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NRC Section XI Report - August 2003

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

NRC Report
Section XI
August 2003

1. Amendments to 10 CFR 50.55a

The final amendment to 10 CFR 50.55a incorporating the 1997 Addenda through 2000 Addenda by reference was published on September 26, 2002 (67 FR 60520) and is available at the Office of Federal Register website: http://www.access.gpo.gov/su_docs/fedreg/a010803c.html

The NRC staff has completed its formal review of the changes to the ASME Code incorporated into the 2001 Edition/Addenda and the 2002 Addenda. NRC rulemaking staff will be available to provide a status briefing for Section XI members on August 25, 2003, in Scottsdale at the Phoenician.

2. ASME Code Cases - Rulemaking/Regulatory Guides

Four regulatory guides were published final in June 2003: Revision 32 to Regulatory Guide 1.84, "Design, Fabrication, and Materials Code Case Acceptability, ASME Section III; Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability," ASME OM Code; Revision 13 to Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1; and Regulatory Guide 1.193, "ASME Code Cases Not Approved for Use." The guides address Supplement 4 to the 1992 Edition through Supplement 11 to the 1998 Edition, and OMN-1 through OMN-13. The ADAMS numbers are: RG 1.184: ML030730417; RG 1.147: ML030730423; RG 1.192 (OM Guide):ML030730430; RG 1.193 (Unapproved code cases): ML030730440; and the response to public comments: ML030730448. The final rule accompanying the guides was published on July 8, 2003 (68 FR 40469) and became effective on August 7, 2003.

The NRC staff has completed its review of the Code Cases in Supplement 12 to the 1998 Edition through Supplement 6 to the 2001 Edition. Draft Revision 33 to RG 1.84, Draft Revision 14 to RG 1.147, and Draft Revision 1 to RG 1.19 are in concurrence. The draft guides are scheduled to be published for public comment November 2003.

3. ASME PRA Standard

Draft Regulatory Guide DG-1122, "An Approach for Determining the Technical Adequacy of PRA Results for Risk-Informed Activities," provided the NRC’s proposed positions on consensus PRA standards and industry PRA documents. ASME and NRC staff have been working to resolve differences on ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications." As a result, the ASME will publish an addendum to the standard in September 2003; the NRC has eliminated many of the proposed conditions in the draft guide. The guide is scheduled to be published in the Fall 2003 for trial use.

A pilot study of the current standard was completed in June 2003. Several deficiencies were identified which will be published in a future revision to the standard.

4. Risk-Informed Activities

The NRC website contains information at http://www.nrc.gov/what-we-do/regulatory/rulemaking/risk-informed/reactor-safety.html relative to risk-informed activities in the Reactor Safety Arena.

The staff met with NEI and other industry representatives in November 2002 to discuss "living program" guidance for risk-informed inservice inspection (RI-ISI) programs. RI-ISI programs must be updated to reflect changes in the plant, and general industry experience that might impact the program. The schedule and extent of such changes is under development. NEI stated that it intended to complete guidance on this process and provide it to the staff for review and comment by the Spring of 2003. The guidance has not yet been submitted and as discussed below, events may overtake the guidance.

The staff has received three ten-year update relief requests from licensees with RI-ISI programs. One relief request review is nearing completion. This licensee did not change the methodology but modified the number of inspections and the locations of inspections between the third and the fourth ten year intervals. After several interactions between the licensee and the staff, the licensee provided a change in risk estimate based on comparing the ASME Section XI program in place before implementation of the original RI-ISI program and the RI-ISI program proposed for the fourth interval.

The staff is in the process of updating the RI-ISI Regulatory Guide (1.178) and SRP Section 3.9.8 and plans to issue the updated guidelines in the next few months. Minimal changes are needed or planned.

5. Generic Activities on PWR Alloy 600/182/82 PWSCC

On February 11, 2003, an Order titled, "Issuance of Order Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors," was issued (EFFECTIVE IMMEDIATELY). Many of the licensees' responses to Bulletin 2002-02 did not describe long-term inspection plans nor a schedule to submit such descriptions. The Order requires specific inspections of the reactor pressure vessel (RPV) head and associated penetration nozzles at pressurized water reactors (PWRs). It has been determined that the current RPV head visual inspection requirements are not sufficient to reliably detect circumferential cracking of RPV head nozzles and corrosion of the RPV head. The Order establishes interim requirements to ensure that licensees implement and maintain appropriate measures to inspect and, as necessary, repair RPV heads and associated penetration nozzles. The long-term resolution of this issue is expected to involve changes to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) and will involve changes to the NRC regulations in 10 CFR 50.55a, "Codes and standards." Licensee responses to the order and other related documents are provided on the website for each operating reactor, grouped by NRC region:

On August 12, 2003, the NRC's Restart Oversight Panel held two public meetings in Oak Harbor, Ohio. The first of these meetings was a public meeting with the licensee to discuss the licensee's performance and progress on their Return to Service Plan. The second was a meeting with the public where the Panel discussed the status of the Oversight Panel activities and responded to questions and concerns from the public. The next meeting is scheduled for September 10, 2003.

The NRC staff has completed all Special Inspection activities, and all regulatory requirements reviews supporting the restart and continued operation of South Texas Project, (STP) Unit 1 and continued operation of STP Unit 2. In a letter to the licensee, dated July 31, 2003, the NRC concluded that STP, Unit 1, complies with all existing regulatory requirements necessary to support the restart of STP, Unit 1, and that the operation of STP, Units 1 and 2, is consistent with the licensee’s obligation to protect the health and safety of the public.

