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NRC Section XI Report – May 2009

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


NRC Report
May 2009

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

 

A final rule was published in the Federal Register [73 FR 52730] on September 8, 2008, incorporating Section III and Section XI of the 2004 Edition of the American Society of Mechanicals Engineers (ASME) Boiler and Pressure Vessel Code into Title 10, Part 50.55a, of the Code of Federal Regulations (10 CFR 50.55a).  The effective date of the rule was October 10, 2008.

 

An amendment to the above rule was published in the Federal Register [73 FR 57235] on October 2, 2008, to correct several paragraph references.

 

A direct final rule has been developed to revise the augmented examination requirements relative to Code Case N-729-1.  During the public comment phase of the proposed rulemaking to incorporate by reference the 2004 Edition into 10 CFR 50.55a, a commenter recommended a change in the percentage of axially oriented flaws for the Code Case N-729-1 specimen set.  Following the publication of the final rule, the commenter requested withdrawal of his recommendation.  The direct final rule will adopt the recommendation for withdrawal by modifying paragraph 10 CFR 50.55a(g)(6)(ii)(D)(4)(ii) to read “At least 20 percent and no more than 60 percent of the flaws shall be oriented axially.”

 

The NRC staff has begun its review of the 2005 Addenda through the 2008 Addenda.  These edition/addenda will be included in the next proposed rulemaking which is scheduled to be published for public comment at the end of 2009.  The NRC staff plans to incorporate the comprehensive inspection requirements of ASME Code Case N-770 by reference into this proposed rule.  The NRC staff also plans to explicitly address endorsement of non-mandatory appendices in this proposed rule.

 

On Wednesday, April 22, 2009, a closure of petition for rulemaking was published in the Federal Register [74 FR 18303].  Specifically, the NRC determined that the issues raised by Raymond A. West in a petition for rulemaking dated December 14, 2007, and revised on December 19, 2007, should be considered in the NRC’s Common Prioritization of Rulemaking process.

 

2.    ASME Code Case Rulemaking/Regulatory Guides

 

Draft Revision 35 to RG 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III,” draft Revision 16 to RG 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,”and draft Revision 3 to RG 1.193 “ASME Code Cases Not Approved for Use” are in final concurrence.  The guides address Code Cases from Supplement 2 to the 2004 Edition through Supplement 0 to the 2007 Edition (Supplement 0, 2007 Edition also serves as Supplement 12 to the 2004 Edition).  The proposed rulemaking and draft regulatory guides will be published in the Federal Register for public comment in May 2009.

 

The NRC staff has completed its review of Supplements 1 - 7, to the 2007 Edition.  Draft revisions 36 to RG 1.84, 17 to RG 1.147, 2 to RG 1.192, and 4 to RG 1.193 are nearly complete.  The goal is to publish these guides for public comment shortly after Revision 35 to RG 1.84, Revision 16 to RG 1.47, and Revision 3 to RG 1.193 have been published as final guides.

 

3.    Risk-Informed Activities

 

Phased Approach to Probabilistic Risk Assessment Quality Regulatory Guide 1.200

 

The increased use of probabilistic risk assessments (PRAs) in the NRC’s regulatory decisionmaking process requires consistency in the quality, scope, methodology, and data used in such analyses.  A key aspect of implementing a phased approach to PRA quality is the development of PRA standards and related guidance documents.  To achieve that objective, professional societies, the nuclear industry, and the staff have undertaken initiatives to develop national consensus standards and guidance on the use of PRA in regulatory decisionmaking.

 

Certain ASME and ANS standards applying to at-power internal events, internal fire events, and external events were combined into a single joint standard, “Standards for Level 1/Large Early Release Frequency (LERF) Probabilistic Risk Assessment Standard for Nuclear Power Plant Applications” (ASME/ANS RA-S-2008).  Accordingly, the staff had initiated work on a revision of RG 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,” that will endorse the joint standard.

