ML24319A108

DG-1263 Rev. 1 (RG 1.224 Rev 0) Establishing Analytical Limits for Zirconium-Based Alloy Cladding - ACRS Version

ML24319A108
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Issue date:	11/15/2024
From:	James Corson NRC/RES/DSA/FSCB
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DG-1263 Rev 1 RG 1.225 Rev 0
Download: ML24319A108 (31)

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U.S. NUCLEAR REGULATORY COMMISSION DRAFT REGULATORY GUIDE DG-1263, Revision 1 Proposed New Regulatory Guide 1.224 Issue Date: Month 20##

Technical Lead: James Corson Pre-Decisional/Public version for meetings with the Advisory Committee on Reactor Safeguards Pre-Decisional/Public version for meetings with the Advisory Committee on Reactor Safeguards This RG is being issued in draft form to involve the public in the development of regulatory guidance in this area. It has not received final staff review or approval and does not represent an NRC final staff position. Public comments are being solicited on this DG and its associated regulatory analysis. Comments should be accompanied by appropriate supporting data. Comments may be submitted through the Federal rulemaking website, http://www.regulations.gov, by searching for draft regulatory guide DG-1263. Alternatively, comments may be submitted to Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.

Comments must be submitted by the date indicated in the Federal Register notice.

Electronic copies of this DG, previous versions of DGs, and other recently issued guides are available through the NRCs public website under the Regulatory Guides document collection of the NRC Library at https://nrc.gov/reading-rm/doc-collections/reg-guides. The DG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML24304A946. The regulatory analysis is associated with a rulemaking and may be found in ADAMS under Accession No. ML24239A776.

ESTABLISHING ANALYTICAL LIMITS FOR ZIRCONIUM-BASED ALLOY CLADDING A. INTRODUCTION Purpose This regulatory guide (RG) defines an acceptable analytical limit on peak cladding temperature (PCT) and integral time at temperature that corresponds to the measured ductile-to-brittle transition for the zirconium-alloy cladding materials tested in the U.S. Nuclear Regulatory Commissions (NRCs) loss-of-coolant accident (LOCA) research program. This analytical limit is based on the data obtained in the NRCs LOCA research program. This RG also describes methods for establishing analytical limits for the ductile-to-brittle transition and breakaway oxidation susceptibility for zirconium-alloy cladding materials not included in the NRCs LOCA research program. Finally, this guide provides a hydrogen generation limit to minimize the risks associated with explosive quantities of combustible gas in containment.

Applicability This RG applies to applicants for and holders of construction permits and operating licenses for power reactors under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 1), and applicants for and holders of standard design approvals, combined licenses, and manufacturing licenses, and applicants for standard design certifications (including an applicant after the Commission has adopted a final design certification regulation), under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 2).

Applicable Regulations

• 10 CFR Part 50, including Appendix A, General Design Criteria for Nuclear Power Plants, provides regulations for licensing production and utilization facilities.

o 10 CFR 50.46a, Alternative acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, provides a voluntary alternative to 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power

DG-1263, Revision 1, Page 2 reactors. Licensees that adopt 10 CFR 50.46a are required to have NRC-approved limits that address cladding degradation phenomena and that avoid explosive concentrations of combustible gas.

o 10 CFR Part 50, Appendix A, General Design Criterion 35, Emergency core cooling, requires that the emergency core cooling system (ECCS) be designed to remove heat from the reactor core following a LOCA to prevent fuel and cladding damage that could interfere with effective core cooling and to limit clad metal-water reaction to negligible amounts.

o 10 CFR 50.46 includes cladding temperature and oxidation limits meant to ensure post-quench ductility (PQD), a limit on allowable hydrogen generation, and a requirement to maintain a coolable core geometry. While 10 CFR 50.46 does not specifically require periodic breakaway oxidation testing or testing for PDQ, the methods in this RG could be used to confirm that new zirconium-based cladding alloys meet the requirements of 10 CFR 50.46.

