Federal Regulations, Codes, & Standards
Users Group ©
NRC Section XI Report - March 2005
Presented By: Mr.
1. Amendments to 10 CFR 50.55a
A final amendment to 10 CFR 50.55a incorporating by reference the 2001 Edition up to and including the 2003 Addenda of the ASME Boiler and Pressure Vessel Code, was published on October 1, 2004 (69 FR 58804). The rule became effective on November 1, 2004.
2. ASME Code Cases - Rulemaking/Regulatory Guides
Three draft regulatory guides were published for public comment on August 3, 2004: Proposed Revision 33 to Regulatory Guide 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III [ADAMS Accession Number ML040850299]; Proposed Revision 14 to Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 [ML040850346]; and Proposed Revision 1 to Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use [ML040850236].” The guides address Code Cases in Supplement 12 to the 1998 Edition through Supplement 6 to the 2001 Edition. The Proposed Rule incorporating the Section III and Section XI regulatory guides was also published on August 3, 2004 [69 FR 46452]. The “Evaluation of Code Cases,” attached to the proposed rule, discussed the staff’s basis for any proposed conditions to Code Cases. The public comment period for the regulatory guides closed on September 2, 2004, and on October 18, 2004, for the proposed rule.
The guides and the responses to public comments are currently being reviewed by the cognizant offices and the Office of the General Counsel. The office reviews are scheduled to be completed in mid-March 2005. Reviews by the Committee to Review Generic Requirements and Advisory Committee on Reactor Safeguards would then take place in early April. Publication of the guides is scheduled for July 2005.
Draft Revision 34 to RG 1.84, Draft Revision 15 to RG 1.147, and Draft Revision 2 to RG 1.193 are nearly complete. The guides address Code Cases in Supplement 7 through Supplement 12 to the 2001 Edition. These guides will be published for public comment shortly after the guides discussed in the above paragraph are published final.
3. Risk-Informed Activities
Plan for Phased Approach to PRA Quality Drafted
The objective of the phased approach to stabilizing the PRA quality expectations and requirements is to achieve an appropriate level of PRA quality for NRC’s risk‑informed regulatory decision making. That is, the phased approach defines the needed PRA quality for all envisioned applications and the process for achieving this quality while the necessary guidance documents defining the PRA quality are developed and implemented. It is expected that meeting the phased approach objective will result in the following:
· Industry movement towards improved and more complete PRAs
· Increased efficiencies in the staff’s review of risk‑informed applications
· Clarification of expectations for 10CFR50.46 and 10CFR50.69 rulemakings
· Continued near‑term progress in enhancing safety through the use of available risk‑informed methods while striving for increased effectiveness and efficiency in the longer term.
The plan describes the phased approach and what activities, on the part of both NRC and industry, are needed to achieve the program objectives. In addition, the action plan discusses the resolution of the following technical issues: model uncertainty; treatment of seismic and other external events; and human performance issues. Phase 1 represents the current situation, where guidance on PRA quality is general, and staff review of the base PRA supporting the activity is performed on a case-by-case basis. Phase 2 occurs when there are PRA standards and the associated regulatory guides in place to address those PRA scope items that are significant to the decision. To be in Phase 2 for an application, the licensee's submittal is expected to be in conformance with the published standards. The PRA standards are being developed on different schedules. As a result, the risk-informed activities will transition to Phase 2 on different schedules according to which scope items are significant to the decision. Phase 3 provides a regulatory framework for the development of a PRA that will be of sufficient quality to support all current and anticipated applications. Phase 3 will be completed by December 31, 2008. A fourth Phase may be developed at a later time.
The plan was issued to the Commission on July 13, 2004, and approved in an SRM dated October 6, 2004. The SRM also approved the staff recommendations that Phase 2 applications will be acceptable after Phase 3 guidance is in place, but once Phase 3 guidance is in place, Phase 1 applications would no longer be acceptable.
