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NRC Section XI Report - April 2008

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

NRC Report
April 2008

1.    Amendments to 10 CFR 50.55a


A proposed amendment to Part10 of the Code of Federal Regulations, Section 50.55a (10 CFR 50.55a) to incorporate by reference the 2004 Edition of the ASME Boiler and Pressure Vessel Code and Code for Operation and Maintenance of Nuclear Power Plant Components was published on April 5, 2007 (72 FR 16731).  The public comment period closed on June 19, 2007.  Responses to the comments have been developed.  It is anticipated that the final rule will be published in May 2008.


2.    ASME Code Case - Rulemaking/Regulatory Guides


The following final regulatory guides (RGs) were noticed in the Federal Register (72 FR 71750) on December 19, 2007:


·        1.84, Revision 34, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III,” [NRC’s Agencywide Documents Access and Management System (ADAMS) No. ML072070407]


·        1.147, Revision 15, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,” (ADAMS No. ML072070419)


·        1.193, “ASME Code Cases Not Approved for Use” (ADAMS No. ML072470294)


The final rule incorporating RGs 1.84 and 1.147 by reference into 10 CFR 50.55a is also available in ADAMS (ML070360713).  The effective date of the rule and hence, RGs 1.84 and 1.147, was January 18, 2008.  RGs 1.84 and 1.147 list the new and revised Code Cases that the NRC has approved or conditionally approved for use.  RG 1.193 lists Code Cases not yet approved for use and is therefore not incorporated by reference into the regulations.


Proposed Revision 35 to RG 1.84, proposed Revision 16 to RG 1.147, and proposed Revision 3 to RG 1.193 are currently in review.  The guides address Code Cases from Supplement 2 through Supplement 0 to the 2007 Edition (Supplement 0, 2007 Edition also serves as Supplement 12 to the 2004 Edition).  The draft guides are expected to be published for public comment in the June/July 2008 time frame.


The NRC staff has nearly completed its review of Supplements 1 and 2 to the 2007 Edition.  There were no nuclear Code Cases in Supplement 3 to the 2007 Edition.


3.    Risk-Informed Activities


Alternative Risk-Informed Inservice Inspection (RI-ISI) Program – Code Case N-716


On September 21, 2007, the NRC conditionally approved an alternative risk-informed inservice inspection program based on Code Case N-716 for Grand Gulf Nuclear Station Unit 1 (the safety evaluation is available in ADAMS [ML072430005]).  On September 28, 2007, the NRC conditionally approved a similar alternative RI-ISI program for Donald C. Cook Nuclear Plants 1 and 2 that is based on Code Case N-716 but deviates and expands upon the Code Case (the safety evaluation is available in ADAMS [ML072620553]).  Since the methodologies deviated from that contained in the Code Case, the safety evaluations only approved the use of the licensee’s proposed methodology and not Code Case N-716.  The Working Group on Risk-Based Implementation has indicated that licensees have expressed great interest in implementing the Code Case.  Accordingly, the staff is considering including the Code Case in draft Revision 16 of Regulatory Guide 1.147 with conditions consistent with the safety evaluations described above.


10 CFR 50.46a - Option 3 Rulemaking (Risk-Informed Emergency Core Cooling System ECCS)


The Advisory Committee on Reactor Safeguards (ACRS) issued a letter on November 16, 2006, recommending that the rule not be issued in its current form.  The letter included three general recommendations: (1) the Rule to risk-inform 10 CFR 50.46 should not be issued in its current form.  It should be revised to strengthen the assurance of defense in depth for breaks beyond the transition break size (TBS); (2) the revision of draft NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process," to include changes resulting from the resolution of public comments, should be completed before the revised Rule is issued; (3) the interpretation that the Rule limits the total increase in core damage frequency (CDF) resulting from all changes in a plant to be "small" (i.e., <10-5/yr) represents a significant departure from the current guidance for risk-informed regulation and should be reviewed for its implications.  NRC staff has provided SECY-07-0082 to the Commission recommending how to proceed with the rulemaking and providing several other options.  The Commission’s August 10, 2007, Staff Requirements Memorandum directed the staff to continue with the rule making but to change the priority from high to medium and provided some addition direction.  The NRC staff is currently developing a schedule to complete this rulemaking.


Reactor Vessel Weld Inspection


Topical Report WCAP-16168-NP Revisions 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," provides the basis for an extension of the reactor vessel weld inspection interval from 10 to 20 years.  The staff has completed its review of this topical report and provided a draft to the PWR Owners' Group.  The staff has resolved all the PWR Owners' Group comments and has placed the final draft of the safety evaluation into concurrence.  The topical report relies extensively on work described in NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)."  The safety evaluation requires licensees utilizing the topical report to use the inspection criteria in the proposed PTS rule, 10 CFR 50.61a, "Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events."


Protection Against Pressurized Thermal Shock Events


On October 3, 2007, the NRC published a proposed change 10 CFR 50.61, “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,” to provide updated requirements for pressurized thermal shock (PTS) events for PWR reactor vessels (72 FRN 56275).  The proposed rule is commonly referred to as 10 CFR 50.61a  The updated technical basis uses many different models and parameters to estimate the yearly probability that a PWR will develop a through-wall crack as a consequence of PTS loading.  These new requirements would be voluntarily utilized by any PWR licensee as an alternative to complying with the existing requirements.  The public comment period closed on December 17, 2007.  The staff has begun developing responses to the comments.  A schedule for the final rule process is under consideration.