 

Regulatory Guide 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,” Revision 2, was issued in March 2009 (ML090410014).  In 1995, the U.S. Nuclear Regulatory Commission (NRC) issued a Policy Statement (Ref. 1) on the use of probabilistic risk analysis (PRA), encouraging its use in all regulatory matters.  This regulatory guide describes one acceptable approach for determining whether the technical adequacy of the 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.  In particular, the regulatory guide defines the quality of a PRA analysis used to support a particular application in terms of its appropriateness with respect to scope, level of detail, and technical acceptability.  This guidance is intended to be consistent with the NRC’s PRA Policy Statement.  It is also intended to reflect and endorse guidance provided by standards-setting and nuclear industry organizations.

 

On February 18, 2009, EPRI submitted a report to the NRC entitled; "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-informed Inservice Inspection Programs, "EPRI, Palo Alto, CA: 2008. 1018427.  The report was submitted as a means of exchanging information with the NRC for the purposes of supporting generic regulatory improvements with respect to application of risk-informed technology to inservice inspection (RI-ISI) programs.  EPRI requested that the NRC issue a safety evaluation report (SER) on the EPRI report.

 

EPRI report 1018427 has been developed to provide guidance in defining which technical elements and supporting requirements of the, plant PRA are applicable to RI-ISI programs.  Also, for those supporting requirements that are applicable to RI-ISI, programs, this report provides guidance on the appropriate capability category.  This guidance was issued for application to both EPRI's traditional RI-ISI methodology (EPRI TR-1 12657) and the streamlined RI-ISI methodology (Code Case N-716).

 

NUREG-1855

 

NUREG-1855, Volume 1, entitled “Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making,“ was published in March 2009 (ML090970525).  The NUREG provides guidance on how to treat uncertainties associated with probabilistic risk assessment (PRA) in risk-informed decisionmaking.  The objectives of this guidance include fostering an understanding of the uncertainties associated with PRA and their impact on the results of PRA and providing a pragmatic approach to addressing these uncertainties in the context of the decisionmaking.  In implementing risk-informed decisionmaking, the NRC expects that appropriate consideration of uncertainty will be given in the analyses used to support the decision and in the interpretation of the findings of those analyses.

 

The NUREG report and Regulatory Guide 1.200 (including EPRI report 1018427) will assist the staff in establishing the technical acceptability of the PRA results to be used in regulatory decision making.  When used in support of an application, these documents will obviate the need for an in-depth review of the base PRA by NRC reviewers, and provide for a more focused and consistent review process.

 

10 CFR 50.46a Rule to Redefine the Large-Break LOCA

 

The NRC intends to publish a revised proposed rule that would establish alternative, risk-informed regulations for emergency core cooling system performance during large-break loss-of-coolant accidents (LOCAs).  The rule would also establish procedures and acceptance criteria for evaluating changes in plant design and operation based upon the results of the new analyses of ECCS performance.  Licensees who perform LOCA analyses using the risk-informed alternative requirements could find that their plant designs are no longer limited by certain parameters associated with previous large break LOCA analyses.  The new requirements could enable some licensees to propose a wide scope of design or operational changes up to the point of being limited by some other parameter associated with any of the required accident analyses.

 

An initial proposed rule was published on November 7, 2005, (70 FR 67598).  However, the nature of the changes made to the rule to incorporate modified Commission guidance made it necessary to allow additional stakeholder comments before issuing a final rule.  The changes made in the revised proposed rule are generally intended to increase the defense-in-depth provided for pipe breaks larger than the transition break size and to modify the risk-informed assessment process required for plant changes made under the rule to more closely follow the existing risk-informed guidance in RG 1.174.

 

The revised proposed rule was made public on April 16, 2009 (www.regulations.gov).  The NRC is planning to issue the revised proposed rule in late-June or early-July 2009 for public comments.  The final rule is scheduled to be issued in mid-2010.

 

 

 

 

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

 

In February 2008 NRC issued Temporary Instruction (TI-172) for regional staff to verify that all pressurized water reactors (PWRs) with dissimilar metal (DM) butt welds are implementing Materials Reliability Program (MRP)-139, “Primary System Piping Butt Weld Inspection and Evaluation Guidelines.”

 

In 2006 ASME started the development of a Code Case for inspection of Alloy 82/182 butt welds.  The Code Case was recently completed.  The staff expects that the Code Case will be acceptable with comments/conditions and will be incorporated by reference directly into the next update to 10 CFR 50.55a.

