United States Nuclear Power

Federal Regulations, Codes, & Standards

Users Group ©


NRC Section XI Report - November 2007

Site Updates


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


NRC Report
November 2007

1.    Amendment to 10 CFR 50.55a – ASME Code Edition/Addenda

 

A proposed amendment to 10 CFR 50.55a that would 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 are being developed.  The final rule is scheduled for publication in January 2008.

 

2.    ASME Code Case - Rulemaking/Regulatory Guides

 

A final rule to incorporate Revision 34 to Regulatory Guide 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III,” and Revision 15 to Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,” into Part10 of the Code of Federal Regulations, Section 50.55a (10 CFR 50.55a) by reference was signed by the Executive Director of Operation on November 1, 2007.  The final rule will be noticed in the Federal Register and will be effective 30 days thereafter.  The final regulatory guides listed above have also been approved and will be noticed in the Federal Register on the same day as the rulemaking.  The effective date of the guides is governed by the final rule.

 

Final Revision 2 to Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use,” has also been approved and is available on the NRC’s regulatory guide website.  No public comments were received on Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use,” Revision 2.  However, changes to this guide were made as a result of public comments received on Regulatory Guide 1.84; specifically, Code Case N-659.  As discussed in the Federal Register Notice (71 FR 62947, dated October 27, 2006), the NRC staff proposed conditional approval of N-659.  Public comments were transmitted expressing concern with a number of the proposed conditions.  The issues are complicated and addressing them is not straightforward.  Accordingly, the NRC has decided to work with ASME International to develop acceptable performance criteria on the use of ultrasonic/radiographic testing and will not endorse N-659 or Revision 1 to the Code Case at this time.  There are several Code actions under development that may be affected, i.e, BC04-247, Code Case N-713, “Use of Ultrasonic Examination in Lieu of Radiography, Section XI, Division 1,” and BC06-1092, “IWA-4520, revise to permit use of Section XI personnel qualifications, methods, and criteria for repair/replacement activities.”

 

Proposed Revision 35 to Regulatory Guide 1.84 and proposed Revision 16 to Regulatory Guide 1.147 are under development.  The guides will include Code Cases from Supplement 2 through Supplement 0 to the 2007 Edition (also considered Supplement 12 to the 2004 Edition).  The draft guides are expected to be published for public comment in Spring 2008.

 

3.    Risk-Informed Activities

 

Regulatory Guide1.200, An approach for Determining the Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities

 

A Federal Register Notice of availability of Revision 1 of Regulatory Guide (RG) 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,” was published on February 8, 2007 (ADAMS No. ML070240001).  The RG describes one acceptable approach for determining whether the quality of a probabilistic risk assessment (PRA), in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors.  NRC Issued Regulatory Issue Summary (RIS) 2007-06 on March 22, 2007 (ADAMs No. ML070650428), discussing NRC expectations on implementing RG 1.200.  The RIS states, in part, that the staff will review routine, limited scope applications (e.g., single allowed outage time extensions, risk-informed inservice inspection, integrated leak rate testing extensions) using its current practices through December 2007.  For all risk-informed applications received after December 2007, the NRC staff will use Revision 1 of RG 1.200 to assess technical adequacy.

 

10 CFR 50.69 - Risk Informed Special Treatment Requirements

 

A Federal Register Notice of the availability of RG 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).  In September 2006, the PWR owners group 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 is intended, in part, to simplify 50.69 applications by providing a template for the contents of the categorization process results and descriptions that should be included in a license amendment request to implement 50.69.  The NRC is reviewing the topical, including a determination of the extent that a standard template format can be developed and endorsed (e.g., quality of the PRA is not addressed).

 

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

 

The Topical report WCAP-16168-NP Rev 1, Risk-informed Extension of the Reactor Vessel In-Service Inspection Interval, requesting an extension of the weld inspection interval from 10 to 20 years is under review.  A meeting at NRC to discuss draft requests for information was held on May 30, 2007.  The topical report relies extensively on work described in NUREG-1874, “Recommended Screening Limits for Pressurized Thermal Shock (PTS)” which the NRC intends to publish in the near future (ADAMs No. ML070740639).

 

Protection Against Pressurized Thermal Shock Events

 

On October 3, 2007, the NRC published a proposal to change 10 CFR 50.61 to provide updated requirements for pressurized thermal shock (PTS) events for PWR reactor vessels (72 FRN 56275).  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.  Public comments should be submitted by December 17, 2007.

