Federal Regulations, Codes, & Standards
Users Group ©
NRC Section XI Report - August 2006
Presented By: Mr. Wally Norris, United States Nuclear Regulatory Commission
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.
A public meeting was held on Monday, May 15, 2006, during
the ASME meetings in
2. ASME Code Cases - Rulemaking/Regulatory Guides
Three final regulatory guides addressing ASME Code Cases were noticed in the Federal Register on September 29, 2005, (70 FR 56938-56939) - Revision 33 to Regulatory Guide 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III; Revision 14 to Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1; and Revision 1 to Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use.” The guides address the Code Cases in Supplement 12 to the 1998 Edition through Supplement 6 to the 2001 Edition. The guides are available electronically in the Regulatory Guides Document Collection of the NRC's public Web site at
Draft Regulatory Guide DG-1135, "ASME Code Cases Not Approved for Use" (proposed Rev. 2 of Regulatory Guide 1.193) was published for public comment on Tuesday, June 6, 2006 (71 FR 32615). The public comment period closes on July 14, 2006. The proposed rule to incorporate Draft Revision 34 to RG 1.84 (Section III Code Cases) and Draft Revision 15 to RG 1.147 (Section XI Code Cases) has been transmitted to the Executive Director for Operations for signature. The regulatory guides are scheduled to be published for public comment in the next few months.
The staff has completed its review of Supplements 2, 3, 4, 5, and 6 to the 2004 Edition and has begun its review of Supplement 7.
3. Risk-Informed Activities
By letter dated August 10, 2004, the Westinghouse Owners
Group (WOG), now known as the Pressurized Water Reactor Owners Group (PWR
Owners Group), submitted Topical Report (TR) WCAP-14572, Revision 1-NP-A,
Supplement 2, "Westinghouse Owners Group Application of Risk-Informed
Methods to Piping Inservice Inspection Topical Report Clarifications, (WCAP‑14572,
Sup. 2) to the
WCAP-14572, Sup. 2, addresses the following three topics:
1) A methodology for evaluating a segment that includes piping with different diameters (i.e., a multiple pipe diameter (MPD) segment) and for selecting locations for examination as an alternative to the previously approved methodology presented in the approved WCAP-14572.
2) The expert panel decision process for moving a segment that, based on the quantitative results, would normally be high-safety significant (HSS) into the low-safety significant (LSS) segment category.
3) Requirements for examination based on the postulated failure modes and configuration of each piping structural element as an alternative to the previously approved methodology presented in the approved WCAP-14572.
By letter dated May 1, 2006, the NRC staff found WCAP-14572, Supp2, to be acceptable for referencing in licensing applications to the extent specified and under the limitations delineated in the TR and in the final Safety Evaluation (ML0061160035). The final SE defines the basis for our acceptance of the TR. Westinghouse subsequently request two clarifying editorial changes to the staff's SE which staff agreed to and an corrected SE was recently issued.
Risk-Informed ISI - 10 year Interval Updates
Many of the RI-ISI programs approved in the late 1990s and early 2000s were approved before the industry Peer Review process was in place. All RI-ISI programs have been approved as living programs that require feedback of new relevant information. The NRC staff considers the completion of an industry peer review (or other PRA reviews) to be new relevant information. Comments generated by any reviews should also be incorporated as new relevant information (or disposition as not important to the RI-ISI program) before completing the 10‑year Interval Update evaluation. Consequently, licensees that have had the original RI-ISI program approved without addressing the Peer Review or other review observations, may be requested to report on their resolution of the observations during the NRC’s review of the 10‑year Interval Update relief request.
10 CFR 50.69 - Risk Informed Special Treatment Requirements
A Federal Register Notice of the availability of Regulatory Guide 1.201, “Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance,” was published on January 27, 2006 (ADAMS No. ML060260361). Based on a public comment, Revision 1 of Regulatory Guide 1.201 was issued for trial use in May 2006 (ML061090627).
