Performance-Based Emergency Core Cooling Systems Cladding Acceptance Criteria

NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50 and 52

[NRC-2008-0332, NRC-2012-0041, NRC-2012-0042, NRC-2012-0043]

RIN 3150-AH42

Performance-Based Emergency Core Cooling Systems Cladding Acceptance Criteria

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Proposed rule.

SUMMARY:

The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to revise the acceptance criteria for the emergency core cooling system (ECCS) for light-water nuclear power reactors. The proposed ECCS acceptance criteria are performance-based, and reflect recent research findings that identified new embrittlement mechanisms for fuel rods with zirconium alloy cladding under loss-of-coolant accident (LOCA) conditions. The proposed rule also addresses two petitions for rulemaking (PRMs) by establishing requirements applicable to all fuel types and cladding materials, and requiring the consideration of crud, oxide deposits, and hydrogen content in zirconium-based alloy fuel cladding. Further, the proposed rule contains a provision that would allow licensees to use an alternative risk-informed approach to evaluate the effects of debris for long-term cooling. The NRC is also seeking public comment on three draft regulatory guides that would support the implementation of the proposed rule.

DATES:

Submit comments on the rule and draft guidance by June 9, 2014. To facilitate NRC review, please distinguish between comments submitted on the proposed rule and comments submitted on the draft guidance. Submit comments on the information collection aspects of this rule by April 23, 2014. Comments received after these dates will be considered if it is practical to do so, but assurance of consideration cannot be given to comments received after these dates.

ADDRESSES:

The methods for accessing information and comment submissions, and submitting comments on the proposed rule are different from the methods for accessing information and comment submissions, and submitting comments on the draft regulatory guides.

Proposed Rule

You may access information and comment submissions related to this proposed rule by searching on

http://www.regulations.gov

under Docket ID NRC-2008-0332. You may submit comments on the proposed rule by any of the following methods:

• Federal rulemaking Web site: Go to

http://www.regulations.gov

and search for Docket ID NRC-2008-0332. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-3422; email:

Carol.Gallagher@nrc.gov.

For technical questions, please contact the individuals listed in the

FOR FURTHER INFORMATION CONTACT

section of this document.

• Email comments to:

Rulemaking.Comments@nrc.gov.

If you do not receive an automatic email reply confirming receipt, then contact us at 301-415-1677.

• Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 301-415-1101.

• Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.

• Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland, 20852, between 7:30 a.m. and 4:15 p.m. (Eastern Time) Federal workdays; telephone: 301-415-1677.

Draft Regulatory Guides

You may access information and comment submissions related to the draft regulatory guides (DGs) by searching on

http://www.regulations.gov

under Docket ID NRC-2012-0041 (DG-1261, “Conducting Periodic Testing for Breakaway Oxidation Behavior” (the NRC's Agencywide Documents Access and Management System (ADAMS) Accession No. ML12284A324)), Docket ID NRC-2012-0042 (DG-1262, “Testing for Post Quench Ductility” (ADAMS Accession No. ML12284A325)), and Docket ID NRC-2012-0043 (DG-1263, “Establishing Analytical Limits for Zirconium-Based Alloy Cladding” (ADAMS Accession No. ML12284A323)), respectively. You may submit comments on the draft regulatory guides by any of the following methods:

• Federal rulemaking Web site: Go to

http://www.regulations.gov

and search for Docket IDs NRC-2012-0041, NRC-2012-0042, and NRC-2012-0043, respectively. Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch, Office of Administration, Mail Stop: 3WFN-06-44M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

Information Collections

You may submit comments on the information collections by the methods described in the

SUPPLEMENTARY INFORMATION

section of this document, under the heading, “Paperwork Reduction Act Statement.”

For additional direction on accessing information and submitting comments, see “Accessing Information and Submitting Comments” in the

SUPPLEMENTARY INFORMATION

section of this document.

FOR FURTHER INFORMATION CONTACT:

Tara Inverso, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-1024, email:

Tara.Inverso@nrc.gov;

or Paul M. Clifford, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-4043, email:

Paul.Clifford@nrc.gov.

SUPPLEMENTARY INFORMATION:

Table of Contents

Executive Summary.

I. Accessing Information and Submitting Comments.

A. Accessing Information.

B. Submitting Comments.

II. Background.

A. Emergency Core Cooling System: Embrittlement Research Findings.

B. Generic Safety Issue (GSI)-191 and Long-Term Cooling.

III. Operating Plant Safety.

A. Emergency Core Cooling System: Embrittlement Research Findings.

B. GSI-191 and Long-Term Cooling.

IV. Advance Notice of Proposed Rulemaking: Public Comments.

V. Proposed Requirements for ECCS Performance During LOCAs.

A. Applicability of Performance-Based Rule: Consideration of PRM-50-71.

B. Performance-Based Aspects of the Proposed Rule.

1. Hydrogen-Enhanced Beta-Layer Embrittlement.

2. Oxygen Ingress From Cladding Inside Diameter.

3. Breakaway Oxidation.

4. Applicability of Ductility-Based Analytical Limits in the Burst Region.

5. Long-Term Cooling.

6. Use of Risk-Informed Approaches To Address Debris for Long-Term Cooling.

C. Corrective Actions and Reporting Requirements.

1. Peak Cladding Temperature and Equivalent Cladding Reacted.

2. Risk-Informed Alternative To Address Debris for Long-Term Cooling.

D. Consideration of PRM-50-84: Thermal Effects of Crud and Oxide Layers.

E. Implementation.

1. Staggered Implementation Schedule.

2. Compliance With Long-Term Cooling Requirements Using Risk-Informed Approach To Address Debris Effects.

VI. Section-by-Section Analysis.

A. Section 50.46c—Heading.

B. Section 50.46c(a)—Applicability.

C. Section 50.46c(b)—Definitions.

D. Section 50.46c(c)—Relationship to Other NRC Regulations.

E. Section 50.46c(d)—Emergency Core Cooling System Design.

F. Section 50.46c(e)—Alternate Risk-Informed Approach for Addressing the Effects of Debris on Long-Term Core Cooling.

G. Section 50.46c(g)—Fuel System Designs: Uranium Oxide or Mixed Uranium-Plutonium Oxide Pellets Within Cylindrical Zirconium-Alloy Cladding.

H. Section 50.46c(k)—Use of NRC-Approved Fuel in Reactor.

I. Section 50.46c(l)—Authority To Impose Restrictions on Operation.

J. Section 50.46c(m)—Corrective Actions and Reporting.

K. Section 50.46c(o)—Implementation.

L. Appendix K to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR), ECCS Evaluation Models.

M. Redesignation of Venting Requirements in § 50.46a.

N. Changes Throughout 10 CFR Parts 50 and 52.

VII. Specific Request for Comments on the Proposed Rule.

A. Fuel Performance Criteria.

B. Risk-Informed Alternative To Address the Effects of Debris.

C. Implementation.

D. Other Issues.

VIII. Request for Comment: Draft Regulatory Guidance.

IX. Availability of Documents.

X. Criminal Penalties.

XI. Agreement State Compatibility.

XII. Plain Writing.

XIII. Voluntary Consensus Standards.

XIV. Finding of No Significant Environmental Impact: Environmental Assessment.

XV. Paperwork Reduction Act Statement.

XVI. Regulatory Analysis: Availability.

XVII. Regulatory Flexibility Certification.

XVIII. Backfitting and Issue Finality.

Executive Summary

Purpose of the Regulatory Action

The proposed rule would adopt performance-based regulatory requirements for determining the acceptability of an ECCS for a nuclear power reactor, including requirements governing the acceptability of the cladding of fuel. (Cladding performance affects the cooling requirements for the ECCS.) The proposed rule would expand the applicability of the rule from uranium oxide pellets within cylindrical zircaloy or ZIRLO

TM

cladding to any light-water reactor (LWR), regardless of fuel design or cladding material. The proposed rule would also replace prescriptive requirements with performance-based requirements. Performance-based ECCS requirements would provide more flexibility for applicants and licensees to meet NRC requirements for emergency core cooling systems in a manner that provides reasonable assurance of adequate protection consistent with the requirements of the Atomic Energy Act of 1954, as amended. The requirements of the proposed performance-based rule also address new technical information on fuel cladding integrity and degradation mechanisms.

The proposed rule would also address two PRMs, PRM-50-71 and PRM-50-84. The PRM-50-71 requests that the NRC expand the applicability of the ECCS rule beyond zircaloy and ZIRLO

TM

cladding materials. The PRM-50-84 requests, among other items, that the NRC require licensees to consider the thermal effects of crud and oxide layers.

Finally, the proposed rule would allow individual nuclear power plant licensees to resolve GSI-191, “Assessment of Debris Accumulation on PWR [Pressurized Water Reactor] Sump Performance,” by using a risk-informed approach for evaluating the effects of debris on long-term cooling.

Summary of the Significant Changes in the Proposed Rule

The proposed rule includes several significant changes to the NRC's existing requirements on the ECCS:

• The proposed rule would replace prescriptive analytical requirements with performance-based requirements. To demonstrate compliance with the requirements, ECCS performance would be evaluated using fuel-specific performance objectives and associated analytical limits that take into consideration all known degradation mechanisms and unique features of the particular fuel system, along with an NRC-approved ECCS evaluation model.

• The proposed rule would apply to all fuel designs and cladding materials. The proposed rule would define two principle ECCS performance requirements:

Core temperature during and following the LOCA does not exceed the analytical limits for the fuel design used for ensuring acceptable performance.

The ECCS provides sufficient coolant so that decay heat will be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

The proposed rule would also include specific performance requirements for fuel designs consisting of uranium oxide or mixed uranium-plutonium oxide fuel pellets within cylindrical zirconium-alloy cladding. New performance objectives and analytical limits may be necessary for other fuel designs, as they are developed. These changes address the requests of PRM-50-71.

• The proposed rule would incorporate the results of recent research findings. The current requirement to maintain the calculated total cladding oxidation below 17 percent would be replaced with a requirement to establish analytical limits on peak cladding temperature (PCT) and integral time at temperature (ITT) that correspond to the measured ductile-to-brittle transition for the zirconium-alloy cladding material. The proposed rule would also address a newly identified phenomenon known as breakaway oxidation by requiring that the total accumulated time that the cladding is predicted to remain above a temperature at which the zirconium-alloy has been shown to be susceptible to breakaway oxidation shall not be greater than a limit that corresponds to the measured onset of breakaway oxidation for that cladding. The proposed rule would also add a requirement to periodically measure breakaway oxidation. Additionally, the proposed rule would require licensees to consider the effects of oxygen diffusion from the cladding inside surfaces, if an oxygen source is present on the inside surfaces at the onset of the LOCA.

• The proposed rule would require that licensees evaluate the thermal effects of crud and oxide layers that accumulate on the fuel cladding during plant operation. Crud is defined as any foreign substance deposited on the surface of the fuel cladding prior to initiation of a LOCA. This addition addresses a request of PRM-50-84.

• The proposed rule contains a provision that would allow licensees to use an alternative risk-informed approach to evaluate the effects of debris for long-term cooling. The proposed rule contains acceptance criteria that would apply to the risk-informed approach and its required content. Additionally, the proposed rule would add reporting requirements that pertain to the risk-informed approach.

Costs and Benefits

The proposed rule, by requiring applicants and licensees to address new technical matters not currently required to be addressed by the NRC's existing ECCS requirements, would provide adequate protection to the health and safety of the public by maintaining that level of protection that the NRC previously thought would be achieved by the current rule. The NRC prepared a draft regulatory analysis for this proposed rule (ADAMS Accession No.

ML12283A188) to identify the benefits and costs of the particular regulatory approach for addressing ECCS performance. The NRC notes that adequate protection must be assured without regard to cost, but if there is more than one way of achieving that level of protection, then costs may be considered. The draft regulatory analysis prepared for this rulemaking was used to help the NRC identify the most effective way of achieving reasonable assurance of adequate protection with respect to protection against LOCAs.

The benefits of maintaining reasonable assurance of protection with respect to protection against LOCAs were not quantified. The NRC estimates that the total cost of the proposed rule would be $35 million (7 percent net present value). The benefits of the proposed rule are several. The proposed rule would result in savings by obviating the need for exemption requests to use additional claddings and exemption requests stemming from the risk-informed alternative. As a more general matter, adopting a performance-based approach to demonstrating ECCS adequacy may afford applicants and licensees greater flexibility in complying with the NRC's ECCS requirements. This may result in reduced applicant and licensee costs with no adverse effect on public health and safety.

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2008-0332, Docket ID NRC-2012-0041, Docket ID NRC-2012-0042, or Docket ID NRC-2012-0043 when contacting the NRC about the availability of information for this proposed rule or draft regulatory guides, respectively. You may access information related to this proposed rulemaking or draft regulatory guides by the following methods:

• Federal Rulemaking Web site: Go to

http://www.regulations.gov

and search for Docket ID NRC-2008-0332 for the proposed rule, and Docket ID NRC-2012-0041, Docket ID NRC-2012-0042, or Docket ID NRC-2012-0043 for the draft regulatory guides.

• NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly-available documents online in the NRC Library at

http://www.nrc.gov/reading-rm/adams.html.

To begin the search, select “

ADAMS Public Documents

” and then select “

Begin Web-based ADAMS Search.

” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to

PDR.Resource@nrc.gov.

The ADAMS accession number for each document referenced in this notice (if that document is available in ADAMS) is provided the first time that a document is referenced. In addition, for the convenience of the reader, the ADAMS accession numbers are provided in a table in the section of this document entitled,

Availability of Documents.

• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include the appropriate NRC Docket ID in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in that docket.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at

http://www.regulations.gov

as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submissions. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

II. Background

A. Emergency Core Cooling System: Embrittlement Research Findings

In SECY-98-300, “Options for Risk-Informed Revisions to 10 CFR Part 50-`Domestic Licensing of Production and Utilization Facilities,' ” dated December 23, 1998 (ADAMS Accession No. ML992870048), the NRC began to explore approaches to risk-informing its regulations for nuclear power reactors. One alternative (termed “Option 3”) involved making risk-informed changes to the specific requirements in the body of 10 CFR part 50. As the NRC began to develop its approach to risk-informing these requirements, it sought stakeholder input in public meetings. Two of the regulations identified by industry as potentially benefitting from risk-informed changes were §§ 50.44 and 50.46. Section 50.44 specifies the requirements for combustible gas control inside reactor containment structures, and § 50.46 specifies the requirements for light-water power reactor emergency core cooling systems. For § 50.46, the potential was identified for making risk-informed changes to requirements for both ECCS cooling performance and ECCS analysis acceptance criteria in § 50.46(b).

PRM-50-71

On March 14, 2000, as amended on April 12, 2000, the Nuclear Energy Institute (NEI) submitted a PRM (ADAMS Accession No. ML003723791) requesting that the NRC amend its regulations in §§ 50.44 and 50.46 (PRM-50-71). The NEI petition noted that these two regulations apply to only two specific zirconium-alloy fuel cladding materials (zircaloy and ZIRLO

TM

). The NEI stated that reactor fuel vendors had subsequently developed new cladding materials other than zircaloy and ZIRLO

TM

and that, in order for licensees to use these new materials under the regulations, licensees needed to request NRC approval of exemptions from §§ 50.44 and 50.46.

On May 31, 2000, the NRC published a notice of receipt (65 FR 34599) and requested public comment. The public comment period ended on August 14, 2000, and the NRC received 11 public comment letters from public citizens and the nuclear industry. Although the majority of the comments generally supported the requests of the PRM, one commenter suggested that the enhanced efficiency of the proposal would be at the expense of public health and safety. The NRC disagrees with that commenter and notes that, while the petition's proposal would remove specific zirconium-alloy names from the regulation, the NRC review and approval of specific zirconium-alloys for use as reactor fuel cladding would be required prior to their use in reactors (with the exception of lead test assemblies permitted in technical specifications). The NRC's detailed discussion of the public comments submitted on PRM-50-71, including a detailed list of commenters, is contained in a separate document, “Section 50.46c and PRM-50-71 Comment Response Document” (ADAMS Accession No. ML12283A213).