 

Repair and Replacement

 

In September 2006, the PWR owners group 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 NRC is reviewing this Topical.

 

On July 11, 2007, a public meeting was held between NRC staff and industry representatives to discuss the details of the Title 10 of the Code of Federal Regulations (10 CFR) 50.69 passive categorization process described in Topical Report (TR) WCAP-16308-NP, “Pressurized Water Reactor Owners Group 10 CFR 50.69 Pilot Program - Categorization Process – Wolf Creek Generating Station.”  The meeting summary is available in ADAMS (ML072010313).  TR WCAP-16308-NP is available in ADAMS (ML062770345).  The meeting was held to discuss several items arising from a May 17, 2007, NRC staff audit (ML071640104) of the 10 CFR 50.69 pilot application documentation, related to the ongoing review of the passive categorization process described in TR WCAP-16308-NP.  The NRC staff opened the meeting by stating that a discussion of issues arising out of its audit and TR WCAP-16308-NP would support the issuance of a request for additional information (RAI) to NEI.  The industry representatives stated that their expectation for the meeting was to clarify the manner in which the American Society of Mechanical Engineers (ASME) Code Case N-660, “Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,” requirements were satisfied during the Wolf Creek pilot categorization and to clarify the intent of any proposed changes to N-660.  Although the primary purpose of the meeting was to discuss the 10 CFR 50.69 passive categorization process described in the TR, there were a few questions regarding the treatment of structures, systems, and components (ADAMS Accession No. ML071930256).  At the meeting conclusion, the NRC staff and industry representatives reached agreement regarding which comments would be formalized as RAIs.

 

A follow-up meeting was held on September 28, 2007, (summary at ML072770519) for the industry to discuss its proposed responses to the NRC staff’s request for additional information (RAI) dated August 27, 2007 (ML072220129).  The NRC letter dated August 27, 2007, contained a total of eleven RAI questions.  The Nuclear Energy Institute (NEI) committed to provide RAI responses to the NRC by October 18, 2007.  The September 28, 2007, meeting was specifically held to ensure that the proposed responses would adequately address the NRC staff’s concern on the passive categorization guidance contained in TR WCAP-16308-NP.  During the meeting, the NRC staff also identified an additional RAI that will require a response before the NRC staff can complete its review.

 

A licensee submitted a relief request under 50.55a(3)(i) to authorize the use of a risk-informed safety classification and treatment for repair/replacement activities in Class 2 and Class 3 moderate energy systems.  Additional information, including a pilot application was received by the NRC on August 6, 2007, requesting completion by April 17, 2008.  The staff has begun its review.

 

During the review of the periodic, 10-year updates of the RI-ISI program, the NRC must develop confidence that the living program requirements are being appropriately implemented using a current PRA of technical adequacy.  The potential impact of the recently issued RG 1.200 on PRA quality on RI-ISI relief requests is under discussion and the staff is working together with NEI and EPRI to assess whether it is feasible to provide more directed PRA quality guidance that could be used in support of RI-ISI updates.

 

ASME Addenda B (2005) To Standard RA-S-2002

 

On July 18, 2007, a public meeting was held between NRC staff and industry representatives to present its proposal for satisfying the PRA quality guidelines for risk-informed inservice inspection (RI-ISI) programs developed using ASME Code Case N-716 “Alternative Piping Classification and Examination Requirements, Section XI Division 1.”  The meeting summary is in ADAMS (ML072130491).  Code Case N-716 has not yet been endorsed for use by the NRC but has been used in two pilot applications under review by the NRC staff.

 

The industry opened the meeting by providing a draft report (Enclosure 2 in ADAMS) that includes a table with all the flooding elements from the ASME standard and the capability category descriptions for each element.  Each element in the table also includes industry’s proposed assessment of which capability category would satisfy the quality requirements for a flooding analysis in support of a RI-ISI program developed according to Code Case N-716.  The NRC staff indicated that the industry proposed approach appears to be reasonable based on the major improvements in the flooding analysis elements in Addendum B to the ASME standard.  The staff indicated that it would determine acceptable capability categories based, at a minimum, on consistency with the analyses found acceptable in the two endorsed RI-ISI methodologies and with RG 1.178, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: For Inservice Inspection of Piping” (ML032510128).  Comparing the industry’s proposed acceptable capability categories with the analysis required by the endorsed documents, the staff did not agree that the proposed capability categories for some elements were sufficient (i.e., the staff recommended that a higher capability category should be required).  An industry final white paper evaluating the staff comments will be submitted to the NRC when it is complete.

