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NRC Section XI Report - May 2005

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


NRC Report
Section XI
May 2005

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.

 

The staff is completing the technical bases for the amendment to 10 CFR 50.55a to endorse the 2004 Edition.  A schedule has not yet been released for the publication date of the proposed rule.

 

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 Committee to Review Generic Requirements and the Advisory Committee on Reactor Safeguards approved publication of the final guides.  Publication of the rulemaking/guides has been re-scheduled for September 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.

 

The staff is currently reviewing Supplements 1, 2, and 3 to the 2004 Edition.

 

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 descried 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.

 

PRA Steering Committee Public Meeting

 

On April 7, 2005, the Probabilistic Risk Assessment (PRA) Steering Committee met with the representatives from Nuclear Energy Institute (NEI), licensees, and members of the public to discuss issues related to current risk-informed initiatives.  Topics discussed at the meeting included the proposed 10 CFR 50.46 rulemaking, long-term direction of risk-informed applications, configuration risk management, review of topical reports involving generic changes, and Mitigating System Performance Index (MSPI) implementation.  The meeting was well attended by external stakeholders and participants provided useful insights on the discussion topics.  The PRA Steering Committee consists of the directors from the Offices of Nuclear Regulatory Research, Nuclear Reactor Regulation, Nuclear Material Safety and Safeguards, Nuclear Security and Incident Response, and Enforcement; one Regional Administrator; and a representative from the Office of the General Counsel.

 

Risk Management Coordinating Committee (NRMCC)

 

This Committee, which is jointly chaired by the Vice Chairman of the ASME BNCS and the Chairman of the American Nuclear Society Nuclear Standards Committee (ANS), was formed on February 20, 2004.  Its purpose is to coordinate 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.  In addition to ASME and ANS, NRC, several consultants, the Westinghouse Owners Group, and NEI are represented on the committee.  Since the initial organization meeting of February 20, 2004, NRMCC convened via teleconference on May 20, 2004, and met again on September 9, 2004 and March 11, 2005.

 

At the March 11, 2005 meeting, the following were discussed:

 

a)    K. Balkey presented the ASME Nuclear Codes and Standards Initiatives on Risk-Informed Classification for Use in ASME Applications.

 

b)    The current status of the ANS Risk Standards was presented.

 

c)    Status of ASME Risk Standards was presented.

 

d)    An early draft of the proposed NRMCC developed Risk Requirements Standard Charter was reviewed.  A final draft is expected to be available for the June 16, 2005 BNCS meeting.

 

e)    A number of integration issues that will have to be addressed in order to combine existing ASME and ANS Risk Standards into one integrated standard.

 

f)     The status if NRC Risk Informed activities was presented by NRC member M. Drouin.

 

The next NRMCC meeting is scheduled for June 2, 2005 at the Washington D.C. ASME Headquarters.

 

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.

 

Final Report on Good Practices for Human Reliability Analysis Published

 

In April 2005, the Office of Nuclear Regulatory Research (RES) published the final version of NUREG‑1792, "Good Practices for Implementing Human Reliability Analysis (HRA)."  This final NUREG report describes those processes, analytical tasks, and judgments that would be expected in an HRA (considering the state-of-the-art) in order for the results to support risk‑informed decisions.  These good practices were developed as part of the NRC's activities to address probabilistic risk assessment (PRA) quality issues and support the implementation of Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities."  Although, the HRA good practices were developed in the context of a risk assessment for nuclear power plant full power operations, some elements of the report may be applicable to other modes of plant operation or nuclear materials applications.  A copy of the publically available report can be obtained through the Agencywide Documents and Access Management System (ADAMS) or at the NRC's website (www.nrc.gov).

 

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

 

Representatives from the ASME and NRC staff met at NRC headquarters on March 22, 2005, to discuss the remaining NRC staff comments on Code Case N-729.  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 concerns remain relative to several provisions.  A letter signed by Michael Mayfield, Director, Division of Engineering, Office of Nuclear Reactor Regulation, dated April 26, 2005, was transmitted to Gary Park, Chair, Subcommittee XI, reiterating the staff’s comments on the proposed Code Case.