After evaluating the petition and public comments received, the NRC

decided that PRM-50-71 should be considered in the rulemaking process. The NRC's determination was published in the

Federal Register

on November 6, 2008 (73 FR 66000). Because most of the issues raised in this PRM pertain to § 50.46, the PRM is addressed in this proposed rule.

1

1

PRM-50-71 also requested changes to § 50.44. Those changes were addressed in a rulemaking that revised that section (68 FR 54123; September 16, 2003) to include risk-informed requirements for combustible gas control. That regulation was also modified to be applicable to all boiling or pressurized water reactors regardless of type of fuel cladding material used.

Staff Requirements Memorandum Direction

On March 31, 2003, in response to SECY-02-0057, “Update to SECY-01-0133, `Fourth Status Report on Study of Risk-Informed Changes to the Technical Requirements of 10 CFR Part 50 (Option 3) and Recommendations on Risk-Informed Changes to 10 CFR 50.46 (ECCS Acceptance Criteria)' ” (ADAMS Accession No. ML020660607), the Commission issued a staff requirements memorandum (SRM) (ADAMS Accession No. ML030910476) directing the NRC staff to move forward to risk-inform its regulations in a number of specific areas. In addition, this SRM directed the staff to modify the ECCS acceptance criteria to provide a more performance-based approach to the ECCS requirements in § 50.46.

Research Results

Separate from the effort to modify the regulations to provide a more risk-informed, performance-based regulatory approach, the NRC had also undertaken a fuel cladding research program to investigate the behavior of high-exposure fuel cladding under accident conditions. This research program included an extensive LOCA research and testing program at Argonne National Laboratory (ANL), as well as jointly-funded programs at the Kurchatov Institute (supported by the French Institute for Radiological Protection and Nuclear Safety and the NRC) and the Halden Reactor project (a jointly-funded program under the auspices of the Organization for Economic Cooperative Development—Nuclear Energy Agency, sponsored by national organizations in 18 countries), to develop the body of technical information needed to support the new regulations.

The effects of both alloy composition and fuel burnup (the extent to which fuel is used in a reactor) on cladding embrittlement (e.g., loss of ductility) under accident conditions were studied in these research programs. The research programs identified new cladding embrittlement mechanisms and expanded the NRC's knowledge of previously identified mechanisms. The research results revealed that alloy composition has a minor effect on embrittlement, but that the cladding corrosion that occurs as fuel burnup increases has a substantial effect on embrittlement. One of the major findings of the NRC's research program was that hydrogen, which is absorbed in the cladding as a result of zirconium oxidation (e.g., corrosion) under normal operation, has a significant influence on embrittlement during a postulated LOCA. Increased hydrogen content increases both the solubility of oxygen in zirconium and the rate at which it is diffused within the metal, thus increasing the amount of oxygen in the metal during high temperature oxidation in LOCA conditions. Further, the NRC's research program found that oxygen from the oxide fuel pellets enters the cladding from the inner surface if a bonding layer exists between the fuel pellet and the cladding, in addition to the oxygen that enters from the oxide layer on the outside of the cladding. Moreover, under some small-break LOCA conditions (such as extended time-at-temperature around 1,000 degrees Celsius (°C) (1832 degrees Fahrenheit (°F))), a phenomenon termed breakaway oxidation can take place, allowing large amounts of hydrogen to diffuse into the cladding, exacerbating the embrittlement process. Breakaway oxidation is defined as the fuel cladding oxidation phenomenon in which weight gain rate deviates from normal kinetics. This change occurs with a rapid increase of hydrogen pickup during prolonged exposure to a high temperature steam environment, which promotes lack of ductility.

The research results also confirmed a previous finding that if cladding rupture occurs during a LOCA, large amounts of hydrogen from the steam-cladding reaction can enter the cladding inside surface near the rupture location. These research findings have been summarized in Research Information Letter (RIL)-0801, “Technical Basis for Revision of Embrittlement Criteria in 10 CFR 50.46” (ADAMS Accession No. ML081350225), and the detailed experimental results from the program at ANL are contained in NUREG/CR-6967, “Cladding Embrittlement during Postulated Loss-of-Coolant Accidents” (ADAMS Accession No. ML082130389). Since the publication of NUREG/CR-6967 and RIL-0801, additional testing was conducted related to the embrittlement phenomenon, which has been documented in supplemental reports. Where the additional testing relates to conclusions and recommendations in RIL-0801, RIL-0801 has been supplemented to reference the additional reports and incorporate findings (“Update to Research Information on Cladding Embrittlement Criteria in 10 CFR 50.46,” dated December 29, 2011 (ADAMS Accession No. ML113050484)).

The NRC publicly released the technical basis information in RIL-0801 on May 30, 2008, and NUREG/CR-6967 on July 31, 2008. Also on July 31, 2008, the NRC published in the

Federal Register

a notice of availability of the RIL and NUREG/CR-6967, together with a request for comments (73 FR 44778). In that notice, the NRC stated that these documents and comments on the documents would be discussed at a public workshop to be scheduled in September 2008. The public workshop was held on September 24, 2008, and included presentations and open discussion between representatives of the NRC, international regulatory and research agencies, domestic and international commercial power firms, fuel vendors, and the general public. A summary of the workshop, including a list of attendees and presentations, is available in ADAMS under Accession No. ML083010496. The NRC has not prepared responses to comments received on the technical basis information as a result of the July 31, 2008,

Federal Register

notice (including comments received at the September 2008 public workshop), because: (i) The public workshop was held, in part, to discuss public comments on the technical basis information, and (ii) further opportunity to comment is available during this proposed rule's formal public comment period.

Based upon a preliminary safety assessment in response to the research findings in RIL-0801, the NRC determined that immediate regulatory action was not required, and that changes to the ECCS acceptance criteria to account for these new findings could reasonably be addressed through the rulemaking process. Recognizing that finalization and implementation of the new ECCS requirements would take several years, the NRC completed a more detailed safety assessment that confirmed current plant safety for every operating reactor. See Section III, “Operating Plant Safety,” of this document for further information.

Since 2002, the NRC has met with the Advisory Committee on Reactor Safeguards (ACRS) multiple times to

discuss the progress of the LOCA research program and rulemaking proposals. Provided in the following table are the dates and ADAMS accession numbers of the relevant ACRS meetings and associated correspondence.

Date

Meeting/Letter

ADAMS

October 9, 2002

Subcommittee Meeting

* ML023030246

October 10, 2002

Full Committee Meeting

* ML022980190

October 17, 2002

Letter from ACRS to NRC staff

ML022960640

December 9, 2002

Response letter from NRC staff to ACRS

ML023260357

September 29, 2003

Subcommittee Meeting

* ML032940296

July 27, 2005

Subcommittee Meeting

* ML052230093

September 8, 2005

Full Committee Meeting

* ML052710235

January 19, 2007

Subcommittee Meeting

* ML070390301

February 2, 2007

Full Committee Meeting

ML070430485

May 23, 2007

Letter from ACRS to NRC Staff

ML071430639

July 11, 2007

Response letter from NRC staff to ACRS

ML071640115

December 2, 2008

Subcommittee Meeting

* ML083520501

* ML083530449

December 4, 2008

Full Committee Meeting

* ML083540616

December 18, 2008

Letter from ACRS to NRC staff

ML083460310

January 23, 2009

Response letter from NRC staff to ACRS

ML083640532

May 10, 2011

Subcommittee Meeting

ML111450409

June 8, 2011

Full Committee Meeting

ML11166A181

June 22, 2011

Letter from ACRS to NRC staff

ML11164A048

June 23, 2011

Subcommittee Meeting

ML11193A035

July 13, 2011

Full Committee Meeting

ML11221A059

July 21, 2011

Response letter from NRC staff to ACRS

ML111861706

December 15, 2011

Subcommittee Meeting

ML120100268

January 19, 2012

Full Committee Meeting

ML12032A048

January 26, 2012

Letter from ACRS to NRC Staff

ML12023A089

February 17, 2012

Response Letter from NRC staff to ACRS

ML120260893

* ADAMS file is a transcript of the ACRS meeting.

PRM-50-84

On March 15, 2007, Mark Leyse (the petitioner) submitted a PRM to the NRC (ADAMS Accession No. ML070871368) requesting that all holders of operating licenses for nuclear power plants be required to operate such plants at operating conditions (e.g., levels of power production and light-water coolant chemistries) necessary to effectively limit the thickness of crud

2

and/or oxide layers on fuel rod cladding surfaces. The petitioner requests that the NRC conduct rulemaking in the following three specific areas:

2

For the purpose of this discussion, the NRC defines “crud” as any foreign substance deposited on the surface of the fuel cladding prior to the initiation of a LOCA. It is known that this layer can impede the transfer of heat.

(1) Establish regulations that require licensees to operate light-water power reactors under conditions that are effective in limiting the thickness of crud and/or oxide layers on zirconium-clad fuel in order to ensure compliance with § 50.46(b) ECCS acceptance criteria;

(2) Amend appendix K to 10 CFR part 50 to explicitly require that steady-state temperature distribution and stored energy in the reactor fuel at the onset of a postulated LOCA be calculated by factoring in the role that the thermal resistance of crud deposits and/or oxide layers plays in increasing the stored energy in the fuel (these requirements also need to apply to any NRC-approved, best-estimate ECCS evaluation models used in lieu of appendix K to 10 CFR part 50 calculations); and

(3) Amend § 50.46 to specify a maximum allowable percentage of hydrogen content in (fuel rod) cladding.

On May 23, 2007, the NRC published a notice of receipt for this petition in the

Federal Register

(72 FR 28902) and requested public comment. The public comment period ended on August 6, 2007. Comments in support of PRM-50-84 were provided by the Union of Concerned Scientists, two individuals, and the petitioner. The NEI and Strategic Teaming and Resource Sharing organization submitted comments in opposition to the petition. After evaluating the public comments, the NRC resolved PRM-50-84 by deciding that each of the petitioner's issues should be considered in the rulemaking process. The NRC's determination, including the NRC's response to public comments received on the petition, was published in the

Federal Register

on November 25, 2008 (73 FR 71564). Although there is no direct relationship between the subject of crud and the anticipated new ECCS acceptance criteria requirements, the petition deals with the NRC's requirements on ECCS performance in § 50.46. Given the comprehensive changes to § 50.46 being addressed in this rulemaking, the NRC is considering the petitioner's proposed changes in this rulemaking.

B. Generic Safety Issue (GSI)-191 and Long-Term Cooling

As a result of evolving staff concerns related to the adequacy of PWR recirculation sump designs, the NRC opened Unresolved Safety Issue (USI) A-43, “Containment Emergency Sump Performance.” The resolution of USI A-43 was subsequently documented in Generic Letter (GL) 1985-022, “Potential for Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage,” dated December 3, 1985 (ADAMS Accession No. ML031150731). The NRC staff found in GL 1985-022 that the 50 percent blockage assumption, identified in Regulatory Guide (RG) 1.82, “Sumps for Emergency Core Cooling and Containment Spray Systems,” Revision 0 (ADAMS Accession No. ML111680318), should be replaced with a more comprehensive requirement to assess debris effects on a plant-specific basis. Following the resolution of USI A-43, industry events at Barsebeck and Limerick Generating Station challenged the conclusion that no new requirements were necessary to prevent the clogging of ECCS strainers at operating boiling water reactors (BWR).

As described in NRC Bulletin 95-02, “Unexpected Clogging of a Residual Heat Removal (RHR) Pump Strainer While Operating in Suppression Pool Cooling Mode,” dated October 7, 1995 (ADAMS Accession No. ML082490807), a safety relief valve at the Limerick Generating Station inadvertently opened and could not be closed, the plant was manually scrammed, and the RHR system was started in the suppression pool cooling mode to remove the heat added by the open relief valve. The A train of the RHR exhibited signs of pump cavitation and was secured. The B train of the RHR was started to remove the heat from the relief valve discharge. After the plant was stabilized, a diver inspected the pump suction strainers and found a mat of fibers and sludge covering them. The licensee determined that the discharge from the relief valve did not contribute debris to the suppression pool.

As described in NRC Bulletin 96-03, “Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors,” dated May 6, 1996 (ADAMS Accession No. ML082401219), a Swedish BWR, Barseback Unit 2, experienced plugging of two containment vessel spray system (CVSS) suction strainers. The strainers were partially plugged with mineral wool (a fibrous insulation) that was dislodged by a steam jet from an open pilot operated relief valve. The operators noticed an indication of high-differential pressure across the strainers and were able to back flush them to keep the CVSS operating.

Also described in NRC Bulletin 96-03 are two ECCS suction strainer plugging events that occurred at the Perry Nuclear Power Plant, a BWR located in the United States. The first event resulted from general maintenance material and dirt in the suppression pool collecting on the RHR suction strainers. The differential pressure caused by the debris resulted in deformation of the suction strainers. After the suppression pool was cleaned and the suction strainers replaced, a second event occurred when several safety relief valves lifted. The RHR system was used to cool the suppression pool after the steam discharge. The suction strainers were inspected and found to be covered with fibrous debris and corrosion products. A test of the system found that the B train pump suction pressure dropped to zero. The fibrous debris originated from temporary drywell cooling filter media that was accidentally dropped into the suppression pool and not retrieved. The fibers created a filtering bed on which particles collected, resulting in a high-resistance debris bed.

In response to these events, the NRC issued generic communications requesting that BWR licensees take appropriate actions to minimize the potential for the clogging of ECCS suction strainers by debris accumulation following a LOCA. The NRC staff concluded that all BWR licensees have sufficiently addressed these bulletins in a memorandum, “Completion of Staff Reviews of NRC Bulletin 96-03, `Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors,' and NRC Bulletin 95-02, `Unexpected Clogging of a Residual Heat Removal (RHR) Pump Strainer While Operating in Suppression Pool Cooling Mode',” dated October 18, 2001 (ADAMS Accession No. ML012970229).

The findings regarding BWR strainers prompted the NRC to open GSI-191, “Assessment of Debris Accumulation on PWR Sump Performance,” to ensure that post-accident debris effects would not impede long-term core cooling at PWRs. After completing its technical assessment of GSI-191, the NRC issued Bulletin 2003-01, “Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors,” dated June 9, 2003 (ADAMS Accession No. ML031600259). This bulletin did not require licensees to immediately perform deterministic evaluations for debris effects, but requested that plants take compensatory measures to reduce risk or otherwise enhance the capability of the ECCS and containment spray system (CSS) recirculation functions. The bulletin also informed licensees that the staff was preparing a generic letter that would request that plants demonstrate through deterministic methods that long-term core cooling would not be compromised by debris effects.

Generic Letter 2004-02, “Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors,” dated September 13, 2004 (ADAMS Accession No. ML042360586), was issued to all operating PWRs requesting that they perform a mechanistic evaluation of the effects of debris on the ECCS and CSS recirculation functions. The affected plants are currently working to address the issues identified by the generic letter. All operating PWRs have installed larger strainers and taken other actions toward the final resolution of the issue. Final closure of the generic letter has been delayed to allow industry and the NRC staff to develop appropriate methodologies for evaluation of debris related issues that were identified after the issuance of the generic letter. The staff generated two SECY papers on this issue to provide options and solicit feedback from the NRC Commissioners. On December 14, 2012, the Commission issued an SRM (ADAMS Accession No. ML12349A378) for SECY-12-0093, “Closure Options for Generic Safety Issue—191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance” (ADAMS Accession No. ML121320270). In this SRM, the Commission directed the following:

The forthcoming § 50.46c proposed rulemaking should contain a provision allowing NRC licensees on a case-by-case basis, to use risk-informed alternatives. The license amendment process would be used to reconstitute the long-term core cooling licensing basis. Stakeholder comments should be solicited on the proposed provision.