 

The NRC staff continues to monitor and evaluate operating experience to ensure that the current inspection schedules are adequate.

 

On October 22, 2008, the NRC issued NRC Regulatory Issue Summary 2008-25, “Regulatory Approach for Primary Water Stress Corrosion Cracking of Dissimilar Metal Butt Welds in Pressurized Water Reactor Primary Coolant System Piping.”  The intent in issuing this regulatory issue summary (RIS) is to inform addressees of the regulatory approach for ensuring the integrity of primary coolant system DM butt welds containing Alloy 82/182 in PWR power plants.

 

In late October 2008, a small inside surface connected circumferential indication was identified in a hot leg nozzle to safe-end Alloy 82/182 weld.  As far as the NRC staff is aware, this is the first circumferential indication that has been identified in the U.S. in an Alloy 82/182 reactor vessel to hot leg piping weld.  The staff completed a preliminary evaluation of this flaw and, based on this evaluation, does not believe that this inspection finding has an effect of the current inspection schedules in MRP-139 or Code Case N-770, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 UNS W86182 Weld Filler Material With or Without the Application of Listed Activities.”

 

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.

 

New Reactor Licensing Status

 

As of May 7, 2009, the status of new reactors licensing under10 CFR Part 52 is as follows:

 

Design Certification

 

NRC has issued four design certifications to date (ABWR, System 80+, AP600, and AP1000).  These are certified in 10 CFR Part 52, Appendices A, B, C, and D, respectively.  The NRC is currently reviewing four design certifications:

 

·                    General Electric-Hitachi’s ESBWR (first passive BWR)

·                    AREVA’s EPR (evolutionary pressurized-water reactor)

·                    Mitsubishi Heavy Industries’ US-APWR (advanced pressurized water reactor)

·                    AP1000 Revision 17 (first amended design certification)

 

Early Site Permits (ESP)

 

NRC has issued three ESPs to date (Clinton, Grand Gulf, and North Anna).  The NRC is currently reviewing one ESP (Vogtle).  To date, there have been no ESPs submitted for greenfield sites.

 

Combined License (COL) Applications

 

NRC is currently reviewing 17 COL applications (26 new reactor units):

·                    1 ABWR         South Texas Project 3 and 4

·                    6 AP1000      Bellefonte 3 and 4, William S. Lee Station 1 and 2, Shearon Harris 2 and 3, Vogtle 3 and 4, V.C. Summer 2 and 3, and Levy County 1 and 2

·                    5 ESBWR      North Anna 3 and Grand Gulf 3*, River Bend 3*, Victoria County 1 and 2*, Fermi 3

·                    4 EPR             Calvert Cliffs 3 , Callaway 2, Nine Mile Point 3, Bell Bend

·                    1 US-APWR Comanche Peak Units 3 and 4

 

*     The reviews of the FSARs for these COL applications are on hold pending possible selection of another standard design.

 

NRO Quality and Vendor Branch 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.

 

NRO Vendor Inspection

 

The NRO vendor inspection program is described in Inspection Manual chapter (IMC) 2507, “Construction Inspection Program, Vendor Inspection.”  This IMC will be implemented by various Inspection procedures (IPs) including:

 

IP 43002:  Routine Inspections of Nuclear Vendors;

IP 43003:  Reactive Inspections of Nuclear Vendors;

IP 43004:  Inspection of Commercial-Grade Dedication Programs;

IP 43005:  NRC Oversight of Third Party Organizations Implementing Quality Assurance Requirements; and

IP 36100:  Inspection of 10 CFR Parts 21 and 50.55(e) Programs for Reporting Defects and Noncompliance.

