piping to and from the diesel generator jacket water coolers.  The relief is for the third 10-year inservice inspection interval at each unit.  There are a total of 8 lines for which the licensee is seeking approval for the use of HDPE pipe.  The NRC staff has concluded that the licensee has provided sufficient information to approve the use of HDPE pipe for the 1A and 2A supply lines to the diesel generator jacket water coolers.  However, the NRC staff has identified that additional information is needed to complete its entire review of this relief request.  By letter dated September 8, 2008 (ML082480700), the NRC issued a request for additional information concerning supply lines 1B, 2B and return lines 1A, 1B, 2A, and 2B.


7.    SECY-08-0140, Development and Regulatory Applications of Consensus Standards


SECY-08-0140, “Development and Regulatory Application of Consensus Standards by U.S. Nuclear Regulatory Commission Staff,” was forwarded to the Commission on September 24, 2008 [ML081260303].  The purpose of the paper is to inform the Commission of the staff’s ongoing activities to develop, promulgate, and endorse voluntary consensus standards in the NRC’s regulatory framework, as a means of increasing the efficiency and effectiveness of the regulatory process. The paper does not propose any changes to current policy, nor does it address any new commitments or resource implications.


8.    Multi-national Design Evaluation Program (MDEP) Codes and Standards Working Group Meeting


On October 29-30, 2008, NRC staff participated in a meeting of the MDEP Codes and Standards Working Group (CSWG) at the offices of the Autorité de Sureté Nucléaire (ASN) in Dijon, France.  The purpose of the meeting was to discuss the preliminary results of the Code-comparison project being performed by standards development organizations (SDOs) of several member countries.  The main objectives of the MDEP/CSWG Code-comparison project are to identify similarities and differences between Codes used in different countries for the design and construction of pressure-boundary components in nuclear power plants and to examine the potential for convergence of code requirements.


The MDEP/CSWG meeting was attended by representatives from OECD/NEA and IAEA and regulators from Canada, Japan, France, Korea, Finland, Russia, China and the U.S.  The SDOs were represented by Code members from France (AFCEN), Korea, (KEA), Japan (JSME), and the U.S. (ASME).  The SDOs presented examples of major differences between their Code and the ASME Boiler and Pressure Vessel Code, Section III and the reasons for their differences.  Each SDO presented the status of its Code-comparison activity, discussed examples of Code differences, and described how the significance of Code differences would be addressed.  The SDOs noted that the Code-comparison table will discuss the source of the differences (e.g., regulatory, technical, industry practice, etc.) and will address specific reasons for the differences.  The first phase of the Code-comparison project will compare Code requirements for ASME Code Class 1 vessels.  In the next phase, the SDOs will include other Class 1 components (i.e., piping, pumps and valves).


At this time, the first phase of the Code-comparison project for KEA’s Korea Electric Power Industry Code (KEPIC) is complete.  The completion of the first draft of JSME’s Code-comparison table for its light water reactor code (S-NC1) is expected to be finished before December 2008.  The first draft of AFCEN’s RCC-M Code is expected to be complete by the end of 2008.  The SDOs will compile the results by January 2009 into a draft Code-comparison table.  The final Code-comparison document for the first phase is scheduled to be issued in May 2009.


9.    Office of Nuclear Regulatory Research (RES) Moving


Between the dates of November 14 and 25, 2008, NRC staff from RES will be moving from their present location at Two White Flint North (TWFN) to 21 Church Street, Rockville, Maryland.  The Church Street location is within walking distance of the Rockville Metro Station on the Red Line (two stops past TWFN).  RES staff will be receiving new phone numbers in addition to having new mail stops.  The Office of New Reactors will be moving into the vacated RES space.

Home Page


Copyright Disclosure