This trial regulatory guide describes a method that the NRC staff considers acceptable for use in complying with the Commissions requirements in §50.69 with respect to the categorization of structures, systems, and components (SSCs) that are considered in risk-informing special treatment requirements. This categorization method uses the process that the Nuclear Energy Institute (NEI) described in Revision 0 of its guidance document NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline."
This trial regulatory guide does not establish any final staff positions, and may be revised in response to experience with its use. This will ensure that the lessons learned from regulatory review of pilot and follow-on applications are adequately addressed in the final regulatory guide, and that the guidance is sufficient to enhance regulatory stability in the review, approval, and implementation of probabilistic risk assessments (PRAs) and their results in the risk-informed categorization process required by §50.69. Trial use is expected to continue through calendar year 2006.
One unresolved issue related to 50.69 is the method to be used to classify passive SSCs primarily for repair and replacement activities. The ASME has developed Code Case N-752, "Risk-Informed Safety Classification and Treatment for Repair / Replacement Activities in Class 2 and 3 Moderate Energy Systems." The NRC staff has not endorsed the methods described in the Code Case.
Request to Extend Reactor Vessel Weld Inspection Interval by One Operating Cycle
Many licensees have submitted requests and been approved to extend the inspection interval for the reactor vessel welds by one operating cycle. The requests referred to NRC guidance provided in a letter from the Nuclear Regulatory Commission to Westinghouse Electric Company, "Summary of Teleconference with the Westinghouse Owners Group Regarding Potential One Cycle Relief of Reactor Pressure Vessel Shell Weld Inspections at Pressurized Water Reactors Related to WCAP-16168-NP, "Risk Informed Extension of Reactor Vessel In-Service Inspection Intervals," dated January 27, 2005. WCAP-16168, Revision 1 was submitted by letter dated January 26, 2006. Based on additional information submitted on June 8, 2006, it has been accepted for NRC staff review.
10 CFR 50.48 Fire Protection
Section 50.48(c), which the Commission adopted in 2004 (69 FR 33536, June 16, 2004), incorporates the 2001 Edition of the National Fire Protection Association (NFPA) standard, NFPA 805, “Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants,” by reference, with certain exceptions. Section 50.48(c) allows licensees to voluntarily adopt and maintain a fire protection program that meets the requirements of NFPA 805 as an alternative to meeting the requirements of 10 CFR 50.48(b) or the plant-specific fire protection license conditions. Licensees who choose to comply with 10 CFR 50.48(c) must submit a license amendment application to the NRC, in accordance with 10 CFR 50.90. Section 50.48(c)(3) describes the required content of the application.
The Nuclear Energy Institute (NEI) developed NEI 04-02, “Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c),” Revision 1, dated September 2005, to assist licensees in adopting 10 CFR 50.48(c) and making the transition from their current fire protection program (FPP) to one based on NFPA 805. Regulatory Guide 1.205, "Risk-informed, Performance-based Fire Protection For Existing Light-water Nuclear Power Plants" was issued in May 2006. This regulatory guide endorses NEI 04-02, Revision 1, as providing methods acceptable to the NRC for implementing NFPA 805 and complying with 10 CFR 50.48(c), subject to the additional regulatory positions contained in Section C of the regulatory guide.
4. Generic Activities on Material Degradation/PWR Alloy 600/182/82 PWSCC
On May 5, 2006, NRC staff met with representatives from NEI and the Electric Power Research Institute (EPRI) Materials Reliability Project (MRP) to discuss NRC staff comments on the industry guidance for the volumetric and visual inspection of dissimilar metal butt welds in pressurized water reactor (PWR) primary systems (MRP‑139). MRP‑139, was approved unanimously by the MRP Executive Committee and was issued to the PWR fleet as a "mandatory" action under the NEI 03‑08 Guideline for the Management of Materials Initiative. The NRR staff previously documented its comments and recommendations on the use of MRP‑139 in a letter dated October 12, 2005, from M. Mayfield (NRC) to A. Marion (NEI). During the meeting, the MRP representatives discussed proposed responses to address the concerns of the NRC staff. Discussions included interactions with the NRC, reporting of inspection findings with the NRC, and Spring 2006 inspection results and implications. MRP representatives indicated that a formal written response to the staff’s concerns documented in the October 12, 2005, letter would be forthcoming. Based on the MRP’s proposed resolution, 17 of 26 staff comments would be addressed satisfactorily.