 

FY 09 Vendor Inspection Plans

 

  • Commercial Grade Dedication organizations
  • Forgings suppliers for AP1000, EPR
  • Manufacturing for SG tubes AP1000 and EPR
  • Manufacturing for ESBWR reactor pressure vessel
  • Manufacturing for valves (all new reactor Design Centers)
  • STP alternate Vendor & ABWR component fabrication in Japan
  • Modular Construction Facilities

 

Vendor Inspection Reports completed, issued and planned inspections

 

  • General Electric, Wilmington, NC, ESBWR Design certification inspection, December 2008 – issued
  • Weir Valves & Controls, Ipswich MA, December 2008, -- issued
  • STP Units 3&4 COL QA Implementation Inspection, Bay City TX, January 2009 – issued
  • Dresser Valves, Alexandria, LA, March 9 - 13, 2009 – issued  April 27, 2009
  • Vogtle COL QA Implementation Inspection, Birmingham AL, March 2009 – completed
  • Wylie Labs, Huntsville, AL, March 13 -17, 2009 -- completed
  • Doosan Heavy Industries, Changwon, Korea, March 30 - April 3, 2009 – Issued Members of the Korean Institute of Nuclear Safety (KINS) and Canadian Nuclear Safety Commission (CNSC) observed the NRC inspection at DHI as part of the multinational design evaluation program’s Vendor Inspection Cooperation Working Group (VICWG)
  • Conval Inc, Somers CT,  Valves – completed April 21-22, 2009

 

Vendor Inspections continue to identify findings related to commercial grade dedication activities and inadequate Part 21 programs for evaluating and reporting of defects that could cause a substantial safety hazard.

 

Previously issued NRC inspection and trip reports can be located at

 

http://www.nrc.gov/reactors/new-licensing/quality-assurance.html

 

6.    MRP-227

 

The Electric Power Research Institute (EPRI) Materials Reliability Project (MRP) submitted MRP-227, “Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), to the NRC to exchange information for the purpose of supporting generic regulatory improvements.  The guidelines in MRP-227 were developed to address the aging of reactor internals components and as a means to satisfy the guidance in Generic Aging Lessons Learned (GALL) Report (NUREG-1801).

 

7.    ASME Letter to NRC – Limitations Imposed by NRC on Edition/Addenda and Code Cases

 

A letter dated March 23, 2009, from Bryan Erler, Vice President, Nuclear Codes and Standards, ASME, to Brian Sheron, Director, Office of Nuclear Regulatory Research, and Eric Leeds, Director, Office of Nuclear Reactor Regulation, discussed the cooperative effort between the NRC staff and Section XI committee members to re-evaluate the conditions imposed in 10 CFR 50.55a by the NRC on certain edition, addenda, and Code Cases.  This is an on-going effort initiated in 2006 to try to address what was, at that time, a growing number of conditions.  As a result of this re-evaluation, a significant number of the conditions have been addressed either through changes to the ASME Code or to 10 CFR 50.55a.

 

The March 23, 2009, letter discusses the results of Phase II of this effort and further changes to the ASME Code to address some of the conditions not considered in the first phase.  Also, the letter provides additional technical justification to request that the NRC consider eliminating a few of the conditions in 10 CFR 50.55a.

 

8.    Meeting on Management of Fatigue

 

A meeting was held on Wednesday, April 29, 2009 between NRC staff and the EPRI/MRP/ Technical Support Committee on thermal stratification fatigue.  The purpose of this meeting was to inform the NRC of progress in the development of MRP guidelines for managing thermal fatigue in reactor coolant loop (RCL) small bore piping.  NRC staff approval or endorsement was not discussed or requested.

 

The EPRI participants discussed technical issues related to the development of industry guidelines for managing thermal fatigue in normally stagnant RCL unisolable small bore subjected to thermal stratification cycling.  These guidelines are described in EPRI report MRP‑146/MRP146S.  The EPRI participants also described the implementation of these guidelines through bi-annual industry workshops and on-line and computer-based NDE training.

 

9.    NRC Regulatory Issue Summary 2009-03

 

On February 12, 2009, the NRC issued Regulatory Issue Summary (RIS) 2009-03, “Process for Scheduling Acceptance Reviews of New Reactor Licensing Applications After April 2009 and Process for Determining Budget Needs for Fiscal Year 2011.”  The NRC issued the RIS to (1) gather information needed for determining fiscal year (FY) 2011 resource and budget needs with respect to future construction-related activities; (2) communicate to stakeholders the agency’s process for scheduling its acceptance reviews, (3) communicate to stakeholders that the NRC has expanded its scheduling process to include all potential 10 CFR Part 52 licensing actions; and (4) request addressees to consider submitting their construction plans and schedules for fabrication of large components and modules to the NRC when available.