 

Representatives from NEI and MRP met with the NRC staff at NRC headquarters on March 24, 2005, to discuss inspections for PWSCC in ASME Class 1 nickel-based components, other than vessel upper head penetrations and steam generator tubes. The NRC staff has decided to prepare a generic letter to address inspections and potential mitigative actions applicable to dissimilar metal butt welds in Class 1 components.  Regulatory action by the NRC staff on the remaining nickel-based alloy components is under consideration.

 

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, Clinton, and Grand Gulf.  All three applications were accepted for docketing in late 2003, and the staff’s safety and environmental reviews of the applications are in progress.  The reviews are expected to take 23, 25, and 27 months, respectively, followed by a mandatory hearing.

 

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.

 

Public Workshop on "Regulatory Structure for New Plant Licensing, Part 1:Technology‑ Neutral Framework"

 

On March 14-16, 2005, the staff from the Offices of Nuclear Regulatory Research (RES), Nuclear Reactor Regulation (NRR), and Nuclear Security and Incident Response (NSIR) held a public workshop to solicit comments on the, "Regulatory Structure for New Plant Licensing, Part 1: Technology-Neutral Framework."  The framework and the associated technology-neutral requirements are intended as the first steps in formulating the technical basis for future rulemaking for technology-neutral regulations for new plant licensing.  The workshop was well attended and included representatives from the American Society of Mechanical Engineers, the American Nuclear Society, the Nuclear Energy Institute, Westinghouse Electric Company, General Atomics, Framatome, the International Atomic Energy Agency, several power reactor licensees, the Federal Emergency Management Agency, and four Department of Energy Laboratories (Brookhaven, Idaho, Lawrence Livermore, and Sandia National Laboratories).

 

The workshop participants agreed that the concepts in the framework are reasonable and that it is feasible to develop both a technology-neutral framework and technology-neutral requirements.  The participants encouraged the staff to continue with similar workshops in the future to further explore the technical issues.  In addition, participants recommended performing a test-case and encouraged expanding the schedule to include the rulemaking phase.  The overall goals of the regulatory structure for new plant licensing include the development of a process for identifying safety issues uniquely associated with reactor technologies that are not addressed by current NRC regulations and providing stability and predictability for licensing of these new reactor technologies.

 

European Pressurized Reactor

 

The NRC met with representatives of Framatome ANP (FANP), Inc. on March 24, 2005 to discuss plans for the European pressurized reactor (EPR) pre-application review.  FANP, a subsidiary of AREVA and Siemens, presented a high level overview of their plans for submitting a Design Certification Document by the end of 2007.  During the pre-application Phase 1, which should continue until December 2005, FANP has stated their intent to engage the staff in several meetings to discuss topics related to the EPR design such as Design Codes and Standards, Probabilistic Risk Assessment, and Safety Analysis.  For Phase II of the proposed pre-application, which is expected to start in January 2006, FANP will request the NRC to review three topical reports on Critical Heat Flux Correlations, Transient and Accident Analysis Code Applicability Report (including fuel mechanical design), and Severe Accident Evaluation. FANP's goal is to deploy EPR plants in the U.S. and is in the process of staffing a dedicated U.S. organization committed to preparing a high quality design certification.  The central task driving the length of the pre-application period is the engineering work required for U.S. design conversion from European standards and the development of a high-quality submittal.

 

The EPR is considered by FANP to be an evolutionary, not revolutionary, reactor design.  FANP  indicated that it currently has support from the Duke energy utility and is looking to solidify wide-spread support from domestic utilities in the U.S.  The EPR design is a 4-loop pressurized water reactor (PWR) with active safety injection systems and is similar to current operating PWRs in the U.S.  However, it has several unique design features such as a double-walled containment, four independent safety injection trains in separate buildings, and a corium spreading area and in-containment reactor water storage tank.  Construction is under way in Finland to build an EPR at the Olkiluoto nuclear power site and is scheduled to go online sometime in 2009.

 

Documents related to future licensing activities can be found on the NRC web site at, http://www.nrc.gov:201/NRC/GENACT/GC/index.html.

 

6.    Agreement on the OECD-NEA Studsvik Cladding Integrity Project (SCIP)

 

On December 16, 2004, the Executive Director for Operations signed the multilateral international Agreement regarding NRC’s participation in the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency’s new Studsvik Cladding Integrity Project (SCIP).  Sixteen (16) additional organizations in ten (10) countries will participate in this research program.