Consistent with this SRM, the proposed rule includes a provision that would allow licensees to use an alternative risk-informed approach to evaluate the effects of debris for long-term cooling.

III. Operating Plant Safety

A. Emergency Core Cooling System: Embrittlement Research Findings

In response to the research findings in RIL-0801, the NRC performed a preliminary safety assessment of currently operating reactors (“Plant Safety Assessment of RIL-0801 (non-proprietary),” dated February 23, 2009 (ADAMS Accession No. ML090340073)). This assessment found that, due to realistic fuel rod power history, measured cladding performance under LOCA conditions, and current analytical conservatisms, sufficient safety margin exists for operating reactors. Therefore, the NRC staff determined that immediate regulatory action was not required, and that changes to the ECCS acceptance criteria to account for these new findings can reasonably be addressed through the rulemaking process.

Recognizing that finalization and implementation of the new ECCS requirements would take several years, the NRC decided that a more detailed safety assessment was necessary. As a voluntary industry effort, the PWR Owners Group (OG) (“Letter Report: OG-11-143 PWROG 50.46(b) Margin Assessment,” dated April 29, 2011 (ADAMS Accession No. ML11139A309)) and BWR OG (“BWROG-TP-11-010 (Rev. 1) Evaluation of BWR LOCA Analyses and Margins Against High Burnup Fuel

Research Findings,” dated June 2011 (ADAMS Accession No. ML111950139)), under the auspices of NEI, submitted ECCS margin assessment reports. After grouping plants based on similar design features, cladding alloys, or ECCS evaluation models and defining cladding alloy-specific analytical limits, the OG reports identified analytical credits or performed new LOCA analyses necessary to demonstrate that the limiting plant within each grouping had positive margin relative to the research findings. The NRC conducted an audit of the OG reports and supporting General Electric—Hitachi (GEH), AREVA, and Westinghouse engineering calculations. Based on the OG reports and supplemental information collected during the audits, the NRC was able to confirm, for every operating reactor, current safe operation. As documented in the audit report and safety assessment (“ECCS Performance Safety Assessment and Audit Report,” dated February 10, 2012 (ADAMS Accession No. ML12041A078)), the NRC intends to verify, on an annual basis, continued safe operation until each licensee has implemented the new ECCS requirements. See Section V.E, “Implementation,” of this document for the staff-recommended implementation plan developed based on this information.

B. GSI-191 and Long-Term Core Cooling

Section II. B., “GSI-191 and Long-Term Cooling,” of this document provides background information on GSI-191 and long-term cooling. That section includes information on action taken by the NRC and licensees to address the potential effects of debris on long-term cooling. These actions have contributed significantly to the safety of operating plants. The NRC staff provided information to the Commission in two SECY papers: SECY-10-0113, “Closure Options for Generic Safety Issue—191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance,” dated August 26, 2010 (ADAMS Accession No. ML101820296); and SECY-12-0093, “Closure Options for Generic Safety Issue—191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance,” dated July 9, 2012 (ADAMS Accession No. ML12130270).

The Commission issued guidance for the closure of the issue in two SRMs associated with each SECY paper. The SRM to SECY-10-0113 (“Staff Requirements—SECY-10-0113—Closure Options for Generic Safety Issue—191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance” (ADAMS Accession No. ML103570354)) was issued on December 23, 2010. With respect to operating plant safety the SRM stated:

The staff should take the time needed to consider all options to a risk-informed, safety conscious resolution to GSI-191. While they have not fully resolved this issue, the measures taken thus far in response to the sump-clogging issue have contributed greatly to the safety of U.S. nuclear power plants. Given the vastly enlarged advanced strainers installed, compensatory measures already taken, and the low probability of challenging pipe breaks, adequate defense-in-depth is currently being maintained.

On December 14, 2012, the Commission issued the SRM to SECY-12-0093 (ADAMS Accession No. ML12349A378). With respect to operating plant safety, the SRM reiterated the direction in SRM-SECY-10-0113.

As directed by the Commission, the NRC staff is currently working with licensees to assure adequate safety by closing the issue and updating their licensing bases to reflect full compliance on a schedule consistent with Commission direction.

IV. Advance Notice of Proposed Rulemaking: Public Comments

On August 13, 2009, the NRC published an Advance Notice of Proposed Rulemaking (ANPR) (74 FR 40767) to obtain stakeholder views on issues associated with amending § 50.46(b). The ANPR indicated that the proposed scope of the rulemaking included four major objectives: (1) Expand the applicability of § 50.46 to include any light-water reactor fuel cladding material; (2) establish performance-based requirements and acceptance criteria specific to zirconium-based cladding materials that reflect research findings; (3) revise the LOCA reporting requirements; and (4) address the issues raised in PRM-50-84 that relate to crud deposits and hydrogen content in fuel cladding. The ANPR provided interested stakeholders an opportunity to comment on the options under consideration by the NRC during a 75-day public comment period. In addition, the NRC asked 12 specific questions in the following categories: Applicability Considerations, New Embrittlement Criteria Considerations, Testing Considerations, Revised Reporting Requirements Considerations, Crud Analysis Considerations, and Cost Considerations. The public comment period ended on October 27, 2009.

The NRC received a total of 19 comment letters during the ANPR's public comment period; these letters were sent from a variety of entities, including one comment from a private citizen, 15 comments from the nuclear industry, one comment from a non-governmental organization, and two comments from the international community. The NRC held a public meeting on April 28-29, 2010, to discuss, among other things, the public comments received on the ANPR. No additional public comments were accepted at this public meeting. The meeting summary is available in ADAMS under Accession No. ML101300490.

As a result of comments received on the ANPR, the NRC has made a number of changes to the proposed rule. A detailed discussion of the public comments submitted on the ANPR, including a detailed list of commenters, is contained in a separate document, “Section 50.46c and PRM-50-71 Comment Response Document” (ADAMS Accession No. ML12283A213). The most significant changes as the result of public comments are:

• The specific experimental technique for measuring cladding ductility (i.e.,

>

1.00 percent permanent strain prior to failure during ring-compression loading at a temperature of 135 °C and a displacement rate of 0.033 millimeters per second (mm/sec)) was removed from the rule and provided as one approved method within DG-1262, “Testing for Postquench Ductility” (ADAMS Accession No. ML12284A325).

• The specific experimental technique for measuring time until breakaway oxidation (i.e., hydrogen uptake reaches 200 weight part per million (wppm) anywhere on a cladding segment subjected to high-temperature steam oxidation ranging from 1200 °F to 1875 °F (649 °C to 1024 °C)) was removed from the rule and provided as one approved method within DG-1261, “Conducting Periodic Testing for Breakaway Oxidation Behavior” (ADAMS Accession No. ML12284A324).

• The proposed risk-informed change to the reporting requirements (objective three of the ANPR) was abandoned. The majority of public comments received on the proposed reporting criteria suggested that the concept was complex, and might promote unnecessary burden or misinterpretation.

• The applicability of the zirconium-based alloy fuel specific performance requirements was expanded to include uranium-plutonium mixed oxide fuel.

• The applicability of the post-quench ductility (PQD) analytical limits in DG-1263, “Establishing Analytical Limits for Zirconium-Based Alloy Cladding” (ADAMS Accession No. ML12284A323), was expanded to

encompass cladding hydrogen concentration up to 800 wppm.

• Many changes and improvements were made in the development of DG-1261, DG-1262, and DG-1263.

• A staged implementation plan was developed.

V. Proposed Requirements for ECCS Performance During LOCAs

The proposed rule would establish a general, performance-based rule governing ECCS performance for LWRs, regardless of fuel design or cladding material. This represents a significant change from the current ECCS regulations, which apply to “uranium oxide pellets within cylindrical zircaloy or ZIRLO

TM

cladding.” Because ECCS system requirements must be expressed independent of fuel type, and because ECCS system performance ultimately must be based upon maintaining the fuel in the reactor in a safe (analyzed) condition, the proposed rule separates the ECCS system requirements from the need for the applicant/licensee to establish the fuel system design performance criteria constituting a safe condition.

In proposed § 50.46c, the specified performance objectives of the systems, structures, and components of the ECCS are to provide residual heat removal during and following a postulated LOCA. As with the current regulations, the ECCS performance is demonstrated by NRC-approved ECCS evaluation models in proposed § 50.46c. Specific performance requirements and analytical limits have been established for fuel designs consisting of uranium oxide or mixed uranium-plutonium oxide pellets within zirconium cladding alloys that account for recent research findings. New performance objectives and analytical limits may be necessary for other fuel designs to take into consideration all degradation mechanisms and any unique features of the particular fuel system that the ECCS is trying to cool.

The proposed rule follows the general regulatory approach of the existing regulations by establishing non-prescriptive, performance-based regulatory language for demonstrating acceptable ECCS system performance and determining the fuel's performance characteristics. The organization and 10 CFR designations of the NRC's requirements governing ECCS (currently in § 50.46) and reactor cooling venting systems (currently in § 50.46a) are expected to change, as a result of: (1) Ongoing rulemaking activities; (2) the proposed implementation schedule for those activities; and (3) the need to maintain the current requirements in place for those licensees that have not transitioned to the new requirements (following the implementation schedule that would be provided in the final rule). A detailed description of the transition of 10 CFR designations is provided in Section VI, “Section-by-Section Analysis,” of this document.

A. Applicability of Performance-Based Rule: Consideration of PRM-50-71

The NRC proposes to expand the applicability of the rule from “uranium oxide pellets within cylindrical zircaloy or ZIRLO

TM

cladding” to any LWR, regardless of fuel design or cladding material. The proposed rule would be applicable to applicants for and holders of construction permits, operating licenses, combined licenses, and standard design approvals and to applicants for certified designs and for manufacturing licenses. The rule would not apply to any licensee that has submitted certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, in accordance with § 50.82(a)(1).

Over the past 10 years, the NRC has granted exemptions from the requirements of § 50.46 (in accordance with § 50.12(a)) to licensees utilizing approved fuel designs with M5 zirconium-based alloy cladding and, more recently, to licensees using approved fuel designs with Optimized ZIRLO

TM

zirconium-based alloy cladding.

The proposed rule includes general performance requirements for future LWR fuel designs and specific performance requirements for the current generation of LWR fuel designs with zirconium-based alloy claddings. As such, it is anticipated that future exemption requests would not be necessary for loading an advanced fuel design or cladding material approved by the NRC through a rulemaking. However, the licensee would still need to submit a license amendment. During this approval process the NRC would determine whether, either: (1) Specified and NRC-approved analytical limits have been established, along with an NRC-approved ECCS evaluation model, which satisfy the specific performance-based requirements for fuel designs consisting of uranium oxide or mixed uranium-plutonium oxide pellets within zirconium-based alloy cladding material; or (2) specified performance objectives and associated analytical limits which take into consideration all degradation mechanisms and any unique features of the particular fuel system have been established, along with an NRC-approved ECCS evaluation model, by which to judge the ECCS performance for new fuel designs.

The NRC recognizes that a small number of fuel rods may experience cladding failuare (i.e., small perforation) during normal operation due to manufacturing defects, debris fretting, grid-to-rod fretting, etc. The allowable number of fuel rod failures during normal operation is not governed by ECCS performance requirements, but limited by 10 CFR part 20, “Standards for Protection against Radiation,” and plant Technical Specifications, which limit reactor coolant activity level to maintain on-site and off-site dose during normal operation, anticipated operational occurrences, and postulated accidents to within prescribed limits. In addition to Technical Specifications limitations, plant administrative limits on reactor coolant activity level further reduce the potential number of failed fuel rods within an operating core.

Due to secondary degradation effects, the performance of these limited failed fuel rods during a postulated LOCA may be difficult to predict, and would most likely be outside the experimental database used to set the NRC-approved analytical limits for coolable geometry (i.e., cladding embrittlement for zirconium-based alloys). However, due to their limited number relative to the total core population, any unforeseen degradation or performance during a postulated LOCA would not challenge the general performance requirements. As such, compliance with ECCS performance requirements of § 50.46c is not required for this limited number of failed fuel rods.

This proposed extension to all LWR fuel types addresses PRM-50-71, which requested that the applicable regulations be amended to allow for the introduction of advanced zirconium-based alloy claddings, thus eliminating the need for a licensee to pursue an exemption for alloys which did not meet the definition of “zircaloy or ZIRLO

TM

.” If the NRC adopts the proposed rule in final form, PRM-50-71 would be granted and resolved.

B. Performance-Based Aspects of the Proposed Rule

The systems, structures, and components of the ECCS are designed to provide residual heat removal during and following a postulated LOCA. Failure of the ECCS to perform its intended function would result in a loss of coolable geometry followed by core reconfiguration. While the principal ECCS performance requirements are simple in nature (i.e., remove residual heat and maintain core temperatures at acceptable levels), the system must be

designed to achieve specified performance objectives, taking into consideration all degradation mechanisms and any unique features of the particular fuel system that the ECCS is intended to cool. Sufficient empirical data must be available for the particular fuel system to identify all degradation mechanisms (e.g., embrittlement, loss of structural integrity) and any unique features (e.g., eutectic or exothermic reactions, combustible gas generation) to specify both acceptable core temperatures and the duration for which the ECCS must remove residual heat. In addition, fuel-specific analytical requirements may be necessary to accurately or conservatively model unique phenomena that impact the ECCS performance demonstration (e.g., fuel rod balloon and burst, cladding inside-diameter oxygen ingress).

To achieve the NRC's goal of a more performance-based rule, significant changes in format and structure are being proposed relative to § 50.46. In place of the current prescriptive § 50.46(b) analytical limits, the proposed rule would define the following principal ECCS performance requirements:

• Core temperature during and following the LOCA event does not exceed the analytical limits for the fuel design used for ensuring acceptable performance. This ensures that the fuel maintains a coolable geometry.

• Sufficient cooling so that decay heat will be removed for the extended period of time required by the long-lived radioactivity remaining in the core so that long-term cooling is ensured.

Complying with these performance requirements provides reasonable assurance that the overall objective of maintaining a coolable core geometry in the event of a LOCA is met. In addition, the proposed rule would dictate specific analytical requirements for demonstrating compliance with the ECCS performance requirements. For instance, to demonstrate compliance with these system performance requirements, ECCS performance would be evaluated using fuel-specific performance objectives and associated analytical limits that take into consideration all degradation mechanisms and unique features of the particular fuel system, along with an NRC-approved evaluation model.

The proposed rule includes specific performance requirements for fuel designs consisting of uranium oxide or mixed uranium-plutonium oxide fuel pellets within cylindrical zirconium-alloy cladding. These performance requirements incorporate the findings of the NRC LOCA research program. New performance objectives and analytical limits may be necessary for other fuel designs.

For uranium oxide or mixed uranium-plutonium oxide fuel pellets within cylindrical zirconium-alloy cladding, all known degradation mechanisms and unique features have been identified, specific performance objectives have been defined, and fuel design-specific performance requirements have been established and included in the proposed rule. For this fuel system design, the performance objective is to maintain the coolable fuel rod bundle array. In other words, the objective is to maintain fuel pellets within the cladding and fuel rods within the fuel bundle lattice. Existing ECCS models and methods are capable of accurately predicting core temperatures and demonstrating ECCS performance, provided this core configuration is maintained. To achieve this performance objective, the ECCS must limit core temperatures to prevent high-temperature cladding failure, prevent brittle cladding failure (i.e., maintain PQD and prevent breakaway oxidation), minimize hydrogen gas generation, and provide for long-term residual heat removal for the long-lived fission decay products associated with uranium oxide or uranium-plutonium oxide fuel.

The following § 50.46(b) requirements would remain unchanged in the proposed § 50.46c:

•

Peak cladding temperature.

The calculated maximum fuel element cladding temperature shall not exceed 2200 °F. The peak cladding temperature requirements currently in § 50.46(b)(1) would be moved to § 50.46c(g)(1)(i).

•

Maximum hydrogen generation.