 

The NRC has received 17 COL applications since 2007, and anticipates submission of approximately six additional COL applications through 2010.  The NRC is reviewing three design certification (DC) applications (Economic Simplified Boiling-Water Reactor, U.S. Evolutionary Power Reactor, and U.S. Advanced Pressurized-Water Reactor), one amended DC application (AP1000 DC Amendment), and one ESP application (Vogtle ESP).  The information requested by this RIS will assist the NRC in effectively scheduling resources to review any potential new applications or modifications to current applications.

 

10.    NRC Regulatory Issue Summary 2009-04

 

On April 3, 2009, the NRC issued Regulatory Issue Summary 2009-04, “Steam Generator Tube Inspection Requirements.”  The RIS was issued to clarify the NRC’s regulatory position on the implementation of the steam generator (SG) tube inspection requirements contained within the technical specifications (TS) of a plant.  The NRC expects addressees to review this RIS for applicability to their facilities and to consider actions as appropriate.

 

After public discussions with the NRC staff, the Nuclear Energy Institute Technical Specification Task Force (TSTF) submitted a proposal containing performance-based programmatic requirements for addressing steam generator tube integrity in the Standard Technical Specifications (STS).  They designated the proposed changes as TSTF-449, “Steam Generator Tube Integrity, Revision 0.”  Based on public interactions with the staff, Revision 0 was modified.  On May 6, 2005, in accordance with the NRC consolidated line item improvement process, the NRC issued a notice of availability (70 FR 24126) of the model application, and referenced the previously (March 2, 2005) noticed no significant hazards consideration determination and model safety evaluation for the adoption of TSTF-449, Revision 4.

 

The revised, performance-based, generic TS focus on ensuring that SG tubes satisfy performance criteria commensurate with the assurance of adequate tube integrity.  All currently operating pressurized-water reactor plants have adopted SG TS similar to those in TSTF-449, Revision 4.  The NRC and the industry have interacted several times concerning proper implementation of the new SG tube inspection programmatic requirements, including public meetings on May 2, 2007 and November 29, 2007.  This RIS clarifies the staff’s position related to issues raised by the industry in implementing these inspection requirements.

 

11.    Round-robin Visual Testing (VT) at PNNL

 

As part of an NRC research program validating the effectiveness and reliability of nondestructive examination methods, a round-robin test is under consideration using commercial VT agencies and qualified examiners to assess the applicability of using VT-1 examinations in lieu of volumetric or surface examinations.  The research would be conducted, in part, to quantify: the essential variables for equipment and procedures; the effects of examination conditions on the acuity of personnel; and quantify examination tolerances for the detectability of the different types of cracks.  Pacific Northwest National Laboratory (PNNL) is currently fabricating mockups with cracks representative of those being reported in the field.  Interested parties should contact Wallace Norris at the USNRC (Tel. 301-251-7650; wallace.norris@nrc.gov or Stephen Cumblidge at PNNL (Tel:  509-372-4054; stephen.cumblidge@pnl.gov).

 

12.    BWR EPU License Amendments

 

The NRC is currently reviewing two BWR extended power uprate (EPU) license amendment requests from Browns Ferry Unit 1 and Monticello, and is expecting an EPU amendment request from Nine Point in the near future.  The NRC is expecting two other EPU license amendment requests in 2010.

 

13.    GE Topical Reports – NRC Staff Reviews

 

The NRC received two topical reports for review:

 

NEDC-33436P: GEH BWR Steam Dryer-Plant Based Load Evaluation (PBLE) from GEH and BWRVIP-194: Steam Dryer Integrity for Power Uprates from BWRVIP.

 

14.    BWRVIP Reports

 

Two topical reports from the Boiling Water Reactor Vessel and Internals Program (BWRVIP) are currently under staff review:

 

BWRVIP-181: Steam dryer Repair Design Criteria, and

BWRVIP-182: Guidance for demonstration of Steam dryer Integrity.


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