In a letter dated September 7, 2005, NEI submitted EPRI MRP Report MRP-169, Rev. 0, Technical Basis for Preemptive Weld Overlays for Alloy 82/182 Butt Welds in PWRs, to the NRC for review and approval of the entire report. In a letter dated December 15, 2005, NEI supplemented its original request seeking NRC’s review and approval of only those sections of the report pertaining to design requirements, design methodology, and the examination volumes of full and optimized structural pre-emptive weld overlays, i.e., 4.0, Design Requirement, and 7.1, Requirements for Types of Examination for Weld Overlays. The staff has reviewed MRP-169 and has identified questions and comments which need to be
addressed in order to complete the review and prepare a safety evaluation.
5. New Reactor Licensing Activities
Final Environmental Impact Statement on Exelon ESP to be delivered to EPA
The NRC staff delivered the final environmental impact statement (FEIS) for the Exelon Early Site Permit (ESP) Application to the Environmental Protection Agency at the end of July 2006. One-hundred and sixty members of the public have requested a copy of the document.
Meeting with NEI and Industry to Discuss Infrastructure Development Schedules
On July 24, 2006, the NRC staff met with representatives
from the Nuclear Energy Institute (NEI) and industry. The purpose of the meeting was to discuss
schedules associated with the Standard Review Plan (SRP) Update and
Associated Regulatory Guides, New Reactor Licensing Rulemaking, and combined
operating license (
6. Examination of Cast Austenitic Stainless Steel
A draft NUREG/CR report was completed in December 2005 on the application of an inside diameter surface eddy current (ET) method to detect surface-breaking cracks in centrifugally and statically cast piping. The report has been reviewed by the NRC staff, and PNNL is currently revising the report. The report is scheduled to be published in December 2006.
The ET technique was successful at detecting all open cracks in the Westinghouse Owners Group (WOG) specimens, and provided insights on discrimination of cracks in the presence of geometrical and metallurgical features. Results from this work will be presented to ASME Code committees during the May and August 2006 meetings.
Measurements of coherent ultrasonic noise from the microstructure of vintage centrifugally cast piping segments on loan from Westinghouse, IHI Southwest, Inc. and EPRI is planned. These measurements will help to understand the nominal background noise one might expect from piping in the field, evaluating how to determine these noise properties on field piping, and how this noise may impact detection capability of the technique.
7. Gen IV Reactor Materials Handbook Advisory Committee Meeting
On February 28 - March 1, 2006, RES staff participated in the inaugural meeting of the Generation IV (Gen IV) Materials Handbook Advisory Committee (AdCom). The Gen IV Materials Handbook is being developed to provide an authoritative single source of highly qualified structural materials information and materials properties data for use in design and analyses of all Generation IV Reactor Systems. The Gen IV Materials Handbook AdCom was established to represent the interests of various stakeholders, including the Department of Energy (which oversees the Gen IV program), NRC, codes and standards organizations, and potential Gen IV reactor systems designers and vendors. The Handbook effort will be coordinated with, and will support, related programs, including the Advanced Fuel Cycle Initiative (AFCI), Nuclear Hydrogen Initiative (NHI), and Global Nuclear Energy Partnership (G-NEP). A key issue will be the standards to be used to assess data quality, particularly for data that may be used in licensee submittals. Subsequent meetings are expected to be held approximately twice a year.
8. ASME Standards Technology LLC Project
The Department of Energy (DOE) enlisted ASME Standards Technology, LLC., to update and expand appropriate materials, construction and design codes for application in future Generation (GEN) IV nuclear reactor systems that operate at elevated temperatures. This would be a multi-year project.
The scope of work, including performance measures, milestones, and deliverables is divided into specific tasks, and there will be 3 project teams. A project team will begin developing codes for metallic materials after it receives the results of DOE-funded research, another team will develop rules governing graphite components of the reactor, and a third team has been formed to write rules for in-service inspection.