The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react. The maximum hydrogen generation limits currently in § 50.46(b)(3) would be moved to § 50.46c(g)(1)(iv).

In the current regulations, the preservation of cladding ductility, via compliance with regulatory criteria on peak cladding temperature (§ 50.46(b)(1)) and local cladding oxidation (§ 50.46(b)(2)), provides a level of assurance that fuel cladding will not experience gross failure and that the fuel rods will remain within their coolable lattice arrays. The recent LOCA research program identified new cladding embrittlement mechanisms that demonstrated that the current combination of peak cladding temperature (2200 °F (1204 °C)) and local cladding oxidation (17 percent equivalent cladding reacted (ECR)) criteria may not always ensure PQD. The impact of these research findings on cladding ductility is addressed in the following section.

1. Hydrogen-Enhanced Beta-Layer Embrittlement

As explained in Section 1.4 of NUREG/CR-6967, oxygen diffusion into the base metal under LOCA conditions promotes a reduction in the size (referred to as beta-layer thinning) and ductility (referred to as beta-layer embrittlement) of the metallurgical structure within the cladding that provides its macroscopic mechanical behavior. The presence of hydrogen within the cladding enhances this embrittlement process.

It is important to recognize that the embrittlement of the cladding is the result of oxygen diffusion into the base metal and not directly related to the rate of growth or overall thickness of a zirconium dioxide layer on the outside cladding diameter. In combination with a limit on peak cladding temperature, the current regulation limits maximum local oxidation to preserve cladding ductility. Maximum local oxidation is used as a surrogate to limit the ITT and associated oxygen diffusion. This surrogate approach is possible because both the rate of oxidation and rate of oxygen diffusion share strong temperature dependence. In the recent LOCA research program, the Cathcart-Pawel (CP) weight gain correlation was used to integrate time-at-temperature and define the point at which ductility was lost (nil ductility). Section 1.3 of NUREG/CR-6967 defines the following equations used to integrate time-at-temperature:

EP24MR14.001

Measurements of weight gain were performed on many of the steam-oxidized cladding samples tested in the LOCA research program. For example, Table 22 of NUREG/CR-6967 provides both measured ECR and calculated Cathcart-Pawel Equivalent Cladding Reacted (CP-ECR) for the zircaloy-2 cladding samples tested. Instead of correlating measured plastic strain or measured offset displacement with measured ECR or measurements of the post-quench cladding microstructure (e.g., beta layer thickness), the research findings correlate the ductile-to-brittle transition to calculated CP-ECR (using the equations previously stated). In this instance, calculated ECR is used to integrate time-at-temperature and requires knowledge of measured ECR. However, an accurate or conservative weight gain model based on measured oxidation, which may be alloy-specific or vary significantly from CP predictions, needs to be used for predicting rate of energy release and hydrogen generation from the metal/water reaction in the LOCA heat balance calculation.

In an attempt to more accurately characterize the degrading phenomenon, the proposed rule would replace the term “maximum local oxidation” with “ITT,” which more directly relates to the parameter of interest (i.e., embrittlement due to oxygen diffusion). This should clarify the need to have: (1) An accurate or conservative weight gain correlation based on measured oxidation for estimating the rate of energy release and hydrogen generation from the metal/water reaction, and (2) a consistent analytical technique to integrate time-at-temperature in both the empirical database (i.e., allowable CP-ECR) and evaluation model (i.e., predicted CP-ECR).

During normal operation, the cladding metal absorbs some hydrogen from the corrosion process. When that cladding is exposed to high-temperature LOCA conditions, the elevated hydrogen levels increase the solubility of oxygen in the beta phase and the rate of diffusion of oxygen into the beta phase. Therefore, even for LOCA temperatures below 1204 °C (2200 °F), embrittlement can occur for time periods corresponding to less than 17-percent oxidation in corroded cladding with significant hydrogen pickup.

Figure 1 illustrates the effect of hydrogen on ring-compression test ductility measurements. Test specimens included high-burnup (a 71- to 74-micrometer corrosion-layer thickness) and as-fabricated (fresh) PWR Zircaloy-4 cladding segments. Cladding samples were oxidized on two sides at approximately 1200 °C (~2200 °F) and cooled at approximately 11 °C per second to 800 °C (1472 °F). As-fabricated samples were quenched at 800 °C, whereas the high-burnup samples were slow-cooled from 800 °C to room temperature.

Figure 1 plots ECR (a parameter correlated with oxygen pickup from the steam) as calculated by the CP-ECR kinetics correlation vs. the offset strain accommodated before cracking in ring compression testing. The offset strain before cracking indicates sample ductility and an offset strain less than 2 percent is considered brittle. Multiple ring compression tests were conducted using rings that had been oxidized to a range of CP-ECR levels from 0-16 percent. The results indicate that high burnup cladding material embrittles more rapidly than fresh material. For these tests, an ECR of 7 percent (where the high burnup material indicated brittle behavior) corresponds to a total (integral) oxidation time of ~155 seconds, while an ECR of 14 percent (where the fresh material first indicated brittle behavior) corresponds to ~300 seconds.

EP24MR14.002

To address this phenomenon (as well as to achieve a more performance-based rule), the NRC proposes to replace the existing prescriptive analytical limits with a performance-based requirement that would require licensees to establish specified and NRC-approved analytical limits on PCT and ITT. These limits should correspond to the measured ductile-to-brittle transition for the zirconium-based alloy cladding based upon an NRC-approved experimental technique. If the peak cladding temperature that preserves cladding ductility is lower than the 2200 °F limit, the licensee should use the lower temperature.

The NRC is issuing draft regulatory guide DG-1263 for comment. The draft regulatory guide provides licensees with “specified and NRC-approved analytical limits on PCT and ITT,” based upon the NRC's LOCA research program's measured ductile-to-brittle transition for zirconium-based alloy cladding. In addition, the NRC is issuing DG-1262 for comment, which provides licensees with “an NRC-approved experimental technique” for conducting PQD measurements and developing analytical limits. These DGs specify an approach acceptable to the NRC. Even if the draft regulatory guides are adopted in final form, licensees may propose alternative approaches to those described in those regulatory guides.

It is important to recognize that a consistent integration technique should be used to quantify time at elevated temperature in both the experiments and evaluation model. For example, the NRC-approved analytical limits on ITT in DG-1263 were based on the NRC's LOCA research program results, which, in turn, integrated time at elevated temperature using the CP weight gain correlation. For consistency with DG-1263, future LOCA analyses should integrate time at elevated temperature using the same CP weight gain correlation when comparing analysis results against these analytical limits. For this case, appendix K to 10 CFR part 50 ECCS evaluation models would continue to use the Baker-Just (BJ) weight gain correlation for estimating the rate of energy release and hydrogen generation from the metal/water reaction.

The NRC's LOCA research program did not investigate cladding degradation mechanisms or develop the technical basis for performance-based requirements beyond the existing 2200 °F peak cladding temperature criterion. Examples of degradation mechanisms beyond cladding embrittlement (via oxygen diffusion) include excessive exothermic metal-water reaction, alloy-specific eutectics, and loss of fuel rod geometry due to plastic flow. As a result, the existing 2200 °F limit (specified in § 50.46c(g)(1)(i) of the proposed rule) remains an absolute upper limit for zirconium-based alloys on PCT. However, as reflected in this proposed requirement, a lower PCT may be required to preserve ductility.

2. Oxygen Ingress From Cladding Inside Diameter

Oxygen sources may be present on the inner surface of irradiated cladding due to gas-phase UO

3

transport prior to gap closure, fuel-cladding-bond formation (uranium dioxide in solid solution with zirconium dioxide), and the fuel bonded to this layer. Under LOCA conditions, this available oxygen may diffuse into the base metal of the cladding, effectively reducing the integral time-at-temperature to nil ductility.

To address this phenomenon, the NRC proposes to add an analytical requirement to the ECCS evaluation model that would require licensees to, if an oxygen source is present on the inside surfaces of the cladding at the onset of a LOCA, consider the effects of oxygen diffusion from the cladding inside surfaces in the ECCS evaluation model.

The NRC recognizes that the availability of a cladding inside diameter (ID) oxygen source and its diffusion into the base metal during a postulated LOCA may depend on several factors (e.g., rod design, power history). As such, applicants are responsible for determining when the fuel-cladding bonding layer is strong enough to allow the diffusion of oxygen from the uranium-oxide fuel to the zirconium cladding and, therefore, must be included in the ECCS evaluation model. It is anticipated that identifying the magnitude and onset of oxygen ID diffusion would be part of the NRC's review and approval of LOCA

evaluation models or vendor fuel designs. A conservative analytical limit is provided in draft regulatory guide DG-1263.

3. Breakaway Oxidation

As explained in Section 1.4.5 of NUREG/CR-6967, zirconium dioxide can exist in several crystallographic forms (allotropes). The normal tetragonal oxide that develops under LOCA conditions is dense, adherent, and protective with respect to hydrogen pickup. However, there are conditions that promote a transformation to the monoclinic phase (i.e., the phase that is grown during normal operation), which is neither fully dense nor protective. The tetragonal-to-monoclinic transformation is an instability that initiates at local regions of the metal-oxide interface and grows rapidly throughout the oxide layer. Because this transformation results in an increase in oxidation rate, it is referred to as breakaway oxidation. Along with this increase in oxidation rate resulting from cracks in the monoclinic oxide, significant hydrogen pickup also occurs. Hydrogen that enters in this manner during a LOCA transient promotes rapid embrittlement of the cladding.

While all zirconium alloys will eventually experience breakaway oxide phase transformation when exposed to long durations of high-temperature steam oxidation, alloying composition and manufacturing process (e.g., surface roughness) influence the timing of this phenomenon.

Any fuel rod that experiences breakaway oxidation during a postulated LOCA will rapidly become brittle and more susceptible to gross failure and hence, is no longer in compliance with General Design Criteria (GDC)-35 requirements for coolable core geometry. To address this phenomenon, the NRC proposes to add a performance-based requirement that the licensee measure the onset of breakaway oxidation for each reload batch on manufactured cladding material and report any changes in the onset of breakaway oxidation at least annually. This requirement, along with a periodic test requirement, would confirm that slight composition changes or manufacturing changes have not inadvertently altered the cladding's susceptibility to oxidation. The NRC is issuing DG-1261, which will provide licensees with “an NRC approved experimental technique” for conducting breakaway oxidation measurements and developing analytical limits. Even if the draft regulatory guide is finalized, licensees may also provide an alternative approach to that proposed in the draft regulatory guide.

4. Applicability of Ductility-Based Analytical Limits in the Burst Region

During a postulated LOCA, a portion of the fuel rod population may be predicted to experience fuel rod ballooning and cladding rupture as a result of rapid depressurization of the reactor coolant system in combination with elevated cladding temperature. The number of burst rods depends on several variables including initial conditions (e.g., fuel rod design, rod internal pressure, rod power) and accident conditions (e.g., break size, cladding temperature). This flawed section of the fuel rod may experience degradation mechanisms beyond oxygen diffusion embrittlement encountered in the remaining portions of the fuel rod, including significant amounts of hydrogen uptake from steam entering the fuel rod through the rupture.

Consistent with the technical basis of the proposed rule, DG-1262 describes an NRC-approved experimental technique for defining the ductile-to-brittle transition. This experimental procedure involves measuring ductility using ring compression testing performed on small, unflawed segments of fuel rod cladding previously exposed to steam oxidation at a defined peak cladding temperature and the integrated time at temperature profile (expressed as CP-ECR). While this experimental approach captures embrittlement of the zirconium metal due to oxygen diffusion and the effects of pre-existing hydrogen on the rate of embrittlement, it does not capture all of the degradation mechanisms experienced in the region of the fuel rod surrounding a cladding rupture. In addition to embrittlement due to oxygen ingress (which is doubled in the burst region due to steam entering cladding rupture), the burst region experiences cladding wall thinning, cladding rupture, and increased hydrogen uptake (hydrogen absorbed from zirconium oxidation on the cladding ID). All of these degradation mechanisms impact the performance of the fuel rod under LOCA conditions. As such, the ductile-to-brittle transition based on ring compression tests of unflawed cladding segments may not fully represent the region of the fuel rod surrounding the cladding rupture.

The rupture region contains non-uniform distributions of: (1) Oxygen concentration within the base metal and zirconium oxide thickness, (2) soluble hydrogen and zirconium hydrides, (3) cladding wall thickness (due to ballooning), and (4) cladding flaws (due to ballooning and rupture). The overall goal of preserving cladding ductility may not apply to the rupture area that contains non-uniform distributions of flaws, cladding thickness, hydrogen distribution, and oxygen levels.

To investigate the mechanical behavior of ruptured fuel rods, the NRC conducted integral LOCA testing, designed to exhibit ballooning and burst, on as-fabricated and hydrogen-charged cladding specimens and high-burnup fuel rod segments exposed to high-temperature steam oxidation followed by quench. The research results and conclusions are documented in the report “Mechanical Behavior of Ballooned and Ruptured Cladding” (ADAMS Accession No. ML12048A475). The integral LOCA testing confirms that continued exposure to a high-temperature steam environment weakens the already flawed region of the fuel rod surrounding the cladding rupture. Hence, limitations on PCT and ITT are necessary to preserve an acceptable amount of mechanical strength and fracture toughness. In addition, this research demonstrated that the degradation in strength and fracture toughness with prolonged exposure to steam oxidation was enhanced with pre-existing cladding hydrogen content.

The research findings from the integral LOCA research presented the NRC with two options for revising the fuel performance requirements: (1) Establish a separate performance requirement within the burst region (i.e., analytical limits that preserve sufficient fracture toughness to ensure burst region survival), or (2) apply the ductility-based analytical limits to the entire fuel rod.

In the absence of a credible analysis of loads, cladding stresses, and cladding strains for a degraded LOCA core, there are no absolute metrics to determine how much ductility or strength would be needed to “guarantee” that fuel-rod cladding would maintain its geometry during and following LOCA quench. It is also not clear what impact severance of some fuel rods into two pieces would have on core coolability. Fragmentation of fuel rod cladding would be more detrimental to core coolability than severance of rods into two pieces. Even minimal ductility ensures that cladding will have high strength and toughness and therefore, high resistance to fracturing. Brittle cladding, on the other hand, might fail at low strength and shatter. Therefore, the intent to maintain ductility is beneficial even without adequate knowledge of LOCA loads. If wall thinning and double-sided oxidation are accounted for, then it was determined that applying the hydrogen-

based embrittlement limit developed in previous work at ANL to limit oxidation in the balloon region of the irradiated fuel rods tested at Studsvik was sufficient to preserve reasonable behavior of the ballooned and ruptured region.

The integral LOCA research concluded that application of the hydrogen-dependent ductility-based analytical limits on PCT and ITT (when applied within the burst region) preserve the mechanical behavior of high-burnup rods tested to that measured for as-fabricated cladding oxidized to 17 percent CP-ECR. Assuming highly conservative upper bounds on thermal expansion loading during quench, the residual mechanical behavior preserved by this limit was determined to be adequate to demonstrate that coolable geometry is maintained. As such, the NRC elected the second regulatory approach to apply a single performance-based requirement to the entire fuel rod. This decision recognizes that portions of the cladding within the burst region may not maintain ductility. This decision is reflected in DG-1263 and supported by the technical basis documented in the staff report,

“

The Mechanical Behavior of Ballooned and Ruptured Cladding” (ADAMS Accession No. ML12048A475).

5. Long-Term Cooling

The current regulation in § 50.46(b)(5) requires that for long-term cooling the calculated core temperature be maintained at an acceptably low value following any calculated successful initial operation of the ECCS. It also requires that decay heat be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

The proposed rule would define a performance-based requirement to ensure acceptable fuel performance during long-term cooling. Specifically, the proposed rule would require that a specified and NRC-approved analytical limit on peak cladding temperature be established that corresponds to the measured ductile-to-brittle transition for the zirconium-based alloy cladding material based upon an NRC-approved experimental technique. It would also require that the calculated maximum fuel element temperature should not exceed the established analytical limit.