 

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

 

The circumferential indications identified in three dissimilar metal (DM) welds on the pressurizer at the Wolf Creek Generating Station raised safety concerns based on the size and location of the indications.  These findings also raised concerns regarding the adequacy of the MRP-139, “Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline,” baseline inspection schedule for pressurizer welds, particularly the deferral of the baseline inspections allowed by the industry’s NEI 03-08, “Guideline for the Management of Materials Issues,” protocol.  The pressurizer surge nozzle-to-safe end weld indications are of concern, as this is the first time that multiple circumferential indications have been identified in this weld.  This condition calls into question the degree of safety margin present in past structural integrity evaluations for flawed DM welds, since multiple stress-corrosion cracking flaws may grow independently and ultimately grow together, significantly reducing the time from flaw initiation to leakage or rupture.  The size of the relief nozzle-to-safe end flaw is also of concern, as this flaw has a much larger aspect ratio than those assumed in the estimates used to establish the basis for the industry-sponsored MRP-139 program.  Larger aspect ratios could result in achieving a critical flaw size and rupture prior to the onset of detectable leakage.  A number of significant meetings have been held on these issues in 2007.

 

Nine licensees with spring 2008 refueling outages had committed to inspect the welds by the end of 2007 if an adequate level of safety from an industry finite element analysis program was not demonstrated to the NRC.  By letter dated February 14, 2007, the Nuclear Energy Institute (NEI) indicated that the Electric Power Research Institute Materials Reliability Program would be undertaking a task to refine the crack growth analyses pertaining to the Wolf Creek pressurizer DM weld ultrasonic indications.  The purpose of the additional analyses was to address the NRC staff’s concerns regarding the potential for rupture without prior evidence of leakage from circumferentially oriented PWSCC in pressurizer nozzle welds.  Industry completed these analyses and documented the results in MRP-216, Revision 1, “Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in Pressurizer Nozzle Dissimilar Metal Welds: Evaluations Specific to Nine Subject Plants,” Final Report Draft A, dated July 10, 2007, (ML071970073), which is also available from the EPRI web site.

 

NRC staff met with industry on August 9, 2007, to provide members of the public with an opportunity to obtain an overview of the entire project.  NRC staff discussed the confirmatory analysis it had performed.  A supporting draft report entitled, “Evaluation of Pressurizer Alloy 82/182 Nozzle Failure Probability (Including Effect of the Fall-06 Wolf Creek NDE Indications), by Structural Integrity Associates, Inc., dated July 14, 2007, is also available in ADAMs (ML071970083).

 

The NRC staff has evaluated the MRP report and documentation provided by the industry (safety assessment is available at ML072470394).  The principal conclusions resulting from this safety assessment are as follows.

 

  • Based on operating experience with circumferentially and axially oriented indications of PWSCC, the NRC staff does not believe that the pressurizer nozzle DM welds are seriously degraded but if circumferential flaws were to occur, the NRC staff does not expect such flaws to grow to rupture without exhibiting leakage.

 

  • For the safety, relief, and spray nozzles, the results of the analyses showed crack arrest for all cases based on weld residual stresses from the finite element calculations.  Through-wall crack growth resulted from analyses based on conservatively applying an ASME Code residual stress.  The limiting cases from these analyses demonstrated acceptable safety factors.

 

  • Analyses of the surge nozzle sensitivity cases showed that safety factors for almost all cases met the NRC staff specified safety factors.  Quantitative evaluation of conservatisms inherent in those few sensitivity cases with low safety factors demonstrates that all surge nozzle analyses have adequate factors of safety.

 

  • All nine plants have adequate safety factors for non-normal thermal loading.

 

  • The NRC staff has therefore concluded that there is reasonable assurance that the nine plants addressed by this evaluation can operate safely until their next scheduled refueling outages in Spring 2008.

 

The NRC staff is developing a temporary instruction (TI) on dissimilar metal butt welds.  TIs are inspection procedures used by regional inspectors.  The objective of this TI is to verify that licensees are implementing mitigation and inspection programs consistent with MRP-139.  This TI is planned to take effect early in 2008.  [Also see Number 8. below]

 

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 updated April 27, 2007, and an estimated schedule by fiscal year for new reactor licensing applications.

