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NRC Section XI Report - May 2006

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


NRC Report
May 2006

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.

 

There will be a public meeting on Monday, May 15, 2006, from 5:30 - 7:30 pm during the ASME meetings in Phoenix, AZ, to discuss the 2004 Edition.  The purpose of the meeting is to inform industry representatives of those provisions in the 2004 Edition where questions have been raised by NRC staff and solicit comments from stakeholders and other interested parties.  It was reported at the ASME meetings in February that the proposed rule would be published for public comment in summer 2006.  The NRC staff has been working on the endorsement of several safety-significant ASME actions however, which has delayed the proposed rule.  Publication of the proposed rule for public comment is now scheduled for early 2007.

 

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

http://www.nrc.gov/reading-rm/doc-collections/reg-guides/power-reactors/active/

 

Draft Revision 34 to RG 1.84, Draft Revision 15 to RG 1.147, and Draft Revision 2 to RG 1.193 have been approved and are awaiting publication.

 

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 U.S. Nuclear Regulatory Commission (NRC) staff for review and approval.

 

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 has found that WCAP-14572, Supp. 2, is 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.

 

Risk-Informed ISI - 10 year Interval Updates

 

Many of the RI-ISI programs approved in the late 1990's and early 2000's 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 other 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 review observations, may be requested to report on their resolution of these 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).  The guide, also issued for trial use on January 27, 2006, (ML060260164), provided guidance for use in developing and assessing evaluation models for accident and transient analyses.

 

Based on a public comment, Revision 1 of Regulatory Guide 1.201 was issued on April 28, 2006 (ML061090627); Federal Register Notice of Availability (ML061090666).  Based on additional comment, a new version of the RG with some changes to clarify staff intent will be issued in May 2006.

 

This trial regulatory guide describes a method that the NRC staff considers acceptable for use in complying with the Commission’s 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.”  The NRC issued a draft of this guide, Draft Regulatory Guide DG‑1121, as part of the §50.69 rulemaking package in May 2003, and solicited public comments specifically concerning the draft guide.  The public comments were considered in the course of preparing new Regulatory Guide 1.201.

 

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.

 

Request to Extend Reactor Vessel Weld Inspection Interval by One Operating Cycle

 

Five licensees have submitted requests 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.  The three original extension requests have been approved.  The remaining two are under review.

 

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

 

By letter dated December 20, 2005, from James E. Dyer, Director, Office of Nuclear Reactor Regulation (NRR), to Gary C. Park, Chairman, ASME Subcommittee on Nuclear Inservice Inspection (SC XI) [ADAMS No. ML053480359], the NRC staff discussed its position that, given the operating experience with primary water stress corrosion cracking (PWSCC) and the serious challenges posed by potential failure of the reactor coolant piping boundary, it was essential that an appropriate regulatory footprint be established for inspection of dissimilar metal (DM) butt welds.  It was requested that Section XI take the actions necessary to develop the needed improvements to existing Code requirements.

 

By letter dated February 10, 2006, from Alex Marion, Senior Director of Engineering, Nuclear Energy Institute, Nuclear Generation Division, to James E. Dyer, Director, NRR (ML0606107612), NEI provided its views on the December 20, 2005, letter from the NRC to the ASME.  NEI stated that it supported the process being pursued by the ASME, but it would not be appropriate to include the guidance in the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) Report MRP‑139, Rev. 0, “Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline,” in the ASME Code.

 

The NRC staff responded in a letter dated March 2, 2006, from John A. Grobe, Director, Division of Component Integrity, NRR, to Alex Marion, NEI (ML0606501931), that the approach of working with the ASME to revise inspection requirements and subsequently revise 10 CFR 50.55a is necessary, in the best interest of the public health and safety, and publicly transparent.

 

On May 5, 2006, NRC staff met with representatives from NEI and EPRI 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.  Discussions are on-going to resolve the remaining issues.

 

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.  Notice has been sent to interested parties to announce the issuance of the design certification final rule (DCR), dated January 27, 2006, for the Westinghouse AP1000 design.  This final rule amends 10 CFR Part 52 to certify the AP1000 standard plant design.

 

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.  The staff completed the safety evaluation reports for North Anna and Grand Gulf in June and October of 2005.  The staff expects to complete the Clinton FSER in February of 2006.  The environmental reviews are in progress and the staff expects to complete those reviews over the next few months.  The completion of the FSERs and EIS will then be followed by a mandatory hearing on all three proposed ESPs.