The staff has initiated the activities necessary to implement the plan. The American Society of Mechanical Engineers (ASME) has published ASME RA-S-2002, “Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications” (April 5, 2002), and Addenda A to this standard (ASME RA-Sa-2003, December 5, 2003). The staff issued Regulatory Guide 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities 1.200 For Trial Use” (February 2004). Regulatory Guide 1.200 endorsed the standard with some clarifications and qualifications as described in Appendix A of the Regulatory Guide.
In mid‑November 2003, the ANS standard ANSI/ANS‑58.21, "American National Standard - External Events in Probabilistic Risk Assessment Methodology," was issued. The staff developed Appendix C to Regulatory Guide 1.200 to endorse ANSI/ANS-58.21. The staff issued Appendix C for public comment on August 31, 2004. The 60-day public comment period expired on October 29, 2004. On November 9, 2004, the NRC staff conducted a meeting to solicit public comments on proposed staff positions on ANSI/ANS-58.21. The staff expects to publish Appendix C to Regulatory Guide 1.200 in final form in the June to August 2005 time frame.
The NRC website contains information at
relative to risk‑informed activities in the Reactor Safety Arena.
Meeting with ASME and Industry Representatives on Proposed Code Cases - N-720 (Section III), and N‑660-1, N-660-2, N-711, and N-716 (Section XI)
A meeting was held on February 10, 2005, between the staff and representatives from the industry and ASME to discuss technical issues raised by the staff relative to each of the Code Cases. The purpose of the meeting was to open a dialogue on the key NRC technical issues early in the consensus standards development process.
With regard to Code Case N-720, “Risk-Informed Safety Classification for Construction of Nuclear Facility Components,” Section III, Division 1, SECY-05-0006, “Second Status Paper on the Staff’s Proposed Regulatory Structure for New Plant Licensing and Update on Policy Issues Related To New Plant Licensing,” was issued on January 7, 2005. Attachment 1 is a draft NUREG entitled, “Regulatory Structure for New Plant Licensing, Part 1: Technology-Neutral Framework Working Draft Report.” The purpose of the draft NUREG is to discuss an approach, scope, and acceptance criteria that could be used to develop a technology-neutral set of requirements for future plant licensing. At the present time, the material contained in the draft NUREG is preliminary and does not represent final staff positions on the issues discussed. As such, certain sections of this document are incomplete and are planned to be completed following receipt of initial stakeholder feedback. The work represented in this document is, however, considered sufficiently developed to illustrate one possible way to establish a technology-neutral approach to future plant licensing and to identify the key technical and policy issues to be addressed. In this regard, a public workshop on the framework has been scheduled for March 14-16, 2005, in the NRC Two White Flint North Auditorium. The contact is Amarjit Singh, 301-415-0250.
Risk‑informing Special Treatment Requirements of 10 CFR Part 50 ‑ Option 2
Final 10 CFR 50.69, "Risk‑Informed Treatment of Structures, Systems, and Components" was approved by the Commission on October 7, 2004, subject to the changes denoted during the affirmation session and documented in the SRM (ADAMS ML042810516). The staff revised the final rule package per the Commission direction, and the final rule was published in the federal register on November 22, 2004 (69 FR 68008).
The final rule can be accessed at http://ruleforum.llnl.gov/cgi-bin/rulelist?type=final.
Risk‑informing 10 CFR 50.46, ECCS ‑ Option 3
The staff has prepared a proposed rulemaking to add a new section to 10 CFR Part 50 to provide an alternative, risk-informed set of requirements for emergency core cooling systems. The Commission SRMs of March 31, 2003, on SECY-02-0057, and July 1, 2004, on SECY-04-0037, provided guidance to the staff on the content of the alternative risk-informed ECCS rule.
The proposed rule would divide the current spectrum of loss of coolant accident (LOCA) break sizes into two regions. The division between the two regions is delineated by a “transition” break size (TBS). The first region includes small breaks up to and including the TBS. The second region includes breaks larger than the TBS up to and including the double ended guillotine break (DEGB) of the largest reactor coolant system pipe. The proposed rule could be voluntarily adopted by light-water reactor licensees. The TBS could be used, for certain purposes, in lieu of the double-ended rupture of the largest pipe in the reactor coolant system (DEGB). Licensees would still need to demonstrate the plant’s ability to mitigate breaks up to the DEGB.