6. Future Reactor Licensing Activities - Advanced Reactor Infrastructure Assessment

The staff issued the draft safety evaluation for the AP1000 design certification in June 2003. The final safety evaluation is expected to be issued in mid-2004. Pre-application review for the ESBWR, ACR-700, and GT-MHR designs will continue, with additional pre-application reviews for the SWR 1000 and IRIS designs starting.

The Office of Nuclear Regulatory Research has developed a 5-year Advanced Reactor Research Program. Most of the major activities are scheduled to be completed by the end of 2007. The objectives of this program are to: Develop an infrastructure of methods, tools, data, and expertise needed to support the certification of advanced reactor designs; Provide the technical bases for regulatory decisions; Support pre-application and design certification reviews; and Develop options and recommendations for advanced reactor policy issues. The program plan also includes the development of a technology-neutral, risk-informed framework for new reactor licensing and regulatory decisions.

7. Risk-Informing 10 CFR Part 50 (Option 2)

The comment period for a proposed rule entitled, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (RIN 3150-AG42)," was to have expired on July 30, 2003. By letter dated July 3, 2003, Nuclear Energy Institute (NEI) requested a 30-day extension to the comment period. In view of the importance of both the proposed rule and the industry's comments on it, the NRC decided to extend the comment period by 30 days as requested. The comment period expires on August 30, 2003. Comments received after this date will be considered if it is practical to do so, but the Commission is able to ensure consideration only for comments received on or before this date. The proposed rulemaking is available for review at http://ruleforum.llnl.gov/cgi-bin/library?source=*&library=SSC_PRULE_lib&file=*&st=prule.

The proposed rule would provide an alternative approach for establishing the requirements for treatment of structures, systems and components (SSCs) for nuclear power reactors using a risk-informed method of categorizing SSCs according to their safety significance. The proposed amendment would revise requirements with respect to "special treatment," that is, those requirements that provide increased assurance (beyond normal industrial practices) that SSCs perform their design basis functions. This proposed amendment would permit licensees (and applicants for licenses) to remove SSCs of low safety significance from the scope of certain identified special treatment requirements and revise requirements for SSCs of greater safety significance. In addition to the rulemaking and its associated analyses, the Commission is also proposing a draft regulatory guide to implement the rule.

8. Implementing Agreement Between the USNRC and the Japan Atomic Energy Research Institute (IAEII) in the Field of Nuclear Reactor Safety Research

The NRC’s Office of Nuclear Regulatory Research (RES) and the Japan Atomic Energy Research Institute (JAERI) are developing an Implementing Agreement to perform cooperative research in the Field of Nuclear Reactor Safety Research. Under this Implementing Agreement, the USNRC and the JAERI will exchange technical information in the areas of (1) thermal-hydraulic code development, (2) severe accidents, (3) plant aging, and (4) safety of high-burnup fuel for light water reactors. Analyses and test data that will be made available to RES will include 1) experiments for advanced light water reactors with passive systems, 2) high temperature gas reactor (HTGR) neutronics, 3) seismic and structural information conducted using their high temperature engineering test reactor (HTTR) design, and 4) high-temperature metallic components and nuclear graphite.

9. Pebble Bed Modular Reactor Workshop

On August 7, 2003, technical staff from the Offices of Nuclear Regulatory Research and Nuclear Reactor Regulation, the US Department of Energy and other interested US government personnel participated in a one-day workshop on the Pebble Bed Modular Reactor (PBMR). The USDOE-sponsored workshop was held at DOE headquarters in Germantown, MD, and involved presentations by PBMR representatives from the Republic of South Africa. The purpose of the seminar was to provide a technical and programmatic update of the PBMR project. The topics included: the status of PBMR program; the design of the PBMR demonstration plant to be built at the Koeberg nuclear plant site in South Africa; the fuel, core design and thermal hydraulic design; and the component testing program, including the fuel qualification testing program. The fundamental "safety case" licensing framework for the plant design has been accepted by the licensing authority in South Africa and the four stage licensing review for the demonstration plant is currently being implemented. The schedule anticipates licenses to be issued for construction, equipment installation, fuel loading and operation in July 2005, December 2005, May 2008 and July 2009, respectively. The PBMR project schedule also involves the resumption of PBMR pre-application activities with the NRC in mid-2004 and a PBMR design certification application by the end of 2006.

10. NRC PRA Steering Committee Meeting with Nuclear Energy Institute (NEI)

On August 13, 2003, a joint meeting of the NRC PRA Steering Committee with the NEI risk-informed regulation working group was held. There was a brief summary of a CSNI/CNRA workshop on the redefinition of the large-break loss of coolant accident (LOCA) which was held in Zurich June 23-24, 2003. The general consensus of the Zurich meeting participants was that there were good reasons for such a redefinition, and that such a redefinition had the potential to both enhance safety and reduce the costs of nuclear power generation. Moreover, the participants believed that an adequate technical basis for the redefinition could be developed.

There was considerable discussion of the rulemaking on the redefinition of the design basis LOCA, and the Commission direction to require a high quality, full scope PRA including low power and shutdown, external events, and level 2 analysis; NEI indicated the difficulty in meeting this requirement. Some of the potential uses of the redefined design basis large break LOCA were identified by the industry.

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