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NRC Section XI Report - May 2004

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


NRC Report
Section XI
May 2004

1. Amendments to 10 CFR 50.55a

 

The final amendment to 10 CFR 50.55a incorporating the 1997 Addenda through 2000 Addenda by reference was published on September 26, 2002 (67 FR 60520) and is available at the Office of Federal Register website: http://www.access.gpo.gov/su_docs/fedreg/a010803c.html.

 

A proposed rule to amend 10 CFR 50.55a to incorporate by reference the 2001 Edition through 2003 Addenda of the ASME Code was published in the Federal Register on January 7, 2004 (69 FR 879).  The public comment period ended on March 22, 2004.  Fifteen comment letters were received, and the NRC staff is in the process of responding to comments.  The final rule is scheduled to be published in September 2004.

 

2. ASME Code Cases - Rulemaking/Regulatory Guides

 

Four regulatory guides were published final in June 2003: Revision 32 to Regulatory Guide 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III; Regulatory Guide 1.192, “Operation and Maintenance Code Case Acceptability,” ASME OM Code; Revision 13 to Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1; and Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use.”  The guides address Supplement 4 to the 1992 Edition through Supplement 11 to the 1998 Edition, and OMN-1 through OMN-13.  The ADAMS numbers are: RG 1.184: ML030730417; RG 1.147: ML030730423; RG 1.192 (OM Guide):ML030730430; RG 1.193 (Unapproved code cases): ML030730440; and the response to public comments: ML030730448.  The final rule accompanying the guides was published on July 8, 2003 (68 FR 40469) and became effective on August 7, 2003.

 

The NRC staff has completed its review of the Code Cases in Supplement 12 to the 1998 Edition through Supplement 6 to the 2001 Edition.  Copies of Draft Revision 33 to RG 1.84, Draft Revision 14 to RG 1.147, and Draft Revision 1 to RG 1.193 have been printed.  The Advisory Committee on Reactor Safeguards (ACRS) and the Committee to Review Generic Requirements (CRGR) have approved the proposed rule for publication.  The proposed rule and draft guides are scheduled to be published for public comment in June 2004.

 

The staff has completed its review of the Code Cases in Supplements 7 through 10 to the 2001 Edition and has begun reviewing Supplements 11 and 12.  The Section XI Code Cases contained in these supplements will be included in Draft Revision 15 to RG 1.147, which has been initiated.

 

3. Risk-Informed Activities

 

Plan for Phased Approach to PRA Quality Drafted

 

On December 18, 2003, the Commission issued a Staff Requirements Memorandum (SRM) (COMNJD-03-2003) that directs the staff to develop an action plan for implementing a phased approach to PRA quality.  An attachment to the memorandum outlines the Commission's expectations for each of the four phases.  In response to the SRM, a plan entitled, Stabilizing the PRA Quality Expectations and Requirements has been drafted.  The staff is currently seeking review and comment from stakeholders inside and outside the NRC.

 

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 draft 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.

 

On March 24, 2004, the project team briefed industry representatives on the draft plan and received initial feedback.  A similar briefing was conducted on March 25, 2004, with the ACRS Subcommittee on Reliability and PRA and on April 15, 2004, with the ACRS Full Committee.

 

Over the next few months, the project team will seek comment from additional internal and external stakeholders.  The team expects to finalize the plan and submit it to the Commission in July 2004.

 

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.

 

IMPORTANT NOTE:  The appropriate level of quality of the PRA's used to support the applications envisioned by ASME Code Cases should be consistent with RG 1.200 as necessary to support the phased approach to PRA quality outlined in COMNJD-03-0002.  PRA quality considerations unique to ASME Code Cases arise from the ability of licensees to implement staff approved Code Cases without prior staff review and approval of the proposed change.  The ASME is encouraged to provide comments to the staff on the draft plan.

 

Risk Management Coordinating Committee (RMCC)

 

The formation of this committee is jointly proposed by the American Society of Mechanical Engineers, American Nuclear Society, and the U.S. Nuclear Regulatory Commission.  The committee will 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.  Since the February 20, 2004, Organization Meeting, the Committee has assembled a draft charter, begun the process of nominating members of the committee, and has started to prioritize potential risk informed standards development projects.  The Committee’s next meeting is scheduled for May 20, 2004, and will be held by telephone conference.

 

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

 

Representatives from the Electric Power Research Institute’s Materials Reliability Program (MRP) met with the NRC staff on April 14, 2004, to discuss their vessel head inspection program.  Specifics of the inspections have yet to be worked out, but the MRP representatives stated that the technical analyses showed that there was a low likelihood of leakage from vessel head penetration nozzles that would be undetected by a bare metal visual inspection.  Thus, MRP representatives believe that ultrasonic testing (UT) inspections do not need to be conducted at every refueling outage, even for plants in the high susceptibility range.  The MRP is also working on the development of a technical basis to differentiate between Alloy 600 and Alloy 690 materials.  The NRC staff stated that it would need quantitative, technically supported data for the agency to permit longer intervals between inspections for reactor heads using alloy 690.

 

The NRC staff has begun the process of converting the vessel head inspection order into a rule.  The rulemaking plan is scheduled to be provided to the Commission in June 2004.  The MRP inspection plan is scheduled to be completed in late summer or early fall which would coincide with the projected comment period for the proposed rule.