Pipe breaks in the smaller break size region are considered much more likely than pipe breaks in the larger break size region. Consequently, each region will be subject to ECCS requirements commensurate with the relative likelihood of breaks in each region. LOCAs in the smaller break size region will continue to be “design-basis accidents” and will continue to be analyzed by current methods, assumptions, and criteria. In the design-basis region, licensees must perform analyses under current ECCS requirements to determine the limiting size and location for breaks up to and including the TBS.
Pipe breaks larger than the TBS, based on their lower likelihood, can be analyzed by the more realistic and less stringent methods established in the new §50.46a. Although LOCAs for break sizes larger than the transition break will become “beyond design-basis accidents,” the NRC will include requirements ensuring that licensees maintain the ability to mitigate all LOCAs up to and including the DEGB of the largest reactor coolant system pipe. Although these breaks would be required to be mitigated, the analysis methods and initial and boundary conditions used may be realistic. Licensees would be allowed to take credit for sufficiently reliable non-safety-related systems without assuming other independent failures and must show that the core remains amenable to cooling. The specific metrics for demonstrating "coolable core geometry" are not necessarily limited to a peak cladding temperature of 2200 degrees F and 17 percent local cladding oxidation as required for breaks smaller than the TBS. Licensees would be able to propose other criteria for assuring coolable core geometry if an adequate technical basis was also provided to support the proposed criteria.
The Staff plans to send the proposed risk-informed ECCS rule to the Commission in March 2005.
Risk Management Coordinating Committee (NRMCC)
The NRMCC coordinates codes and standards activities associated with risk management for current and new nuclear power plants, nuclear facilities, and the transportation and storage of nuclear material. At the meeting on September 9, 2004, the NRMCC voted to: endorse a strategy for the development of a single PRA standard for requirements that would include: 1) at-power and low power/shutdown internal events - Level 1 and (LERF) (including internal fire) and external events (including external flooding, seismic, and wind); 2) a level of detail consistent with ASME Standard RA‑S‑2002; 3) requirements in the single standard based on requirements in existing standards (ASME - internal-events - full power PRA and ANS - external - events - full power PRA) or standards now in the final stages of development (ANS - low power shutdown (LPSD) PRA, and ANS - internal fire PRA). Guidance on how to perform a PRA for the scope covered by the single requirements standard would continue to be provided in separate guides.
Also at the September 9 meeting, NRMCC voted to form a Joint Consensus Committee reporting to either the ANS Nuclear Standards Board or the ASME BNCS and proceed with the development of the single PRA standard. Responsibility for the standard would be assigned to a single society (ASME or ANS) with support, recognition, and revenue-sharing by the other Society.
Since the September 9 meeting, two Task Groups have been working on different aspects of developing the single PRA standard. One TG is developing a ASME/ANS PRA requirements standard (level 1 internal/external events and fire at power and shutdown). This TG is composed of the ASME Committee On Nuclear Risk Management (CNRM) augmented by ANS Risk-Informed Standards Committee members and reports to the ASME BNCS. A second TG is developing a standard addressing PRA Level 2 and 3 PRAs. For this effort ANS has the lead and reports through the ANS Nuclear Standards Committee. Ultimately the two documents will be merged into a single PRA standard.
The next meeting of the NRMCC is scheduled
for March 11, 2005, at the ASME’s offices in
4. Generic Activities on PWR Alloy 600/182/82 PWSCC
Preliminary comments were submitted to the ASME Task Group on Alloy 600/82/182 Issues by the NRC staff on October 18, 2004. The staff is very supportive of the industry’s efforts regarding the development of Code inspection requirements and criteria for upper vessel heads and penetrations but expressed concerns with the format of the first draft of the Code Case. Many of these concerns were addressed during the December 2004 meetings in the extensive changes to the Code Case.