 

6.    Visual Testing NUREG

 

NUREG/CR-6943, A Study of Remote Visual Methods to Detect Cracking in Reactor Components,” has been posted to the NRC public website.  The study concludes that a significant fraction of the cracks that have been reported in nuclear power plant components are at the lower end of the capabilities of the VT equipment currently being used.  The study also suggests that inspection conditions need to be nearly ideal to detect these cracks.  The research has shown that the use of visual inspections in lieu of ultrasonic testing must be carefully considered taking into account factors such as surface conditions and expected crack opening displacement.

 

7.    Treatment of Operational Leakage

 

On August 30, 2007, a meeting was held between NRC staff and industry representatives to discuss the development of NRC staff and industry guidance related to operational leakage in ASME Code Class pressure boundary components.  The meeting summary is available in ADAMS at ML072540808.  This project was undertaken to address industry concerns raised in the NEI White Paper on Operational Leakage provided to the NRC in May 2006 (ML061320347).  NEI provided Revision 1 to this white paper in October 2006 (ML063250490) and Revision 2 in May 2007 (ML071590195).

 

The NRC interim guidance indicates that American Society of Mechanical Engineers (ASME) Code Class components with through wall leakage in high energy piping systems should be declared inoperable immediately.  The draft revision to Part 9900 indicates that it is the NRC staff view that for ASME Class 2 moderate or high energy (HE) components and Class 3 HE components with identified through-wall leakage, the staff considers that it may not be feasible to make an immediate operability determination that a reasonable expectation of operability exists.  Industry representative discussed some scenarios that could occur in ASME code Class 2 components and Class 3 HE components to illustrate their view that such an immediate operability determination may be feasible.

 

As a result of discussions, industry representatives indicated that they had a better understanding of the NRC staff view and offered to prepare draft industry guidance to address the NRC staff concerns in this area.  NRC staff indicated it would review NEI comments in meeting handouts and consider revising the draft Part 9900 Appendices.  During this meeting, NRC staff and industry representatives also discussed use of the ASME Code and ASME Code Cases for evaluation of structural integrity and for structural integrity acceptance criteria for pressure boundary components with through-wall leakage.

 

NRC staff met with representatives from the industry on October 18, 2007, to continue discussions regarding industry concerns with Regulatory Issue Summary (RIS) 2005-20, “Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, “Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and Operability,” as delineated in NEI White Paper dated May 2, 2006 (ML061320347), Revision 1 submitted October 24, 2006 (ML07240526), and Revision 2 submitted on May 11, 2007 (ML071590195).  The summary of a meeting held on August 30, 2007, is available at ML072540764.  During this meeting, NRC staff and industry representatives also discussed use of the ASME Code and ASME Code Cases for evaluation of structural integrity and for structural integrity acceptance criteria for pressure boundary components with through-wall leakage.

 

8.    10th Meeting of the International Cooperative Program for the Inspection of Nickel Alloy Components (PINC)

 

On October 3-5, 2007, staff from the Office of Nuclear Regulatory Research (RES) conducted this meeting that was and hosted by the Valtion Teknillinen Tutkimuskeskus (VTT) Technical Research Centre of Finland.  The meeting focused on the reliability of non-destructive examinations (NDE) for primary water stress corrosion cracking (PWSCC).  Participants reviewed the progress of the Atlas Database of PWSCC Morphology and NDE Responses and discussed round robin tests that are assessing the performance of various NDE techniques to detect and size PWSCC cracks in piping and dissimilar metal welds.  Additional participants are sought for trials on mockups simulating the seal weld of a bottom mounted instrumentation (BMI) vessel penetration.  Staff visited the Finnish nuclear regulator (STUK) to discuss this program and the codes and standards governing NDE in plant construction and in-service inspection.

 

9.    Public Meeting with Toshiba on Requested Pre-application Review of the “Super-Safe, Small, and Simple” (4S) Reactor Design

 

On October 23, 2007, the staff held a public meeting with representatives from the Toshiba Corporation (Toshiba) and its project partners, Westinghouse Electric Company and Japan’s Central Research Institute of Electric Power Industry, to discuss Toshiba’s request for a pre-application review of the 4S reactor design.  Pre-applicant presentations provided an overview of the 4S design, safety features, and technology base and were followed by discussions of potential focus topics and plans for the requested pre-application design review activities.  A statement provided by representatives from the city of Galena, Alaska, indicates that a formal request for a pre-application review of their siting-related white papers will be submitted in the near future.


Home Page

 

Copyright Disclosure