 

On April 22, 2004, NRC Information Notice 2004‑08, “Reactor Coolant Pressure Boundary Leakage Attributable to Propagation of Cracking in Reactor Vessel Nozzle Welds,” was issued to all operating boiling‑water reactors.  The purpose of the information notice was to alert licensees to cracking identified in the nozzle‑to‑cap weld of control rod drive (CRD) return line penetration N10 at Pilgrim Nuclear Power Station.  The leakage was from the nozzle‑to‑cap weld area of penetration N10.  The licensee used a Performance Demonstration Initiative qualified manual ultrasonic testing technique to determine that the N10 nozzle‑to‑cap weld contained an unacceptable flaw that was approximately 4.45cm (1.75 inches) long in the circumferential direction.  Observations by the nondestructive examination inspector suggested that the flaw initiated at the inner diameter of the weld, in the area of previous weld repairs.  The through‑wall location appeared to be close to the centerline of the weld.

 

The reactor pressure vessel nozzle is made of SA‑508, Class 2 low‑alloy steel, while the CRD return line cap is made of Alloy 600.  The weld is fabricated with Alloy 82/182 material, and the nozzle side of the weld is buttered with Alloy 182 material.  The licensee concluded that the root cause of the cracking in the nozzle‑to‑cap weld of the CRD return line in penetration N10 was interdendritic SCC (IDSCC), given that the flaw was completely contained within the weld.  The licensee believes that the IDSCC was induced by a combination of a crevice condition and weld repair stresses resulting from previous local weld repairs.  The licensee performed a weld overlay repair to stop the leakage, using Code Case N‑504‑2, "Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping" (with modifications), and Code Case N‑638, “Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique," as an alternative to the requirements in IWA‑4000.

 

The Advisory Committee on Reactor Safeguards (ACRS) Materials and Plant Operations Subcommittees are scheduled to hear presentations on reactor pressure vessel head degradation from industry and the Staff on June 1, 2004, followed by an abbreviated presentation at the June 2nd full committee meeting.  The purpose is to present to the ACRS the current state of integrity of the plants and the progress of life management plans by both industry and staff.

 

5. Examination Requirements for Cast Austenitic Piping Welds Greater Than 2 inches Thickness

 

Efforts to develop qualification requirements for Appendix VIII, Supplement 9, “Cast Stainless Steel,” have been hindered by the current difficulties in volumetrically examining cast austenitic stainless steel (CASS) piping from the outside diameter (OD).  Under sponsorship of the NRC, Pacific Northwest National Laboratory (PNNL) has been conducting a series of studies on twenty  PNNL and Westinghouse Owners Group (WOG) specimens using eddy-current examination (ET) from the inside diameter (ID) of the specimens.  This has been shown to be quite successful as all of the cracks were detected and length sized to standards contained in Appendix VIII.  Thus, preliminary tests indicate that this surface examination method may prove to be very effective at detecting surface breaking cracks.

 

PNNL has begun conducting Low Frequency (LF) Synthetic Aperture Focusing Technique (SAFT) and phased array examination of the WOG specimens from the OD.  ET scans on CASS surge line piping has been conducted to assess the background noise level of this pipe material.  It is not as “noisy” as the worst CASS in the WOG specimens but is noisier than the quiet WOG CASS material and the PNNL material.  A ring specimen cut off a pipe (on loan from the EPRI NDE Center) from a canceled plant is being polished to etch the grain structure and photograph to document the microstructure.  This pipe specimen will then be examined using both LF-SAFT and phased array.  ET scans of the ID of the pipe specimen will be conducted to determine the noise properties and to perform UT scans to assess the coherent noise properties for UT.  Based on these measurements, a determination will be made as to the best path forward.  Meaningful results from these efforts should be available by the September 2004 meetings in New Orleans.

 

6. Workshop on International Developments and Cooperation on Risk-Informed Inservice Inspection and Nondestructive Testing Qualification

 

NRC staff participated and presented papers at the Committee on the Safety of Nuclear Installations (CSNI) "Workshop on International Developments and Cooperation on Risk-Informed Inservice Inspection (RI-ISI) and Nondestructive Testing (NDT) Qualification" at the Swedish Nuclear Inspectorate (SKI) Offices on April 13 and 14, 2004, in Stockholm, Sweden.  Objectives of the Workshop were to discuss experiences, research, and initiatives related to RI‑ISI and NDT.  Attendees included regulatory, industry, and research organization representatives.  U.S. participants also included representatives of EPRI and  vendors.  A major initiative is related to development of structural reliability methodology.  This is of significant interest to NRC, and complements activities related to development of probabilistic fracture mechanics (PFM) codes to evaluate effects of degradation on structural integrity.

 

NRC staff also attended annual meetings of the Metals Subgroup and the Main Working Group on Integrity of Components and Structures (WGIAGE) on April 15 and 16, respectively.  Items of special interest included: ongoing benchmark project on methods to evaluate pressurized thermal shock (PTS); ongoing survey of member countries on experiences and research activities related to Primary Water Stress Corrosion Cracking and Alloy 600 and 690; and a possible new initiative on benchmarking of PFM codes.  Aging management practices are also discussed and plans are to issue the presentations in a comprehensive report.


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