 

This project will provide experimental data that are needed for improving the understanding of the dominant cladding failure mechanisms under normal operation and transient conditions in light water reactor (LWR) fuels.  Four basic phenomena will be addressed: (1) pellet‑cladding mechanical interaction, (2) iodine stress‑corrosion cracking, (3) hydrogen embrittlement, and (4) delayed hydride cracking.  This will be accomplished by testing irradiated fuel in a hot cell and a test reactor.  Special arrangements have been made to provide specimens from some of these fuel rods to Argonne National Laboratory (ANL) for further testing.  These will include de-fueled lengths of Westinghouse ZIRLO cladding from high-burnup fuel rods irradiated in North Anna.  It should be noted that these will be the first advanced cladding specimens from high-burnup fuel made available to NRC for testing.

 

The research conducted and the fuel rod specimen test data obtained under this Agreement will help NRC develop fuel damage limits that are used in licensing reviews.

 

7.    NUREG/CR-6861, Barrier Integrity Research Program - Final Report

 

In December 2004, the Office of Nuclear Regulatory Research (RES) published NUREG/CR-6861 entitled, "Barrier Integrity Research Program - Final Report."  Argonne National Laboratory (ANL) conducted a comprehensive review of leakage operating experience by developing a database of leakage events, an evaluation of the capabilities of the current leakage monitoring systems, and an identification of new systems that are potentially more capable than current systems.  RES initiated the ANL program as a result of Davis-Besse lessons-learned task force (DBLLTF) recommendations.  This NUREG/CR will be used by an RES and Office of Nuclear Reactor Regulation working group which is deliberating on improvements to reactor coolant system unidentified leakage requirements.

 

8.    Meeting at Electric Power Research Institute (EPRI) Non-Destructive Examination (NDE) Center

 

On January 11, 2005, Office of Nuclear Regulatory Research (RES) management and staff met with staff of the Electric Power Research Institute (EPRI) Non-Destructive Examination (NDE) Center to identify opportunities to coordinate research efforts and avoid unnecessary overlaps.  The EPRI steering committee of member utilities recently authorized a significant increase in funding for research to improve the reliability and speed of in-service inspections.  The NRC staff described the Program for the Inspection of Nickel Alloy Components (PINC), an international cooperative research group being formed to assess NDE techniques for detecting primary water stress corrosion cracking (PWSCC) in components made of Alloy 600 and related materials.  EPRI will be participating in the PINC program.

 

9.    Public Meeting on the Mitigating Systems Performance Index (MSPI)

 

On March 16, 2005, staff from the Offices of Nuclear Regulatory Research (RES), Nuclear Reactor Regulation (NRR) and the Regions participated in a public meeting on the Mitigating Systems Performance Index (MSPI).  The purpose of the workshop was to address MSPI technical and implementation issues resulting from the first industry workshop held in February of this year.  Although a majority of the major technical issues were substantially resolved during the meeting, industry representatives did not come to agreement on certain issues related to PRA quality.  Industry had previously made a proposal on PRA quality to staff and is currently re-evaluating its position.  The staff will continue to work with industry and public stakeholders to resolve potential changes to the original proposal.  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.  Industry-wide implementation of the MSPI is scheduled to begin in January 2006.

 

Publication of NUREG-1816, “Independent Verification of the Mitigating Systems Performance Index (MSPI) Results for the Pilot Plants”

 

On March 21, 2005, the Office of Nuclear Regulatory Research published NUREG-1816, "Independent Verification of the Mitigating Systems Performance Index (MSPI) Results for the Pilot Plants."  An electronic version of this NUREG was previously made publicly available through ADAMS in February of this year.  The primary goal of the research described in this report was to provide independent verification of the results of the MSPI pilot program.  On the basis of this research, the staff concluded that, in general, the MSPI is capable of differentiating risk-significant changes in plant performance and addresses some problems associated with the safety system unavailability performance indicators currently used in the reactor oversight process.  This research activity supported the development of the MSPI and ongoing staff efforts to implement the MSPI on an industry-wide basis.


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