United States Nuclear Power

Federal Regulations, Codes, & Standards

Users Group ©


NRC Section XI Report - December 2003

Site Updates


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


NRC Report
Section XI
December 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 and 2003 Addenda. The proposed rule is in final concurrence and is scheduled to be published for public comment in the December 2003/January 2004 timeframe.

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 1.193 have been reviewed by the cognizant offices. The Committee to Review Generic Requirements (CRGR) and Advisory Committee on Reactor Safeguards (ACRS) reviews have been completed. The draft guides are scheduled to be published for public comment in January/February 2004.

The staff is currently reviewing Code Cases in Supplements 7 through 10 to the 2001 Edition. The Section XI Code Cases contained in these supplements will be included in Draft Revision 15 to RG 1.147, which will be initiated in 2004.

3. 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.

On November 20, 2003, members of the NRC staff met with representatives of nuclear utilities and the Nuclear Energy Institute to discuss plans for a pilot program to guide implementation of Draft Regulatory Guide DG-1122, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." DG-1122 describes one acceptable approach for determining that the quality of a PRA (in toto or for those parts that support a specific application) is sufficient to provide confidence in the results such that they can be used in regulatory decision-making for light water reactors. The guide will be issued for trial use in late December 2003. Confirmation of the pilot applications will occur when NRC receives formal letters of intent from the licensees involved. Submittals for the Columbia, San Onofre and Limerick applications could be received as early as January or February 2004. The South Texas and Surry applications are more complex and are expected later in 2004.

The American Nuclear Society (ANS) has published its Standard for External Events PRA. The scope of a PRA covered by this Standard is limited to analyzing accident sequences initiated by external events that might occur while a nuclear power plant is at nominal full power. It is further limited to requirements for (i) a Level 1 analysis of the core damage frequency (CDF) and (ii) a limited Level 2 analysis sufficient to evaluate the large early release frequency (LERF). The scope of a Seismic Margin Assessment (SMA) covered by this Standard is limited to analyzing nuclear power-plant seismic capacities according to either the so-called EPRI method or the so-called NRC method. External events covered within the Standard's scope include both natural external events (e.g., earthquakes, high winds, and external flooding) and human-made external events (e.g., airplane crashes, explosions at nearby industrial facilities, and impacts from nearby transportation activities). Appendix A contains an extensive list of most of the external events generally included within an external-events PRA and hence within the standard's scope. The NRC staff will review the Standard and NRC staff positions developed from the review will then be incorporated in NRC guidance for ensuring the technical adequacy of PRAs, i.e. DG-1122 and SRP 19.1.

The staff updated the RI-ISI Regulatory Guide (1.178) and SRP Section 3.9.8. The updated versions were issued in September 2003. Minimal changes were incorporated.

4. 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.

During the Panel Discussion that closed the Nickel-Base Alloy - Vessel Penetration Conference, which was held in Gaithersburg, MD, from 09/29/03 - 10/02/03, Richard Barrett, Director of the Division Engineering, NRR, stated that NRR was considering rulemaking to address the examination of reactor pressure vessel heads.

On August 21, 2003, the NRC issued NRC BULLETIN 2003-02, "Leakage from Reactor Pressure Vessel Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity." The purpose of the bulletin was to: (1) advise PWR addressees that current methods of inspecting the RPV lower heads may need to be supplemented with additional measures (e.g., bare-metal visual inspections) to detect reactor coolant pressure boundary (RCPB) leakage; (2) request PWR addressees to provide the NRC with information related to inspections that have been or will be performed to verify the integrity of the RPV lower head penetrations, and (3) require PWR addresses to provide a written response to the NRC in accordance with the provisions of Section 50.54(f) of Title 10 of the Code of Federal Regulations (10 CFR 50.54(f)).

On September 5, 2003, the NRC issued Temporary Instruction 2515/152, "Reactor Pressure Vessel Lower Head Penetration Nozzles." The objective of the Temporary Instruction (TI) was to support the review of licensees’ RPV lower head inspection activities that are implemented in response to Bulletin 2003-02. The TI validates that a plant is meeting its inspection commitments using procedures, equipment, and personnel that have been demonstrated to be effective in detecting signs of leakage from the RPV lower head penetration (LHP) nozzles and the detection of RPV lower head degradation.

5. Future Reactor Licensing Activities - Advanced Reactor Infrastructure Assessment

The draft AP1000 Design Certification Safety Evaluation Report was issued in mid-June 2003 with approximately 170 Open Items. Since that time, both the staff and Westinghouse have concentrated efforts on understanding and resolving the Open Items. The NRC’s Office of Nuclear Regulatory Research (RES) is performing confirmatory analyses of the Economic and Simplified Boiling Water Reactor (ESBWR) response to off-normal conditions. On September 4, 2003, a a status report on RES’s independent analysis of the ACR-700 was forwarded to Atomic Energy of Canada, Ltd. Requests for additional information have been transmitted to General Atomics relative to the review of the Gas Turbine Modular Helium Reactor (GT-MHR). The NRC staff has been informally informed by the Pebble Bed Modular Reactor (PBMR) project that it intends to begin pre-application discussions with the NRC by mid-2004 and formal pre-application work in FY2005/6.

RES has initiated an Advanced Reactor Research Program to develop a technology-neutral, risk-informed framework for new reactor licensing and regulatory decisions. The Framework for a Risk-Informed Regulatory Structure for Advanced Reactors will help to ensure that a structured and systematic approach is used during the development of the regulations that will govern the design, construction, and operation of advanced reactors. 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 staff held a public workshop/meeting in November 2003 to discuss the framework.

RES staff has completed draft NUREG, "Regulatory Guidance for Assessing Exemption Requests from Nuclear Power Plant Licensed Operator Staffing Requirements Specified In 10CFR50.54(m)," dated September 2003. The draft NUREG describes the process recommended for reviewing and making decisions on exemption requests from the regulation. The guidance is based on function and task analyses and the anticipated role of the operator, given the concept of operations, as opposed to a prescriptive rule that specifies a fixed number of licensed operators per site. This approach is consistent with Commission direction to use performance-based approaches when feasible. The staff will make the draft NUREG available for public comment within the next two months.

NRC is also supporting meetings sponsored by the International Atomic Energy Agency (IAEA), including various Coordinated Research Projects dealing with advances in fuel technology for high temperature gas cooled reactors and the effects of near-field earthquakes on nuclear facilities. Several international cooperative research agreements have been signed.

6. 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.

7. Meeting on ACR-700 Phenomena Identification and Ranking Technique (PIRT)

On October 30 and 31, 2003, RES conducted PIRT meetings on the Advanced CANDU Reactor design, ACR-700. The PIRT process is used to identify key phenomena and processes that are important to understanding plant behavior under normal and accident conditions. The strength of the PIRT process is in the importance ranking of these phenomena, and in the identification of data and models to support their predictions. The staff will use PIRT information to identify gaps in tools and data to help prioritize their development. The PIRT meetings were the first in a series of expert panel working meetings for applying the process to the ACR-700 review. This initial meeting included introductory technical presentations by the vendor, Atomic Energy of Canada Limited (AECL), and employed technical experts from the NRC, national laboratories, contractors, consultants, and universities. The next PIRT meeting is scheduled for December 11-12, 2003.


Home Page

 

Copyright Disclosure