6. Use of Risk-Informed Approaches To Address Debris for Long-Term Cooling

The proposed rule would allow all entities to use an alternative risk-informed approach to evaluate the effects of debris for long-term cooling. The adverse effects of debris on ECCS performance have been documented in the NRC's actions to resolve GSI-191, “Assessment of Debris Accumulation on PWR Sump Performance.” Debris may cause increased head loss across the ECCS and CSS pump suction strainer and restrict the flow of water to the ECCS and CSS pumps. Debris may also pass through the strainer and cause blockage of components or the core, or damage to components downstream of the strainer. For these reasons, the effects of debris on long-term ECCS cooling performance must be evaluated. However, the NRC believes that risk-informed methodologies have progressed to the point where the NRC may allow their use in considering the effects of debris on the adequacy of long-term ECCS cooling performance. The entity's application and the NRC's review and approval of the application will close that entity's required actions under GSI-191.

For the purpose of § 50.46c provisions on the risk-informed alternative to long-term cooling, debris is material within containment that may be transported to the suction strainer(s) for the ECCS and CSS. Debris includes (but is not limited to) loose materials that may transport and materials that may be damaged by a LOCA jet to the extent that they become transportable. Debris sources of interest typically include insulation, coatings, dust, dirt, concrete, fire barrier material, signs and tags, and materials left in containment; however, debris may originate from other sources. Debris may also result from chemical interactions that cause precipitation of materials. Debris may cause increased head loss across the strainer and restrict the flow of water to the ECCS and CSS pumps. Debris may also pass through the strainer and cause blockage of components or the core, or damage to components downstream of the strainer.

The proposed § 50.46c provisions allowing a risk-informed approach for evaluating the effects of debris on long-term cooling performance would require that the defense-in-depth philosophy and safety margins be maintained and, as a result, defense-in-depth and safety margins must be explicitly considered. This consideration of defense-in-depth and safety margins is consistent with the NRC's general guidance regarding risk-informed decisionmaking contained in RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant Specific Changes in the Licensing Basis,” Revision 2, dated May 2011 (ADAMS Accession No. ML100910006). The RG 1.174 provides guidance on an acceptable approach to risk-informed decision-making, consistent with the Commission's Policy Statement on the Use of Probabilistic Risk Assessment (PRA) dated August 16, 1995 (60 FR 42622). The RG sets forth a set of five key principles, four of which are relevant to the proposed rule:

• Maintain the defense in depth philosophy;

• Maintain sufficient safety margins;

• Any changes allowed must result in no more than a small increase in core damage frequency or risk, consistent with the intent of the Commission's Safety Goal Policy Statement; and

• Incorporate monitoring and performance measurement strategies.

The proposed rule is consistent with the defense in depth principle of RG 1.174. Defense-in-depth has traditionally been applied in reactor design and operation to provide multiple means of accomplishing safety functions and to prevent the release of radioactive material. The applicant would need to address the intent of the general design criteria (or similar licensing basis design criteria), national standards, and engineering principles (e.g., single failure criterion) in evaluating the impact of the alternative approach on defense-in-depth. Defense-in-depth is considered sufficient if the overall redundancy and diversity among the plant's systems and barriers, including the containment and its support systems, is sufficient to ensure that the risk acceptance criteria of § 50.46c(e)(1)(i) are met, and the following attributes are maintained:

• Reasonable balance is preserved among prevention of core damage, prevention of containment failure or bypass, and mitigation of consequences of an offsite release.

• There is not an over-reliance on programmatic activities to compensate for weaknesses in plant design.

• System redundancy, independence, and diversity are preserved commensurate with the expected frequency of challenges, consequences of failure of the system, and associated uncertainties in determining these parameters.

• Defenses against potential common cause failures are preserved and the potential for the introduction of new common cause failure mechanisms are assessed and addressed.

• Independence of barriers is not degraded.

• Defenses against human errors are preserved.

• The intent of the plant's design criteria is maintained.

Regarding the maintenance of sufficient safety margins, the applicant would need to address the impact of implementing the alternate approach on

current safety margins. Consistent with RG 1.174, Revision 2, sufficient safety margins are considered to be maintained when:

• Codes and standards or their alternatives approved for use by the NRC are met.

• Safety analysis acceptance criteria in the licensing basis are met or proposed revisions provide sufficient margin to account for analysis and data uncertainty.

The risk-informed provisions for considering the effects of debris on long-term cooling would also require that any potential net increase in risk from implementation of the risk-informed approach be assessed and that reasonable confidence is provided that this change in risk is small. The NRC regards “small” changes for plants with total baseline core damage frequencies (CDF) of 10

−

4

per year or less to be CDF increases of up to 10

−

5

per year and plants with total baseline CDF greater than 10

−

4

per year to be CDF increases of up to 10

−

6

per year. However, if there is an indication that the CDF may be considerably higher than 10

−

4

per year, the focus of the applicant should be on finding ways to decrease rather than increase CDF and the licensee may be required to present arguments as to why steps should not be taken to reduce CDF in order for the alternate approach to be considered. For plants with total baseline large early release frequency (LERF) of 10

−

5

per year or less, small LERF increases are considered to be up to 10

−

6

per year, and for plants with total baseline LERF greater than 10

−

5

per year, small LERF increases are considered to be up to 10

−

7

per year. Similar to the CDF metric, if there is an indication that the LERF may be considerably higher than 10

−

5

per year, the focus of the licensee should be on finding ways to decrease rather than increase LERF and the licensee may be required to present arguments as to why steps should not be taken to reduce LERF in order for the alternate approach to be considered. This perspective is consistent with the guidance in Section 2.2.4 of RG 1.174, Revision 2.

Finally, § 50.46c contains requirements that would ensure that the plant-specific PRA is of sufficient scope, level of detail, and technical adequacy for this approach and is updated and maintained over time and that the risk-informed approach is evaluated periodically. The technical adequacy of the plant-specific PRA would be assessed by the NRC taking into account appropriate standards and peer review results. The NRC has prepared an RG (RG 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,” dated March 2009 (ADAMS Accession No. ML090410014)) on determining the technical adequacy of PRA results for risk-informed activities. As one step in the assurance of technical adequacy, the PRA must have been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC. Therefore, the NRC staff would rely on the NEI Peer Review Process, as modified in the NRC's approval, or the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Peer Review Process, as modified in the NRC's approval; both processes are documented in RG 1.200. Changes and data, including: (1) Operational practices; (2) the facility configuration; (3) plant and industry experience; and (4) structure, system, and component (SSC) performance would be required to be fed back into the PRA and the § 50.46c risk-informed analyses and, when appropriate, adjustments would be made to maintain the validity of these processes. In addition, § 50.46c contains requirements for corrective action and reporting, to the NRC, conditions where the established risk-informed approach results exceed the risk acceptance criteria. Together, these requirements would maintain the validity of the risk-informed approach such that the risk-informed decisionmaking principles would continue to be satisfied over the life of the facility.

In as much as § 50.46c contains requirements that would (1) provide reasonable confidence that any net risk increase from implementation of its requirements is small; (2) maintain defense-in-depth; (3) maintain safety margins; and (4) require the use of monitoring and performance measurement strategies, the proposed rule is consistent with the Commission's policy on the use of PRA for risk-informed decision-making and, more importantly, would maintain adequate protection of public health and safety.

Future Development of Draft Guidance for the Risk-Informed Alternative

South Texas Project Nuclear Operating Company (STPNOC) submitted a letter of intent to pilot a risk-informed approach for addressing GSI-191 (ADAMS Accession No. ML103481027) in December 2010. Subsequently, the NRC received a pilot submittal from STPNOC on January 31, 2013 (ADAMS Accession No. ML13043A013), supplemented on June 19, 2013 (ADAMS Accession No. ML131750250). In parallel with the NRC's review of the application, the NRC will develop draft guidance for the risk-informed alternative to address the effects of debris on long-term cooling. That draft guidance will be published for comment upon completion, which is currently anticipated for early- to mid-calendar year 2015. The NRC will then evaluate public comments received on the draft guidance, and develop the final guidance on a timeline that ensures all guidance (both for the risk-informed alternative and the new proposed embrittlement criteria) is available when the NRC staff provides the final § 50.46c rule to the Commission (currently scheduled for February 2016).

C. Corrective Actions and Reporting Requirements

1. Peak Cladding Temperature and Equivalent Cladding Reacted

The ANPR identified the third objective of the rulemaking as the revision of the LOCA reporting requirements. Specifically, the ANPR indicated that the NRC considered revising the reporting criteria by redefining what constitutes a significant change or error in such a manner as to make the reporting requirements dependent upon the margin between the acceptance criteria limits and the calculated values of the respective parameters (i.e., PCT or CP-ECR). After reviewing the public comments received, the NRC recognizes that the proposed reporting requirements specified in the ANPR were complex, and might, as a result, promote unnecessary burden or misinterpretation. As such, the reporting requirements of this proposed rule would not incorporate a dependence on margin between the acceptance criteria and calculated parameters.

The proposed rule would add a reporting requirement and definition of significant change or error based on predicted changes in maximum local oxidation (i.e., ECR), reformat the reporting section to clarify existing requirements, and add a reporting requirement based on periodic breakaway oxidation measurements. Any changes or errors that prolong the temperature transient may further challenge the ITT analytical limit; however, they may not significantly change the predicted PCT. As such, this change or error would not be captured in the reporting requirements. To improve the reporting and evaluation of changes or errors of this type, the NRC would expand the definition of significant change or error to include maximum local oxidation. The

threshold for a significant change or error, 0.4 percent ECR, would be equivalent to a change in calculated ECR for a 50 °F change in cladding temperature.

The definition of a significant change or error (i.e., 50 °F PCT, 0.4 percent ECR) is specific to zirconium-alloy cladding. A new definition of significant change or error may be necessary for other cladding materials. In addition, the proposed rule would require the use of maximum local oxidation (i.e., percent ECR) to evaluate the impact of a change or error on the predicted ITT.

Reporting requirements with respect to any “change to or error discovered in an NRC-approved ECCS evaluation model or in the application of such a model” have been a source of confusion. Two common misconceptions are: (1) Baseline values when estimating a significant change or error (i.e., greater than 50 °F), and (2) 30-day reporting including “a proposed schedule for providing a reanalysis.” When estimating a significant change or error, the proposed rule provides threshold values for both PCT and local oxidation. The baseline predictions used to assess a significant change or error should be the PCT and maximum local oxidation values documented in a plant's updated final safety analysis report (UFSAR). These values should represent the latest LOCA analyses that were submitted and reviewed by the NRC staff as part of a license amendment request (e.g., power uprate, fuel transition) as amended by prior annual reports. The following example illustrates the NRC's position:

In 2007, a licensee submits new LOCA analyses as part of an extended power uprate license amendment request with a predicted PCT of 1900 °F and maximum local oxidation (MLO) of 2.4 percent ECR. The 2008 and 2009 annual reports identify no changes or errors. In 2010, two errors in the ECCS evaluation model are discovered and documented in the annual report with an estimated impact on PCT of +25 °F and −20 °F and estimated impact on MLO of +0.08 percent ECR and −0.01 percent ECR. A 30-day notification was not required since the estimated impact was below the threshold for a significant change or error. At this point, the licensee should update the UFSAR, document the error notification, and identify the baseline for judging future changes or errors as 1905 °F PCT and 2.5 percent ECR.

When a change to or error in an ECCS evaluation model is discovered, the licensee would be responsible for estimating the magnitude of changes in predicted results to: (1) Determine if immediate steps are necessary to demonstrate compliance or bring plant design or operation into compliance with § 50.46c requirements, and (2) identify reporting requirements. Under the proposed rule, a licensee's obligation to report and take corrective action varies depending upon whether the licensee's situation falls into one of three possible scenarios, as described in this document:

1.

Change, error, or operation that does not result in any predicted response that exceeds any acceptance criteria and is itself not significant.

The licensee must:

a. Submit an annual report documenting the change(s), error(s), or operation along with the estimated magnitudes of changes in predicted results.

b. Revise the UFSAR.

c. Use the UFSAR PCT/ECR predictions as a baseline for future evaluations.

2.

Change, error, or operation that does not result in any predicted response that exceeds any acceptance criteria but is significant.

The licensee must:

a. Submit a 30-day report documenting the change(s), error(s), or operation, estimated magnitudes of changes in predicted results, and the schedule for providing a new analysis of record (AOR). The NRC will review the new AOR.

b. Revise the UFSAR to include new AOR.

c. Use the UFSAR PCT/ECR predictions as a baseline for the future evaluations.

3.

Change, error, or operation that results in any predicted response that exceeds acceptance criteria.

The licensee must:

a. Take immediate actions to bring the plant into compliance with acceptance criteria.

b. Report the change, error, or operation under §§ 50.55(e), 50.72, and 50.73, as applicable.

c. Submit a 30-day report documenting the change(s), error(s), or operation, estimated magnitudes of changes in predicted results, and the schedule for providing a new AOR. The NRC will review the new AOR.

d. Revise the UFSAR to include new AOR.

e. Use the UFSAR PCT/ECR predictions as the baselines for future evaluations.

The proposed reporting requirements in § 50.46c(m) reflect reformatting of the current reporting provisions in order to separately identify these three scenarios and clarify their respective requirements.

The proposed rule would also add the requirement to report results of breakaway oxidation measurements to the NRC. The licensees would be required to measure breakaway oxidation prior to each reload batch, and report the measurements within the calendar year following the testing. The breakaway oxidation phenomenon is explained in detail in sub-section B.3, “Breakaway Oxidation” of this section, “Proposed Requirements for ECCS Performance During LOCAs.” This reporting requirement would be specific to zirconium-alloy cladding and may not be applicable to other cladding materials.

2. Risk-Informed Alternative To Address Debris for Long-Term Cooling

Section 50.46c(e) of the proposed rule would require reasonable confidence that any calculated increase in CDF or LERF associated with debris is small. In the context of this paragraph, the calculated increases in CDF and LERF represent the difference between the as-built, as-operated plant (accounting for the effects of debris) and the “baseline” plant where the effects of debris are assumed to be negligible. This approach quantifies the portions of CDF and LERF attributable to debris and designates them as ΔCDF and ΔLERF. These metrics inform the NRC staff's decision on whether the effects of debris are acceptably small and consistent with the Commission's Safety Goal Policy Statement.

Subsequent changes to the plant or the PRA model may change the baseline CDF and LERF values as well as ΔCDF and ΔLERF. Because the NRC staff's original decision was based in part on these metrics, subsequent changes to their values should be assessed to ensure that the bases for this decision are still valid. It should be noted that the cumulative effects of operating changes (including plant modifications, procedural changes, and SSC performance) must be maintained within the rule's risk acceptance criteria over the life of the plant and, therefore, the evaluation of subsequent changes needs to address the cumulative effect of these changes.

Therefore, the proposed rule contains a corrective action and reporting requirement that would ensure that changes and errors are evaluated, reported to the NRC (as appropriate), and corrected in a timely manner (as appropriate). Consistent with the NRC's integrated approach to decisionmaking, changes that can impact risk, defense-in-depth, or safety margins need to be evaluated and, as appropriate, reported to the NRC. These terms, while frequently used, can have different definitions to different stakeholders. Therefore, the NRC intends to ensure that licensees using the risk-informed

approach to debris update their UFSAR to list applicable plant-specific capabilities of defense-in-depth and safety margins with respect to the proposed rule.

In addition, the NRC's approval under § 50.46c(e)(3) would specify the circumstances under which the entity would be required to notify the NRC of changes or errors in the risk evaluation approach used to address the effects of debris on long-term cooling. This requirement would ensure that if errors in the approach are identified subsequent to the NRC approval or if the entity seeks to change specific aspects of their approach that were determined by the NRC to be important to the NRC approval, such as the scope or level of detail of the PRA, these circumstances would be clearly identified in the NRC's approval. These requirements would ensure conditions that result in exceeding the § 50.46c(e) acceptance criteria are identified, corrected, and reported in a timely manner, and thus, ensure the effects of debris on long-term core cooling continue to be appropriately addressed.