 

The staff is currently engaged in certification review of General Electric's ESBWR design.  The design certification application was submitted to the NRC in August 2005, and once accepted for docketing, the certification review is expected to take 48 months.  In addition, the staff is currently engaged with vendors in pre‑application review of several other new and innovative reactor designs, including the EPR design (Framatome ANP), the Pebble‑Bed Modular Reactor (PBMR) design (PBMR Pty Ltd.), the Advanced Candu Reactor (ACR)‑700 design (Atomic Energy of Canada, Ltd.), and the International Reactor Innovative and Secure (IRIS) design (Westinghouse Electric Company).

 

Meeting with Pebble Bed Modular Reactor (PBMR) Pty Ltd.

 

On March 15 and 16, 2006, NRC staff met with PBMR Pty, Ltd. to discuss plant operations, design approach, and response to events for the PBMR.  These discussions are in support of future white papers that PBMR Pty, Ltd. plan to submit to the NRC as part of the pre-design certification application review.

 

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 is being reviewed by the NRC staff.  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.

 

NRC-funded research using low-frequency (500 kHz) phased array ultrasonic testing, as applied from the outside surface, is continuing at PNNL as a potential method to detect cracks in cast piping components.  Early tests with a prototype 500 kHz array showed good results for flaws that are 25-30% through-wall in depth, using simple line scans and steering the ultrasonic beam for inspection angles from 30 to 60 degrees.  A newly designed 500 kHz array with a smaller footprint and enhanced ultrasonic properties is currently being evaluated in the laboratory.  This array uses state-of-the-art piezo-composite elements and allows both beam steering in the active aperture and beam skewing in the passive aperture.  This means that cracks in the WOG specimens having major cross-sections, or branches, that are oriented off-axis to the primary ultrasonic beam may now be better detected and imaged.  Preliminary results show an improved signal-to-noise response with less input gain using the new array.  The smaller footprint of the new array also allows access to more of the WOG specimens with challenging outside surface geometries.

 

Data acquisition on the WOG specimens is on-going, to be followed by measurements of coherent ultrasonic noise from the microstructure of vintage centrifugally cast piping segments on loan from Westinghouse, IHI Southwest, Inc. and EPRI.  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.

 

An international workshop on future directions for the inspection of cast austenitic stainless steel (CASS) was held on Saturday, May 13, 2006, at the conclusion of the EPRI/JRC 5th International Conference on NDE in Relation to Structural Integrity for Nuclear and Pressurized Components.  The workshop was held to review the current state-of-the-art in the inspection of CASS piping, and to discuss what the next steps should be.

 

7.    9th Semiannual Meeting of the Organization for Economic Co-operation and Development (OECD) Piping Failure Data Exchange (OPDE) Program

 

On April 20 and 21, 2006, NRC staff attended the OECD OPDE meeting to present the status of recent U.S. efforts in this area.  Representatives from Canada, Germany, Japan, Korea, Spain, Sweden, and Switzerland also participated.  The purpose of the OPDE Program is to assemble a database of relatively rare, light-water-reactor (LWR) piping failures and precursor events.  These operating experience data are used to develop realistic, pipe failure estimates that can be utilized by OPDE members to improve the safety of their LWRs.

 

8.    Spring Meeting of the Third International Steam Generator Tube Integrity Program (ISG‑TIP)

 

On April 26-27, 2006, NRC staff held a meeting, attended by representatives from Canada, Korea, and the United States.  This international program provides a forum for discussing ongoing research in order to minimize duplication of effort.  Participants made presentations on non-destructive examination, tube integrity analysis, and steam generator corrosion research.  Additionally, the participants discussed the closure of the current 5-year program and what research may be conducted if a fourth ISG‑TIP is pursued.  The research findings from this program are expected to be useful in the NRC staff’s evaluations of licensees’ non‑destructive examination results and in the staff’s review of burst and leakage models used by the industry.

 

9.    23rd International Common-Cause Failure Data Exchange Project (ICDE) Steering Group Meeting

 

On April 25-27, 2006, NRC staff attended a meeting to review progress to date and discuss future activities, including:  updating the common-cause failure (CCF) event data for the six components currently in the ICDE database, updating CCF analyses and insights for several components with emphasis on inspection and maintenance applications, and adding several components for data collection (control rod drive assemblies, level-measurement devices, and heat exchangers).  The results are compared to U.S. events to provide insights for use by staff during inspections.

 


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