Repair and Replacement


In September 2006, the Pressurized Water Reactor Owners Group (PWROG) submitted WCAP-16308-NP, Revision 0, “Pressurized Water Reactors Owners Group 10 CFR 50.69 Pilot Program - Categorization Process - Wolf Creek Generating Station.”  The Topical includes, in part, an alternative methodology to the NRC endorsed Code Case N-660 for categorization of passive components. The PWROG, through NEI, responded to NRC's August 27, 2007, request for additional information on October 22, 2007.  The current review schedule includes an April 7, 2008, public meeting to discuss the staff's draft safety evaluation.


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


Prior to 2005, inspection of dissimilar metal (DM) butt welds was performed in accordance with ASME Code, Section XI, requirements.  In late 2005, the industry implemented an initiative for more aggressive inspection schedules than required by Code.  This initiative is documented in MRP-139, “Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline.”  The base line inspection schedules in MRP-139 are based on temperature and size; for example, pressurizer inspections were required to be completed first.  The on-going inspection frequency is based on the mitigation method applied to a weld (for example, the most frequent inspections would occur in unmitigated pressurizer welds).  The staff is monitoring the implementation of the MRP-139 program for managing PWSCC in DM butt welds.  NRC recently issued Temporary Instruction (TI-172) for regional staff to verify that all PWRs with DM butt welds are implementing MRP-139.


For the longer term, the NRC staff requested that ASME develop a Code Case for inspection of Alloy 82/182 butt welds.  ASME has been actively working on the Code Case and has been responsive to NRC input on the Code Case.  The Code Case is nearing completion.  Based on progress to date, the staff expects that the Code Case will be acceptable for referencing and that the Code Case will be incorporated by reference into 10 CFR 50.55a.


In Oct 2006 inspections were performed at Wolf Creek prior to the pressurizer DM welds being mitigated by application of weld overlays.  Large circumferential flaw indications were found in the Wolf Creek pressurizer safety, relief, and surge nozzle welds.  Based on industry and NRC advanced finite element (AFE) fracture mechanics analyses, NRC staff agreed to industry’s original planned inspection schedules, i.e., some plants inspected their welds up to several months later than called for in MRP‑139.


Recently, two B&W plants experienced PWSCC in decay heat removal to hot leg welds.  The staff evaluated the experience from these plants and concluded that these experiences are within the assumptions upon which the current inspection schedules are based.


Recently a potential safety issue was identified related to inspections of nozzles from a retired pressurizer.  These inspections caused the staff to question whether the initial flaw characterization was inconsistent with the advanced finite element analysis basis for pressurizer inspections scheduled in spring 2008.  Based on quick turn around but more advanced inspection techniques used by industry and witnessed by the NRC staff, the staff concluded that the flaws were fabrication-induced and there was no structurally significant PWSCC in the welds.  The NRC staff continues to monitor and evaluate operating experience to ensure the current inspection schedules are adequate.


5.    New Reactor Licensing Activities


The New Reactor Licensing public web-site [http://www.nrc.gov/reactors/new-reactor-licensing.html] has a list of expected new nuclear power plant applications, and an estimated schedule by fiscal year for new reactor licensing applications.


On April 9, 2008, a public meeting was held in Rockville, MD, to discuss the 10 CFR Part  52 licensing process and the relationship to the ASME Boiler and Pressure Vessel Code for the design and construction of nuclear power plants.  The purpose of the meeting was to promote communication between the NRC and the ASME relative to new reactor licensing activities and to identify potential changes to ASME Code requirements to support the new Part 52 licensing process.  The NRC discussed Part 52 and its relationship to other regulations, such as 10 CFR 50.55a.  The ASME provided an overview of standards actions and task teams related to new reactor developments (including world-wide activities) and ASME Code structure and philosophy.  The ASME has initiated a new joint Section III / Section XI committee, the Task Group on New Reactors, that will hold its first meeting in Vancouver, BC, on April 21, 2008.  The NRC will participate in the committee.


Information Notice


On April 7, 2008, NRC Information Notice 2008-04: Counterfeit Parts Supplied to Nuclear Power Plants, was issued to inform the industry of the potential for counterfeit parts to enter supply chains.


6.    Operating Reactors


Information Notice


On March 19, 2008, the NRC issued NRC Information Notice 2008-02: Findings Identified During Component Design Bases Inspections, to inform addressees of issues identified during recent component design bases inspections (CDBIs) regarding the capability of selected components to perform their design bases safety functions.  The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.


7.    NUREG/CR-6960


NUREG/CR-6960, “Crack Growth Rates and Fracture Toughness of Irradiated Austenitic Stainless Steels in BWR Environments,” was published in March 2008.  Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic stainless steels, including weld heataffected-zone materials, that were irradiated to fluence levels as high as ≈ 2 x 1021 n/cm2 (E > 1 MeV) (≈ 3 dpa) in a boiling heavy water reactor at 288-300°C.  The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined.  The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated.  A fracture toughness trend curve that bounds the existing data has been defined.  The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.


8.    NUREG/CR-6945


NUREG/DR-6945, “Fabrication Flaw Density and Distribution in Repairs to Reactor Pressure Vessel and Piping Welds,” was published in April 2008.  This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping.  Welds from four different reactor pressure vessels and a collection of archived pipes were studied to develop empirical estimates of fabrication flaw densities.  This work indicates that large flaws occur in these repairs, and that the repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities.  A generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors was developed.  The purpose of the generalized flaw distribution is to predict component-specific flaw densities.  The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments.


Construction records were evaluated relative to fabrication processes and product forms used.  An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated.  A description of repair flaw morphology is provided with a discussion of fracture mechanics significance.

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