The revised Code Case was letter balloted in January
2005. The staff submitted additional
concerns regarding some of the changes made and reiterated a few concerns. The comments will be discussed during the
Section XI Task Group on Alloy/600/82/182 meeting on February 28, 2005, in
Public Meeting with
On January 18, 2005, Office of Nuclear Regulatory Research (RES) staff met in a public meeting with staff from Argonne National Laboratory (ANL) and contractor representatives of the Materials Reliability Program (MRP), an industry program funded by the Electric Power Research Institute (EPRI). The ANL scientists and MRP contractors presented summaries of their respective programs on the corrosion of reactor structural materials in simulated pressurized water reactor coolant of various concentrations and at various temperatures and exposure durations. Discussion addressed the significance of the finding that high ratios of lithium and boron markedly reduce corrosion rates. The data generated by both programs will be included in the upcoming revision of the EPRI Boric Acid Corrosion Guidebook, scheduled for completion in 2007. The results of the two programs will help inform future regulatory actions regarding inspection of pressure boundary materials exposed to boric acid solutions.
5. New Reactor Licensing Activities
The final safety evaluation report (FSER) and Final Design Approval (FDA) for the AP1000 design certification were issued on September 13, 2004. The staff is scheduled to complete rulemaking to certify the AP1000 design in December 2005.
The staff is reviewing early site permit (ESP)
applications for three sites: North Anna,
Pre-application reviews for the ESBWR (New Simplified Boiling Water Reactor by General Electric), ACR‑700 (Advanced CANDU Reactor by Atomic Energy of Canada Limited), and IRIS (International Reactor Innovative and Secure by Westinghouse Electric Company) designs will continue. One topic being addressed in the ACR-700 pre‑application review is unique features that may impact ASME Code Sections III and XI. The design uses materials and fabrication techniques recognized by the Canadian Standards Association but not by the ASME code. Design certification review of the ESBWR and ACR‑700 are expected to start in 2005 and 2006, respectively. A design certification application for IRIS is also possible in 2006.
On August 30, 2004, the Office of Nuclear Regulatory Research (RES) and the Office of Nuclear Reactor Regulation (NRR) staff met with Mr. Edward Wallace, Senior General Manager ‑ US Programs, PBMR (Pebble Bed Modular Reactor) Pty. LTD, Republic of South Africa (RSA). The purpose of the meeting was to update the staff on the status of the PBMR project, including the PBMR plant design and the company’s plan to: (1) conduct preapplication interactions with the staff on selected key technical topics beginning in April 2005; and (2) submit the PBMR design for NRC certification in the first quarter of CY 2007. According to PBMR Ltd, construction on the demonstration PBMR plant at the Koeberg site, RSA, will begin in April 2007 and operation is envisioned to commence in 2010.
The staff held a meeting with
PBMR Ltd. on November 3, 2004, to discuss PMBRs plan for pre-application
interactions, submitting the PBMR design certification application, and
The Nuclear Regulatory Commission staff will hold its 17th
annual RIC Tuesday, March 8, Wednesday, March 9, and Thursday, March 10,
2005, at the Bethesda North Marriott Hotel and
6. First Industry Workshop on the Mitigating Systems Performance Index (MSPI)
On February 9 and 10, 2005, staff from the Offices of Nuclear Regulatory Research (RES), Nuclear Reactor Regulation (NRR) and the Regions participated in the first of three public workshops on the MSPI. The workshop was widely attended and included representatives from Institute of Nuclear Power Operations (INPO), Nuclear Energy Institute (NEI), and all seventy power reactor licensees. The purpose of the workshop was to introduce the industry participants to the MSPI, provide training, and prepare licensees for industry‑wide implementation of the MSPI. RES and NRR staff also participated in a panel discussion regarding the PRA quality requirements for MSPI implementation. The MSPI is a more risk‑informed performance indicator that addresses many of the concerns with the current Safety System Unavailability performance indicator used as part of the Agency’s Reactor Oversight Process. The second workshop is scheduled for June 2005.