The corrective action and reporting requirements for the aspects of the rule related to entities using the risk-informed alternative approach of § 50.46c(e) would be established in § 50.46c(m)(4). The proposed rule recognizes that there are different corrective and reporting requirements for different entities, as depicted in Table 1, Corrective Actions and Reporting: Risk-Informed Approach.

Table 1—Corrective Actions and Reporting: Risk-Informed Approach

Entity (and applicable proposed requirement)

Requirement to re-evaluate?

Requirement to report?

Requirement to make

necessary changes?

Design certification applicant before issuance of final design certification rule (covered by § 50.46c(m)(4)(i))

No (But known errors and discoveries must be corrected)

Yes (Submit amended application)

Yes (Changes in amended application).

Design certification applicant during the period of validity under § 52.55(a) and (b)—not currently referenced in any combined operating license (COL) application or COL (covered by § 50.46c(m)(4)(ii))

No

Yes (Only if referenced in a COL; then within 30 days)

No.

Design certification applicant during the period of validity under § 52.55(a) and (b)—once referenced in a COL application or COL (covered by § 50.46c(m)(4)(iii))

Yes

Yes

No.

Design certification renewal applicant (covered by § 50.46c(m)(4)(iv))

Yes

Yes (as part of renewal application)

Yes.

Combined license applicant (covered by § 50.46c(m)(4)(v))

No (But known errors and discoveries must be corrected)

Yes (Submit amended application)

Yes (Changes in amended application).

Combined license holder before finding under § 52.103(g) (covered by § 50.46c(m)(4)(vi))

No

Yes

Yes.

Operating license holder or combined license holder after finding under § 52.103(g) (covered by § 50.46c(m)(4)(vii))

Yes

Yes

Yes.

For design certification applicants (i.e., prior to issuance of the final design certification rule), the proposed rule would require that, if any errors are discovered, the applicant must submit a report to the NRC within an amended application. That amended application would describe any changes to the certified design and/or changes in the analyses, evaluations, and modeling (including the debris evaluation model and the PRA and its supporting analyses); and would demonstrate that the acceptance criteria in § 50.46c(e)(1) are met.

For design certification applicants during the period of validity under § 52.55(a) and (b) that are not currently referenced in any COL application or COL, there would be no evaluation, reporting, or change requirement. However, once the design certification is referenced by a COL applicant, any information regarding compliance with § 50.46c(e)(1) must be reported in accordance with the requirements in 10 CFR part 21.

For design certification applicants during the period of validity under § 52.55(a) and (b) that

are

referenced in a COL application or COL, the proposed rule would require the design certification applicant to evaluate and report any information concerning compliance with the acceptance criterion of § 50.46c(e)(1). However, there would be no requirement to make changes to the analyses, evaluations, and modeling until the time of renewal.

For design certification renewal applicants, the proposed rule would require the applicant to re-evaluate the analyses, evaluation, and modeling; report any changes or errors; and include in its application any necessary changes to the certified design, debris evaluation model, PRA, or supporting analyses to demonstrate that the renewed certified design meets the acceptance criteria in § 50.46c(e)(1).

For combined license applicants, the proposed rule would require the applicant to report any errors that are discovered within 30 days of the completion of that determination. The combined license applicants would be required to report the errors and make any necessary changes to the analyses, evaluation, or modeling within the amended application.

For combined licenses before the finding under § 52.103(g), the proposed rule would require that any errors that are discovered be updated in the analyses, evaluations, and modeling no later than the scheduled date for initial fuel loading under § 52.103(a). The licensee must also confirm that the acceptance criteria of § 50.46c(e)(1) continue to be met. Once this update is submitted, and until the Commission has made the finding under § 52.103(g), the licensee shall re-perform the review to ensure the acceptance criteria of § 50.46c(e)(1) continue to be met in a timely manner; this ensures that updating occurs if there are extended delays in the scheduled date for initial fuel loading. If the licensee determines that any acceptance criterion of § 50.46c(e)(1) are not met, then the licensee would be required to submit an application for amendment of its

combined license and departure from a referenced design certification rule, if applicable.

For operating licenses and combined licenses after the finding under § 52.103(g), the proposed rule would require that the licensee re-evaluate the analysis, evaluation, and modeling by no later than 48 months after the last review to confirm that the acceptance criteria of § 50.46c(e)(1) continue to be met. The licensee would also be required to take action in a timely manner to bring the licensee into compliance and report any failure to meet the acceptance criteria of § 50.46c(e)(1). Further, the amended application for the combined license would be required to include a request for exemption from a referenced design certification rule but would not need to address the criteria for obtaining an exemption.

D. Consideration of PRM-50-84: Thermal Effects of Crud and Oxide Layers

Determination of PRM

This proposed rule would address issues raised in a PRM that was submitted by Mark Leyse on March 15, 2007, and docketed as PRM-50-84. The petition requests that the NRC conduct rulemaking in three specific areas:

(1) Establish regulations that require licensees to operate light-water power reactors under conditions that are effective in limiting the thickness of crud and/or oxide layers on zirconium-clad fuel in order to ensure compliance with § 50.46(b) ECCS acceptance criteria;

(2) Amend appendix K to 10 CFR part 50 to explicitly require that the steady-state temperature distribution and stored energy in the reactor fuel at the onset of the postulated LOCA be calculated by factoring in the role that the thermal resistance of crud deposits and/or oxide layers plays in increasing the stored energy in the fuel. (These requirements also need to apply to any NRC-approved, best-estimate ECCS evaluation models used in lieu of appendix K to 10 CFR part 50 calculations); and

(3) Amend § 50.46 to specify a maximum allowable percentage of hydrogen content in [fuel rod] cladding.

On May 23, 2007 (72 FR 29802), the NRC published a notice of receipt for this petition in the

Federal Register

and requested public comment on the petition. The public comment period ended on August 6, 2007. After evaluating the public comments, the NRC decided that each of the petitioner's issues should be considered in the rulemaking process. On this basis, the NRC closed the docket on the petition for rulemaking. The NRC's determination, and evaluation of public comments received, was published in the

Federal Register

on November 25, 2008 (73 FR 71564).

Technical Issues in PRM-50-84

Licensees use approved fuel performance models to determine fuel conditions at the start of a LOCA, and the impact of crud and oxidation on fuel temperatures and pressures may be determined explicitly or implicitly by the system of models used. With the addition of an unambiguous regulatory requirement to address the accumulation of crud and oxide during plant operation, the NRC believes that fuel performance and LOCA evaluation models must include the thermal effects of both crud and oxidation whenever their accumulation would affect the calculated results. The NRC notes that licensees are required to operate their facilities within the boundary conditions of the calculated ECCS performance. During or immediately after plant operation, if actual crud layers on reactor fuel are implicitly determined or visually observed after shutdown to be greater than the levels predicted by or assumed in the ECCS evaluation model, licensees would be required to determine the effects of the increased crud on the calculated results. In many cases, engineering judgment or simple calculations could be used to evaluate the effects of increased crud levels; therefore, detailed LOCA reanalysis may not be required. In other cases, engineering judgment is used to determine that new analyses would be performed to determine the effect the new crud conditions have on the final calculated results. If unanticipated or unanalyzed levels of crud are discovered, then the licensee must determine if correct consideration of crud levels would result in a reportable condition as provided in the relevant reporting paragraphs. Should this proposed rule be adopted in final form, the NRC believes this regulatory approach to address crud and oxide accumulation during plant operation would satisfactorily address the issues raised by the petitioner's first request.

The formation of cladding crud and oxide layers is an expected condition at nuclear power plants. Although the thickness of these layers is usually limited, the amount of accumulated crud and oxidation varies from plant to plant and from one fuel cycle to another. Intended or inadvertent changes to plant operational practices may result in unanticipated levels of crud deposition. The NRC agrees with the petitioner (the petitioner's second request) that crud and/or oxide layers may directly increase the stored energy in reactor fuel by increasing the thermal resistance of cladding-to-coolant heat transfer, and may also indirectly increase the stored energy through an increase in the fuel rod internal pressure. As such, to ensure that licensee ECCS models properly account for the thermal effects of crud and/or oxide layers that have accumulated during operations at power, the proposed rule would add a requirement to evaluate the thermal effects of crud and oxide layers that may have accumulated on the fuel cladding during plant operation. If the NRC adopts the proposed rule in final form, then the second request of PRM-50-84 would be resolved.

The petitioner's third request is for the NRC to establish a maximum allowable percentage of hydrogen content in fuel rod cladding. The purpose of this request is to prevent embrittlement of fuel cladding during a LOCA. Although the NRC has decided not to propose the specific rule language recommended by the petitioner, the proposed new zirconium-specific requirements, if adopted in final form, would address the petitioner's third request by considering cladding hydrogen content in the development of analytical limits on integral time at temperature.

The NRC believes that this proposed rule addresses each of the three issues raised in PRM-50-84. If the NRC adopts the proposed rule in final form, PRM-50-84 would be granted in part and resolved.

E. Implementation

The proposed rule would specify the dates for compliance with the rule for existing operating license holders as well as holders of new reactor construction permits, combined licenses, and applicants for standard design certifications. The proposed rule sets forth a staggered schedule for compliance with the final rule, depending upon existing margin to the revised requirements with respect to embrittlement and the anticipated level of effort to demonstrate compliance. Apart from this staggered schedule for compliance, the rule also allows licensees the alternative of voluntarily seeking to meet the long-term cooling requirements of the proposed rule (and other changes as permitted by the risk-informed alternative and noted in the application) using a risk-informed approach, which could be accomplished in advance of the date for compliance

with the rule as set forth in the staggered schedule.

1. Staggered Implementation Schedule

For existing operating nuclear power reactors, the proposed rule includes a staged schedule for implementation. The NRC has developed this staged implementation to improve the efficiency and effectiveness of this migration toward the new ECCS requirements for the existing operating fleet. As part of this plan, licensees have been divided among three implementation tracks based upon existing margin to the revised requirements and anticipated level of effort to demonstrate compliance. The purpose of the staged implementation approach is to bring licensees into compliance as quickly as possible, while accounting for: (1) Differences between realistic and appendix K to 10 CFR part 50 LOCA models; and (2) the level of effort and scope of analyses required for compliance. Table 2 provides an overview of the implementation schedule for the existing fleet. Note that the compliance schedule requirement represents the date that the licensee submits either the letter report or license amendment request (as opposed to the date of NRC approval). The proposed track assignments for every operating reactor is provided in Table 1 of proposed § 50.46c(o). Table 1 of proposed § 50.46c(o) would be updated, as necessary, to capture the implementation track assignments for all operating reactors at the time the final rule is issued. Applications for a 10 CFR part 50 operating license under review on the effective date of the rule would be assigned an implementation track based on the factors used in establishing the three tracks (as described in Table 1). An applicant for a new 10 CFR part 50 operating license submitted or docketed after the effective date of the rule must comply with the provisions of the rule. The NRC notes that Vermont Yankee Nuclear Power Station is listed in the implementation track assignments. Although Vermont Yankee submitted a notification of permanent cessation of power operations under § 50.82(a)(1)(i) (see ADAMS Accession No. ML13273A204), that notification contained only an estimate of the date of cessation. Vermont Yankee plans to supplement that letter with a (firm) date of cessation, as required per §§ 50.82(a)(1)(i) and 50.4(b)(8). Watts Bar, Unit 2, and Bellefonte, Units 1 and 2, have construction permits in effect or in the process of being reinstated. However, the ECCS margin to the proposed rule's requirements on embrittlement for each of these plants is not yet known. (A final safety analysis report (FSAR) has not been approved for these plants.) The NRC will determine the appropriate track for each plant once its ECCS margin to embrittlement is finalized. At that point, that plant would be added to Table 1 of proposed § 50.46c(o) in the appropriate track, and the title of Table 1 would be modified accordingly.

Table 2—Implementation Plan

Implementation track

Basis

Anticipated level of effort

Number of units

BWR

PWR

Compliance demonstration

1

All plants which satisfy new requirements without new analyses or model revisions

Low

27

37

No later than 24 months from effective date of rule.

2

PWR plants using realistic large-break (LB) LOCA models requiring new analyses. BWR/2 plants

Medium

2

13

No later than 48 months from effective date of rule.

3

PWR plants using appendix K LB and small-break (SB) models requiring new analyses. BWR/3 plants

Medium-High

6

15

No later than 60 months from effective date of rule.

To support the implementation of the proposed requirements on individual plant dockets, fuel vendors would be encouraged to submit for NRC review alloy-specific hydrogen uptake models and any LOCA model updates (e.g., incorporation of CP weight gain correlation) no later than 12 months from the effective date of the final rule. Upon approval, these models and methods could be used to demonstrate the ECCS performance against the new analytical limits. For Track 1 plants that would not require new ECCS evaluations, licensees should complete any necessary engineering calculations, update their plant UFSAR, and provide a letter report to the NRC documenting compliance with § 50.46c. The NRC recognizes that to demonstrate compliance, these plants would need to utilize newly-approved hydrogen uptake models and integrate time at temperature using the CP weight gain correlation (for appendix K to 10 CFR part 50 models).

For any unit at a plant that would require a new ECCS evaluation, including adopting a previously approved realistic evaluation model, revising an existing evaluation model, performing a new LOCA break spectrum analysis, performing a multiple rod survey (e.g., burnup-rod power tradeoff), or making changes to a technical specification or core operating limit report (COLR), licensees would need to submit the new LOCA AOR and, where applicable, a license amendment request updating the COLR list of approved methods.

The NRC has developed a phased implementation approach for applicants and holders of standard design approvals, design certifications, combined licenses, and manufacturing licenses granted under 10 CFR part 52.

The proposed implementation plan for reactors approved under 10 CFR part 52 would allow the applicant for a design certification, standard design approval, or manufacturing license either submitted to, or docketed by, the NRC prior to the effective date of the rule, to come into compliance with the rule at the time of any application for renewal.

An applicant for a design certification, standard design approval, or manufacturing license submitted or docketed after the effective date of the rule must comply with the provisions of the rule.

The holder of a combined license granted prior to the effective date of the rule would be permitted to operate the plant for one fuel cycle before demonstrating compliance with the rule. Doing so would permit adequate time to submit demonstration of compliance with the rule prior to

achieving fuel burnup for which the cladding limitations are imposed by the rule. In this case the holder of the combined license would be required to remain in compliance with the ECCS performance acceptance criteria in place at the time the combined license was granted.

Applicants for combined licenses docketed after the effective date of the rule must comply with the provisions of the rule.

The proposed rule reflects the NRC's determination that reactor designs reviewed and approved under 10 CFR part 52 should have the same constraints as the reactors operating under 10 CFR part 50 with respect to development, submittal, and approval of ECCS performance models necessary to demonstrate compliance with this rule. Alloy-specific hydrogen uptake models and all ECCS performance model updates would be expected to be submitted in a timely manner for NRC review and approval so that demonstration of the ECCS performance with respect to the analytical limits would not impact plant operation more than is necessary.

The proposed rule also reflects the NRC's expectation that, for new reactors licensed to operate prior to the effective date of the rule, operation for at least the initial fuel cycle using fuel that has not been analyzed under the proposed rule's provisions accounting for burn-up effects does not present an adequate protection concern. During the initial fuel cycle, the NRC believes that burn-up effects would not be limiting, and the current ECCS rule's acceptance criteria are sufficient during the initial fuel cycle to provide reasonable assurance of adequate protection with respect to overall ECCS performance.

