Federal Regulations, Codes, & Standards
Users Group ©
NRC Section XI Report - February 2004
Presented By: Mr.
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.
A proposed rule to amend 10 CFR 50.55a to incorporate by reference the 2001 Edition through 2003 Addenda of the ASME Code was published in the Federal Register on January 7, 2004 (69 FR 879). Public comments must be submitted by March 22, 2004.
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
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 approved for publication by the cognizant offices. The proposed rule and draft guides should be published for public comment in April 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 February 5, 2004, Nuclear Regulatory Commission (NRC)
staff met with representatives from Nuclear Energy Institute (NEI) and
industry at the NRC’s office in
On February 19, 2004, the ACRS Subcommittee for Reliability and PRA will review the resolution of public comments on the proposed 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components," and the staff's draft Regulatory Guide endorsing NEI 00-04, "10 CFR 50.69 Structures, Systems, and Components Categorization Guideline." On March 25, 2004, the staff will brief the subcommittee on the status of risk-management technical specification initiatives. On April 1, 2004, the staff brief the subcommittee on its draft action plan for implementing a phased approach to PRA quality, called for in staff requirements memorandum (COMNJD-03-2003). On April 15-17, 2004, the full committee will discuss all three of these activities.
A public meeting with stakeholders was held on January 22,
2004, to discuss pilots of the PRA quality guidance, Regulatory Guide 1.200,
and SRP 19.1. Industry was represented
by NEI and representatives from the five pilot application licensees: San
Onofre, Surry, South Texas Project, Limerick, and
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 1.200, “An Approach for Determining the Technical Adequacy of
Probabilistic Risk Assessment Results for Risk-Informed Activities.” RG 1.200 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 March 2004.
Confirmation of the pilot applications will occur when NRC receives
formal letters of intent from the licensees involved. Submittals for the
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. RG‑1.200 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 January 6, 2004, NRC, ASME, and utility representatives met at NRC headquarters to discuss: the status of Task Group on Alloy 600/182/82 Issues activities, status of Task Group on Boric Acid Corrosion activities, the use of mechanical nozzle seal assemblies (MNSAs) as a Section XI repair method, and the ASME nuclear codes and standards process and the management of emerging issues.
Following are the highlights of the meeting:
· The NRC reiterated its desire to rely on the ASME Code in general, and for the ASME to address emerging issues relative to components within the scope of the ASME Code under most circumstances.
· Because organizations outside of the ASME are often involved in addressing emerging issues (e.g., EPRI Materials Reliability Project), a process was developed for addressing future correspondence between the NRC and ASME such as requests to initiate new activities or status of on-going efforts.
· Currently, the ASME has no formal mechanism for notifying Code users of emerging issues. ASME staff recognize the need for such a mechanism and are considering various options.
· The NRC expressed its desire for the ASME to develop a “fast-track” for the Code to address emerging issues.
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 NRC staff and Westinghouse have concentrated efforts on understanding and resolving the Open Items. As of mid-February 2004, all major thermal-hydraulic issues have been resolved, and steady progress is being made towards completion of a draft Final Safety Evaluation Report by mid-2004. 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 NRC 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.
6. Organizational Meeting of Nuclear Risk Management Coordinating Committee
On February 20, 2004, the initial meeting of this committee was held at