2. Compliance With Long-Term Cooling Requirements Using Risk-Informed Approach To Address Debris Effects

Implementation of the alternative approach to addressing the impact of debris on long-term cooling is independent from implementation of the requirements related to the embrittlement research findings. The NRC would allow partial early implementation of the proposed requirements of § 50.46c, limited to this alternative approach. In other words, an applicant may elect to submit its risk-informed alternative under § 50.46c(e) prior to demonstrating compliance with the other requirements of § 50.46c. In this case, the licensee would have to receive NRC approval on both its risk-informed submittal and the analytical limit for long-term cooling required under § 50.46c(g)(1)(v) prior to using the risk-informed approach. The NRC is proposing to allow early implementation because the NRC encourages licensees to complete resolution of GSI-191 and this risk-informed alternative is one way of resolving the issue.

The NRC has determined that a licensee's decision to use a risk-informed methodology to evaluate the effects of debris on ECCS and CSS with respect to long-term cooling following a LOCA should be reviewed and approved by the NRC prior to implementation. The ECCS and CSS are significant safety systems that provide necessary defense-in-depth. The design bases for the ECCS are of high regulatory significance to the NRC, as reflected in the detailed requirements applicable to the ECCS (and the associated fuel system) in § 50.46 and appendix K to 10 CFR part 50. In addition, the design bases for the ECCS and the CSS affect the design bases for many other SSCs throughout the nuclear power plant. Therefore, changes to the design assumptions for the ECCS and CSS may have significant effects on the design bases for other SSCs throughout the plant. These potential effects include changes in the consequences of postulated accidents, margins of safety, and defense-in-depth.

The NRC also determined that § 50.59, properly implemented, would not allow a change to the design bases of a plant to use a risk-informed methodology for evaluating the effects of debris on long-term cooling. A risk-informed methodology for addressing the effects of debris on long-term cooling is a departure from the method of evaluation described in the current UFSAR, as updated and used in establishing the design bases in the safety analysis as defined in § 50.59(a)(2). Hence, under § 50.59(c)(2)(viii), a licensee's departure from the existing methodology for evaluating long-term cooling must be reviewed and approved by the NRC as a license amendment.

In sum, given the importance of the ECCS and CSS, the “cascading” effects of changes in ECCS and CSS design on the design bases of other SSCs of a nuclear power plant, the NRC believes that a licensee's decision to use a risk-informed methodology to evaluate the effects of debris on ECCS with respect to long-term cooling should be reviewed and approved by the NRC. Under the proposed rule, the NRC's review and approval is accomplished through the license amendment process in accordance with §§ 50.90 through 50.92.

VI. Section-by-Section Analysis

The organization and 10 CFR designations of the NRC's requirements governing emergency core cooling (currently in § 50.46) and reactor cooling venting systems (currently in § 50.46a) are expected to change. These changes would result from:

(1) The current schedule for Commission serial adoption of two rulemakings: (i) The finalization of the proposed rule on risk-informed changes to ECCS systems, currently referred to as the § 50.46a rulemaking, followed by; (ii) the finalization of this proposed rule on performance-based changes to ECCS requirements and cladding acceptance criteria, currently referred to as the § 50.46c rulemaking;

(2) The proposed schedule for implementation of these rules; and

(3) The need to maintain current requirements in place for those reactors that have not transitioned to the new requirements under the implementation schedule to be specified in the final rule.

The following table shows how the organization and 10 CFR designation of these rules will evolve, if the NRC sequentially adopts the two final rules and licensees complete implementation of the alternate cladding requirements. The NRC notes that, in an SRM, “SRM-SECY-10-0161—`Final Rule: Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements (10 CFR 50.46a)',” dated April 26, 2012 (ADAMS Accession No. ML12117A121), the Commission approved the NRC staff's request to withdraw SECY-10-0161, “Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements (10 CFR 50.46a),” from Commission consideration (ADAMS Accession No. ML121500380). The NRC does not plan to publish a notice in the

Federal Register

withdrawing the § 50.46a proposed rule. The NRC staff plans to resubmit the draft final rule for Commission consideration in conjunction with the Near-Term Task Force (NTTF) Recommendation 1 activities. (For information on NTFF Recommendation 1, see “Recommendations for Enhancing Reactor Safety in the 21st Century,” dated July 12, 2011, ADAMS Accession No. ML 112510271.) Therefore, the § 50.46a rulemaking still may be finalized before the § 50.46c rulemaking, as assumed in the following table.

Existing NRC requirements and proposed new regulations (bolded rules are currently in effect)

Rulemaking and implementation activities

Adoption of final risk-informed ECCS requirements (§ 50.46a)

Initial codification of final performance-based fuel cladding requirements

End of phased implementation period for performance-based cladding requirements

§ 50.46 ECCS Acceptance Criteria

§ 50.46 ECCS Acceptance Criteria (

unchanged

)

§ 50.46 ECCS Acceptance Criteria (

unchanged

)

§ 50.46 ECCS Acceptance Criteria (

see discussion for § 50.46c under this column

).

Risk-Informed ECCS Requirements

(currently designated in final rulemaking package as § 50.46a)

§ 50.46a Risk-Informed ECCS Requirements

§ 50.46a Risk-Informed ECCS Requirements

§ 50.46a Risk-Informed ECCS Requirements.

§ 50.46a Reactor Coolant Venting Systems

Redesignated as § 50.46b

NA (

Redesignation as § 50.46b completed

)

NA (

Redesignation as § 50.46b completed

).

Performance-based ECCS and Cladding Requirements

(currently designated in draft proposed rulemaking package as § 50.46c)

NA

§ 50.46c Alternate Fuel Cladding Requirements

NA

(Administrative rulemaking would: (i) remove superseded fuel cladding requirements in § 50.46, and (ii) redesignate § 50.46c as § 50.46.).

A. Section 50.46c—Heading

A new section, § 50.46c, would be created in 10 CFR part 50 by this rulemaking. The heading of § 50.46c would be “Emergency core cooling system performance during loss-of-coolant accidents.”

B. Section 50.46c(a)—Applicability

Paragraph (a) would define the applicability of the proposed rule, which remains limited to LWRs, but would be expanded beyond fuel designs consisting of uranium oxide pellets within cylindrical zircaloy or ZIRLO

TM

cladding. The proposed rule would also be applicable to applicants for and holders of construction permits, operating licenses, combined licenses, and standard design approvals, and also to applicants for standard design certifications and for manufacturing licenses.

C. Section 50.46c(b)—Definitions

Paragraph (b) would provide definitions for terms used in this section. The definitions of

Loss-of-coolant accident

and

Evaluation model

would remain unchanged from those currently located in § 50.46(c)(1) and (c)(2), respectively.

The definition of

Breakaway oxidation

and

Debris evaluation model

would be added.

D. Section 50.46c(c)—Relationship to Other NRC Regulations

Paragraph (c) would describe the relationship of § 50.46c to other NRC regulations. The description in proposed paragraph (c) would remain largely unchanged from that of the current regulation found in § 50.46(d). However, the description would be revised to make clear that an approach approved by the NRC under § 50.46c(e) may also be used when evaluating the effects of debris to demonstrate compliance with other requirements of this part, including GDC-35, GDC-38, and GDC-41 (as allowed by § 50.46c and requested in the application).

E. Section 50.46c(d)—Emergency Core Cooling System Design

Paragraph (d)(1) would define performance-based requirements for the ECCS. Paragraph (d)(2) would require that ECCS performance be demonstrated using an NRC-approved ECCS evaluation model meeting specific requirements for a range of postulated LOCAs of different sizes, locations, and other properties, sufficient to provide assurance that the most severe postulated LOCA has been identified. The provisions for a realistic ECCS model or appendix K to 10 CFR part 50 model would remain unchanged from the current regulation found in § 50.46(a)(1)(i) and (ii), respectively. Similarly, the model requirement that calculated changes in core geometry must be addressed would remain unchanged from the current regulation found in § 50.46(b)(4). Paragraph (d)(2)(iii) would explicitly require that the ECCS evaluation model address calculated changes in core geometry, and consider factors that may alter localized coolant flow or inhibit delivery of coolant to the core. Demonstration of ECCS performance in the post-accident recovery period, or long-term cooling, is expected to consider inhibition of core flow that can result from such factors as, but not limited to, pump damage, piping damage, boron precipitation, and deposition of debris and/or chemicals associated with the long-term cooling mode of recirculation coolant collection from the reactor building sump. Consideration of debris and/or chemical deposition is already required by the current rule, and the proposed rule does not alter the current efforts to address such factors under programs such as GSI-191. Demonstration of consideration of such factors may also be achieved through analytical models that adequately represent the empirical data obtained regarding debris deposition. The proposed rule would alternatively allow the use of risk-informed approaches to evaluate the effects of debris on localized coolant flow and delivery of coolant to the core during the long-term cooling (post-accident recovery) period.

In addition, paragraph (d)(2)(iv) of the proposed rule would specifically require that ECCS performance be demonstrated for both the accident and the post-accident recovery and recirculation period.

Paragraph (d)(2)(v) would require that the ECCS model address the fuel system modeling requirements in paragraph (g)(2) if the reactor uses uranium oxide or mixed uranium-plutonium oxide pellets within zirconium cladding (e.g., currently operating reactors).

Paragraph (d)(3) would provide the ECCS evaluation model documentation requirements currently provided in appendix K, Section II, “Required Documentation.”

F. Section 50.46c(e)—Alternate Risk-Informed Approach for Addressing the Effects of Debris on Long-Term Core Cooling

Paragraphs (d)(2)(iii) and (e) would allow entities to use a risk-informed approach for addressing the effects of debris on long-term core cooling. Paragraphs (e)(1)(i) through (e)(1)(iv) would provide the acceptance criteria for an acceptable alternative risk-informed approach for addressing the effects of debris on long-term core cooling and would establish minimum requirements for the plant PRA and how it is to be used in the alternate risk-informed approach. These proposed requirements are intended to ensure that the implementation of the alternate risk-informed approach to address debris

effects on long-term core cooling would provide reasonable confidence that any resulting increase in CDF and LERF will be small, and that sufficient defense-in-depth and safety margins are maintained. These proposed requirements are consistent with the key principles of risk-informed decisionmaking described in RG 1.174, Revision 2.

Paragraph (e)(1)(i) of the proposed rule would require that there be reasonable confidence that any potential risk increase be small. Paragraph (e)(1)(ii) would require that sufficient defense-in-depth and safety margins be maintained as part of the implementation of the alternate risk-informed approach. Further, paragraphs (e)(1)(iii) and (iv) would contain the minimum requirements for the plant PRA and how it is to be used in the alternate risk-informed approach.

Paragraph (e)(2) would require those applicants seeking to use the alternative risk-informed approach under paragraph (e)(1) to submit an application that contains the information provided in paragraphs (e)(2)(i) through (e)(2)(v).

Paragraph (e)(2)(i) would require applicants to follow established regulatory guidance that the NRC expects to finalize concurrent with the final rule. If an applicant wishes to use a different approach, the submittal must provide a sufficient description of how the alternative risk-informed approach would be conducted and why it is acceptable.

Paragraph (e)(2)(ii) would require that initiating events from sources both internal and external to the plant and for all modes of operation, including low power and shutdown modes, be considered when evaluating the effects of debris on long-term core cooling using the alternate approach. This aspect of the rule recognizes that the minimum PRA that would be required by paragraph (e)(1)(iv) may not address all sources of initiating events and modes of operations, and as such, other approaches may be used. Therefore, the application would need to describe the measures taken to assure the scope, level of detail, and technical adequacy of all the analyses performed to address severe accidents are sufficient for this application and address the full spectrum of initiating events and modes of operation.

Paragraph (e)(2)(iii) would specifically address the need to provide the results of the PRA review process. This aspect includes such items as any peer reviews performed, any actions taken to address peer review findings that are important to the application, and any efforts to compare the plant-specific PRA to the ASME/ANS PRA standard, as endorsed by the NRC in RG 1.200.

In paragraph (e)(2)(iv), the applicant would be required to include information about the evaluations they conduct to provide reasonable confidence that any potential increase in risk would be small. The applicant would be required to provide sufficient information to the NRC, describing the evaluations and the basis for their acceptability as appropriately representing the potential increase in risk from implementation of the requirements in this rule.

In paragraph (e)(2)(v), the applicant would be required to provide a description of the analytical limit on long-term peak cladding temperature established in accordance with paragraph (g)(1)(v).

Paragraph (e)(3) would provide that the NRC may approve an application to implement the alternative risk-informed approach if it determines that the proposed approach satisfies the requirements of paragraph (e)(1) and establishes an acceptable long-term peak cladding temperature limit. The NRC staff would review the description of the alternative risk-informed approach set forth in the application, and the associated evaluations, to confirm that it contains the elements required by the rule. The NRC staff would also review the information provided about the plant-specific PRA and other systematic evaluations used to evaluate severe accidents in support of the application to assure that the scope, level of detail, and technical adequacy of the analyses are commensurate with the reliance on the risk information. This aspect of the review would involve the NRC assessment of the information provided about: 1) the peer review process to which the plant-specific PRA was subjected, 2) the reliance on other systematic evaluations to address areas not covered by the plant-specific PRA, and 3) the approach for maintaining sufficient defense-in-depth and safety margins. The NRC staff intends to use review guidance for this purpose. The NRC's approval of the use of the risk-informed approach to address long-term cooling would specify the circumstances under which the entity would be required to notify the NRC of changes or errors in the risk evaluation approach used to address the effects of debris on long-term cooling. Depending upon the nature of the underlying application (e.g., license, design certification rule, or design approval), the approval and notification requirement will be implemented through a license condition, a provision in the design certification rule, or a condition of the design approval, as applicable.

Paragraph (f) would be added to reserve rulemaking space for future amendments to § 50.46c.

G. Section 50.46c(g)—Fuel System Designs: Uranium Oxide or Mixed Uranium-Plutonium Oxide Pellets Within Cylindrical Zirconium-Alloy Cladding

This section would be added to set forth fuel design specific analytical limits and performance-based requirements by which to judge the overall ECCS performance in accordance with paragraph (d)(1) for LWRs using uranium oxide or mixed uranium-plutonium oxide pellets within cylindrical zirconium alloy cladding. The fuel performance criteria in paragraph (g)(1) and fuel system modeling requirements in paragraph (g)(2) are based on the established degradation mechanisms and performance objectives for this specific fuel type.

Paragraph (g)(1)(i) would establish an analytical limit on peak cladding temperature to avoid cladding embrittlement, high temperature failure modes, and run-away exothermic oxidation. Except as calculated in paragraph (g)(1)(ii), the calculated maximum fuel element cladding temperature should not exceed 2200 °F. This requirement remains unchanged from the current requirement at § 50.46(b)(1).

Paragraph (g)(1)(ii) would require that the zirconium alloy cladding maintains sufficient post-quench ductility in order to avoid gross failure. This requirement replaces the current prescriptive analytical limit, 17 percent ECR, in § 50.46(b)(2).

Paragraph (g)(1)(iii) would be added to establish a performance-based requirement to preclude breakaway oxidation in order to avoid cladding embrittlement and gross failure. Breakaway oxidation is a new requirement relative to § 50.46(b).

Paragraph (g)(1)(iv) would establish an analytical limit on maximum hydrogen generation to avoid an explosive concentration of hydrogen gas. This requirement would be the same as that of the current regulation in § 50.46(b)(3).

Paragraph (g)(1)(v) would be added to establish a performance-based requirement to ensure acceptable fuel performance during long-term cooling. This performance requirement is consistent with the current requirement to “maintain the calculated core

temperature at an acceptably low value” located in § 50.46(b)(5).

Paragraph (g)(2) would establish fuel design specific modeling requirements that are needed in addition to the generic ECCS evaluation model requirements in paragraph (d)(2). Paragraph (g)(2)(i) would require consideration of oxygen diffusion from the cladding inside surface. This would be a new ECCS evaluation model requirement.

Paragraph (g)(2)(ii) would be added to include a requirement to evaluate the thermal effects of crud and oxide layers that may have accumulated on the fuel cladding during plant operation.

Paragraphs (h) through (j) would be added to reserve rulemaking space for future amendments to § 50.46c, including any changes that stem from using newly designed fuel and cladding materials.

H. Section 50.46c(k)—Use of NRC-Approved Fuel in Reactor

Paragraph (k) would prohibit licensees from loading fuel into a reactor, or operating the reactor, unless the licensee either determines that the fuel meets the requirements in paragraph (d), or complies with technical specifications governing lead test assemblies in its license.

I. Section 50.46c(l)—Authority To Impose Restrictions on Operation

Paragraph (l) would provide that the Director of the Office of Nuclear Reactor Regulation or the Director of the Office of New Reactors may impose restrictions on reactor operation if it is found that the evaluations of ECCS cooling performance submitted are not consistent with the requirements of this section. The authority to impose restrictions would be expanded, relative to the authority currently granted in § 50.46(a)(2), to address licenses issued under 10 CFR part 52.

J. Section 50.46c(m)—Corrective Actions and Reporting

Paragraph (m) would provide reporting requirements applicable to the ECCS evaluation model and reporting requirements applicable to entities that elect to use the risk-informed alternative to address the effects of debris on long-term cooling. Paragraphs (m)(1) through (m)(3) would apply to all entities subject to § 50.46c; paragraphs (m)(4) would apply to those entities demonstrating acceptable long-term core cooling under the provisions of paragraph (e).

Paragraph (m)(1) would establish required action and reporting requirements if an entity identifies any change to, or error in, an ECCS evaluation model or the application of such a model, or any operation inconsistent with the evaluation model. For clarity, this paragraph was divided into three categories of changes or errors, each with its own proposed actions and reporting. These requirements are unchanged from the current § 50.46(a)(3), with the exception of conforming to analytical limits established in the proposed rule.

Paragraph (m)(1)(i) would establish required action and reporting requirements if an entity identifies any change to, or error in, an ECCS evaluation model or the application of such a model, or any operation inconsistent with the evaluation model, that does not result in any predicted response that exceeds any acceptance criteria and is itself not significant.

Paragraph (m)(1)(ii) would establish required action and reporting requirements if a licensee identifies any change to, or error in, an ECCS evaluation model or the application of such a model, or any operation inconsistent with the evaluation model, that does not result in any predicted response that exceeds any acceptance criteria but is significant (as defined in paragraph (m)(2)).

Paragraph (m)(1)(iii) would establish required action and reporting requirements for an entity who identifies any change to, or error in, an ECCS evaluation model.

Paragraph (m)(1)(iv) would require an amendment to a design certification application reflecting any reanalysis required by paragraph (m)(1)(ii) to be submitted by the applicant in concert with the reanalysis.

Paragraph (m)(2) would be added to provide the definition of a significant change or error. The definition would be expanded, relative to the 50 °F change in calculated peak cladding temperature in § 50.46(a)(3)(i), to include a 0.4 percent ECR change in calculated cladding oxidation.

Paragraph (m)(3) would require the onset of breakaway oxidation to be measured for each reload batch, and would require any changes in the time to the onset of breakaway oxidation to be assessed against the integral time and to be reported annually. This would be a new reporting requirement.

Paragraph (m)(4) would establish required action and reporting requirements for entities choosing to implement the alternative risk-informed approach for addressing the effects of debris on long-term core cooling. Paragraph (m)(4) would specify the evaluation, reporting, and change requirements for the various categories of entities that may elect to use the risk-informed approach.

Paragraph (n) would be added to reserve rulemaking space for future amendments to § 50.46c.

K. Section 50.46(o)—Implementation

This section would establish the implementation requirements and schedule for the existing fleet and for new reactors. Paragraph (o)(1) would require construction permits under 10 CFR part 50 issued after the effective date of the rule to comply with the requirements of § 50.46c.

Paragraph (o)(2) would require operating licenses under 10 CFR part 50 based upon construction permits (including deferred and reinstated construction permits) to comply with the requirements of § 50.46c by no later than the time frame established for operating reactors in the implementation table. Until that point, the construction permits identified by this paragraph must comply with § 50.46.

Paragraph (o)(3) would require operating licenses under 10 CFR part 50 issued after the effective date of the rule to comply with the requirements of § 50.46c.

Paragraph (o)(4) would require operating licenses under 10 CFR part 50 (as of the effective date of the rule) to comply with the requirements of § 50.46c by no later than the applicable date set forth in the implementation table for operating reactors.

Paragraph (o)(5) would require standard design certifications, standard design approvals, and manufacturing licenses under 10 CFR part 52, whose applications (including applications for amendment) are docketed after the effective date of the rule (including branches of these certifications whose applications are docketed after the effective date of the rule), to comply with the provisions of the rule. Applicants submitting after the rule has been adopted should have had ample time to develop and receive approval for the analysis methods necessary to comply with the provisions of the rule.

Paragraph (o)(6) would require standard design certifications under 10 CFR part 52 issued before the effective date of the rule to comply no later than the time of renewal of certification. Similar to the requirements of paragraph (o)(5), such applicants will have had ample time necessary to comply with the provisions of the rule.

Paragraph (o)(7) would require standard design certifications, standard design approvals, and manufacturing licenses, along with new branches of certifications under 10 CFR part 52

whose applications are pending as of the effective date of the rule to comply with § 50.46c no later than the time of renewal. Those entities that are in the approval process at the time the rule becomes effective will be required to comply at time of renewal. This will provide ample time to develop and receive approval for the methodologies necessary to comply with the rule. Paragraph (o)(8) would require combined license applications under 10 CFR part 52 that are docketed after the effective date of the rule to comply with the provisions of the rule.

Paragraph (o)(9) would require applications for combined licenses under 10 CFR part 52 that are docketed or issued after the effective date of the rule to comply with § 50.46c no later than completion of the first refueling outage after the initial fuel load. Those entities that are issued combined licenses prior to the effective date of the rule must comply with the rule no later than the first refueling outage after initial fuel load. This affords those entities ample time to develop and submit the necessary methodologies.

Entities that elect to use the voluntary alternative to the long-term cooling requirements of the proposed rule using a risk-informed approach can do so in advance of the date for compliance with the rule. In this case, the entity would have to receive NRC approval on both its risk-informed submittal and the analytical limit for long-term cooling required under § 50.46c(g)(1)(v) prior to using the risk-informed approach.

L. Appendix K to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR) ECCS Evaluation Models

In appendix K, a new paragraph II.6 would be added to clarify that, for those entities that have implemented § 50.46c, the requirements for documentation are located within § 50.46c(d)(3).

M. Redesignation of Venting Requirements in § 50.46a

This proposed rule would redesignate the current § 50.46a, “Acceptance criteria for reactor coolant system venting systems,” as proposed § 50.46b. A new § 50.46a would be added and reserved for future use as the rulemaking to provide a risk-informed alternative to the LOCA technical requirements.

N. Changes Throughout 10 CFR Parts 50 and 52

Several administrative changes would be made throughout 10 CFR parts 50 and 52 in order to conform with the proposed rule and proposed redesignation of the venting requirements in current § 50.46a. Section 50.8 would be amended to add the proposed rule to the list of approved information collections. Where §§ 50.34(a)(4), 50.34(b)(4), 52.47(a)(4), 52.79(a)(5), 52.137(a)(4), and 52.157(f)(1) refer to § 50.46, the proposed rule would add “and § 50.46c, as applicable.” Where §§ 50.34(a)(4), 52.47(a)(4), 52.79(a)(5), 52.137(a)(4), and 52.157(f)(1) refer to § 50.46a, the proposed rule would instead refer to § 50.46b.

Changes are also made to GDC-35, GDC-38, and GDC-41 in appendix A to 10 CFR part 50 to promulgate the acceptability of using a risk-informed alternative for long-term cooling when demonstrating compliance with these regulations, as allowed by § 50.46c and requested in the application.

VII. Specific Request for Comments on the Proposed Rule

In addition to the request for general comments on the proposed rule, the NRC also requests specific comments on the following topics:

A. Fuel Performance Criteria

NRC Question 1.

Performance-Based Peak Cladding Temperature Limit.

The NRC is proposing, in § 50.46c(g)(1)(i), to maintain the existing prescriptive criterion on PCT for zirconium alloy cladding. Limits on cladding temperature are necessary to protect against a loss of coolable geometry resulting from brittle failure upon quench, to protect against high-temperature ductile failure, and to prevent reaching the point at which the zirconium-water reaction would become autocatalytic. In the original § 50.46 rulemaking, the 2200 °F limit on PCT was based on cladding embrittlement (i.e., protection against brittle failure upon quench), which was determined to be more limiting than either high temperature ductile failure or autocatalytic oxidation. The NRC's LOCA research program did not investigate cladding degradation mechanisms or develop the technical basis for performance-based requirements beyond the existing 2200 °F PCT criterion. Since the cladding embrittlement mechanism, oxygen diffusion, is strongly dependent on temperature, there exists an upper temperature at which the allowable time duration to nil ductility approaches zero (i.e., PCT °limit). As described in Section V.B.1 of this document, recent research has confirmed that 2200 °F remains an appropriate upper limit to protect against cladding embrittlement since nil ductility is achieved rapidly at higher temperature. As such, the proposed § 50.46c maintains the 2200 °F prescriptive PCT criterion.

The NRC requests comment on the proposed rule's retention of the prescriptive PCT criterion, specifically:

a. In place of the prescriptive PCT criterion, should the NRC adopt performance-based requirements for zirconium alloy cladding to protect against high temperature ductile failure and autocatalytic oxidation?

b. Do established testing procedures already exist for demonstrating acceptable high temperature cladding performance and defining acceptance criteria to meet these new performance-based requirements?

NRC Question 2.

Periodic Breakaway Testing.

To address the breakaway oxidation phenomenon, the NRC proposes to add a performance-based requirement in § 50.46c(m)(3) that the licensee measure the onset of breakaway oxidation periodically on manufactured cladding material and report any changes in the onset of breakaway oxidation at least annually. This requirement, along with a periodic test requirement (defined as each reload batch in the proposed rule language), would confirm that slight composition changes or manufacturing changes have not inadvertently altered the cladding's susceptibility to breakaway oxidation. The NRC is considering adopting, as a final rule, a requirement that each licensee measure breakaway oxidation behavior for each re-load batch. The NRC requests specific comment on the type of data reported and the proposed frequency of required testing. The objective of periodic testing is to prevent affected fuel from being loaded into a reactor. At the same time, the objective is to do so without adding ineffective requirements and unnecessary burden. Other sampling approaches may be more effective. For example, should the licensee be required to report data relevant solely to their reload fuel batch or should the licensee be able to report representative data based on periodic testing (e.g., test every 10,000 rods, tubing lot, or ingot) of the same zirconium-based alloy cladding compiled during the period from the last report?

NRC Question 3.

Analytical Long-Term Peak Cladding Temperature Limit.

Section 50.46c(g)(1)(v) of the proposed rule would require that a specified and NRC-approved limit on long-term peak cladding temperature be established which preserves a measure of cladding ductility throughout the period of long-term demonstration (e.g., 30 days). The current regulation at § 50.46(b)(5) stipulates that long-term temperature be maintained “at an acceptably low

value.” The proposed rule would define the performance-based metric to judge an acceptably low temperature. The overall goal of preserving ductility would provide reasonable assurance that the fuel rods will maintain their coolable bundle array. The NRC is requesting input regarding this performance objective to determine if this is the most suitable performance-based metric to demonstrate long-term cladding performance.

Alternatively, the proposed rule could establish an analytical limit of long-term fuel rod cladding temperature related to observed corrosion behavior. For example, the Pressurized Water Reactor Owners Group (PWROG) has applied as a long-term core cooling acceptance criterion that the cladding temperature be maintained below 800 °F (see Topical Report (TR) Westinghouse Commercial Atomic Power (WCAP)-16793-NP, Revision 2, “Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid,” Appendix A (ADAMS Accession No. ML11292A021)). Doing so will ensure that additional corrosion and hydrogen pickup over a 30-day period will not significantly affect cladding properties. The NRC seeks comment on the acceptance criterion for long-term cooling and whether there is justification for a different temperature limit (other than the 800 °F provided in the WCAP).

B. Risk-Informed Alternative To Address the Effects of Debris

NRC Question 4.

Acceptance Criteria for Risk-Informed Alternative.

Section 50.46c(e) of the proposed rule contains the high-level acceptance criteria for an alternative that would allow entities to use, on a case-by-case basis, a risk-informed approach to address the effects of debris on long-term core cooling. In addition, the NRC will develop draft regulatory guidance for this provision concurrent with the staff's review of the STPNOC's pilot application for a risk-informed approach to address the closely related topic of GSI-191. The NRC seeks comment on whether the detailed acceptance criteria should be set forth in § 50.46c, or in the associated regulatory guidance.

NRC Question 5.

Regulatory Approach for Risk-Informed Regulation.

The NRC seeks comment on whether the risk-informed alternative offered by this regulation should require meeting numeric-risk acceptance criteria as a matter of compliance (similar to § 50.48c) or whether other risk-informed approaches that use risk-importance insights to establish measurable criteria or performance objectives, such as those in use by §§ 50.62, 50.63, and 50.65, or approaches using both risk importance and numeric-risk acceptance criteria, such as those in use by § 50.69, would be preferable.

NRC Question 6.

Operational Modes Considered in Risk-Informed Alternative.

Deterministic evaluations of GSI-191 are currently required only for those modes of operation where both recirculation from the sump is relied upon and the plant accident can cause high pressure jets that can result in generation and transport of debris to the sump. By contrast, probabilistic evaluations generally consider all modes of operation. The NRC seeks comment on whether the risk-informed approach provided in § 50.46(e) could generically exclude any plant operational modes (e.g., low power or shutdown) from consideration. If so, what are the bases for excluding these operational modes from consideration?

NRC Question 7.

Reporting Criteria for the Risk-Informed Alternative.

The NRC is proposing in § 50.46c(m) corrective actions and reporting criteria specific to the risk-informed approach for addressing the effects of debris on long-term cooling. These criteria are performance-based and similar in concept to the reporting criteria in § 50.69. Per proposed § 50.46c(m), the NRC's approval of the entity's risk-informed application would specify the circumstances under which the licensee or design certification applicant shall notify the NRC of changes or errors in the risk evaluation approach. In addition, the proposed rule would require entities to review the analyses, evaluations, and modeling for changes and errors and incorporate changes to the design, plant, operational practices, and operation experience. The entity would then be required to update the debris evaluation model and the PRA and its supporting analyses, and re-perform the evaluations of risk, defense-in-depth, and safety margins to confirm the acceptance criteria for the risk-informed approach continue to be met. The NRC seeks specific comment on the reporting criteria for the risk-informed approach.

Alternatively, the NRC seeks comment on whether the reporting criteria for the risk-informed approach should be more prescriptive and establish requirements similar to those for the ECCS model (i.e., § 50.46c(m)(1) through (m)(3)). For instance, should the rule establish values for changes in Δ CDF, Δ LERF, defense-in-depth, and safety margins that would trigger specific reporting actions? If so, what values should reporting criteria establish as reporting triggers and what are the bases for selecting those values?

NRC Question 8.

Exemptions Needed to Implement the Risk-Informed Alternative.

One objective of the proposed rule is to allow entities to submit a risk-informed alternative to address the effects of debris on long-term core cooling without the need to submit an exemption request. The NRC identified that, in order to eliminate the need for an exemption, changes may be necessary in GDCs 35, 38, and 41, as provided in the proposed rule. The NRC seeks input on whether conforming changes to other regulations would be necessary or desirable. Such conforming changes may avoid the need for entities wishing to use the risk-informed alternative to request exemptions from those regulations in order to effectively implement the risk-informed alternative. If you believe it is necessary or desirable to provide a conforming change to a regulation in order to avoid an exemption fr

This text is long and has been trimmed here. Open the source document for the complete record.