10 CFR Part 52 governs the issuance of early site permits, standard design certifications, combined licenses, standard design approvals, and manufacturing licenses for nuclear power facilities. The regulations in 10 CFR 52.47, Contents of applications; technical information; 10 CFR 52.79, Contents of applications; technical information in final safety analysis report; 10 CFR 52.137, Contents of applications; technical information; and 10 CFR 52.157, Contents of applications; technical information in final safety analysis report, for applications for standard design certifications, combined licenses, standard design approvals, and manufacturing licenses for nuclear power facilities, respectively, state that the applicable 10 CFR Part 50 regulations cited above are also required for applicants for and holders combined licenses, standard design approvals, and manufacturing licenses and applicants for a standard design certification (including an applicant after the Commission has adopted a final design certification regulation).

There are no similar requirements related to ECCS performance for applicants for and holders of early site permits.

Related Guidance NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP) (Ref. 3), provides guidance to the NRC staff for the review of license applications and license amendments for nuclear power plants.

o SRP Section 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary, provides guidance for reviewing LOCAs.

o SRP Section 4.2, Fuel System Design, provides guidance for reviewing reactor fuel designs.

RG 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance (Ref.

4), provides guidance for calculating realistic or best-estimate ECCS performance during LOCAs.

RG 1.203, Transient and Accident Analysis Methods (Ref. 5), describes a process that the staff considers acceptable for use in developing and assessing evaluation models that may be used to analyze transient and accident behavior that is within the design basis of a nuclear power plant.

DG-1263, Revision 1, Page 3 Draft Regulatory Guide (DG)-1261 (proposed new RG 1.222), Conducting Periodic Testing for Breakaway Oxidation Behavior (Ref. 6), describes a method to address a zirconium-based cladding alloy embrittlement mechanism known as breakaway oxidation.

DG-1262 (proposed new RG 1.223), Testing for Post-Quench Ductility (Ref. 7), provides a method for measuring the ductile-to-brittle transition for a zirconium-based cladding alloy as a function of hydrogen content.

Purpose of Regulatory Guides The NRC issues RGs to describe methods that are acceptable to the staff for implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific issues or postulated events, and to describe information that the staff needs in its review of applications for permits and licenses. Regulatory guides are not NRC regulations and compliance with them is not required. Methods and solutions that differ from those set forth in RGs are acceptable if the applicant provides sufficient basis and information for the NRC staff to verify that the alternative methods comply with the applicable NRC regulations.

Paperwork Reduction Act This RG provides voluntary guidance for implementing the mandatory information collections in 10 CFR Parts 50 and 52 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et. seq.). These information collections were approved by the Office of Management and Budget (OMB), under control numbers 3150-0011 and 3150-0151, respectively. Send comments regarding this information collection to the FOIA, Library, and Information Collections Branch, Office of the Chief Information Officer, Mail Stop: T6-A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-0011 and 3150-0151), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street, NW, Washington, DC, 20503.

Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number.

DG-1263, Revision 1, Page 4 TABLE OF CONTENTS A. INTRODUCTION................................................................................................................................... 1 B. DISCUSSION.......................................................................................................................................... 5 C. STAFF REGULATORY GUIDANCE.................................................................................................... 9 D. IMPLEMENTATION............................................................................................................................ 18 GLOSSARY............................................................................................................................................... 19 REFERENCES........................................................................................................................................... 20 APPENDIX A............................................................................................................................................... 1 APPENDIX B............................................................................................................................................... 1

DG-1263, Revision 1, Page 5 B. DISCUSSION Reason for Issuance Licensees implementing the alternative ECCS requirements in 10 CFR 50.46aand using uranium oxide or mixed uranium-plutonium oxide pellets within zirconium-alloy cladding must address cladding degradation phenomena. This guide provides methods to establish analytical limits to prevent cladding embrittlement and to ensure cladding post-quench ductility following a postulated LOCA. This guide also provides a hydrogen generation limit that precludes the formation of a combustible gas mixture in the reactor coolant system or in containment. This limit is identical to the historical limit included in 10 CFR 50.46(b)(3). The basis for this limit is discussed elsewhere (e.g., in the Atomic Energy Commission opinion paper (Ref. 8) that established the ECCS requirements in 10 CFR 50.46).

Background

In 1996, the NRC initiated a fuel-cladding research program intended to investigate the behavior of high-exposure fuel cladding under accident conditions. This program included an extensive LOCA research and testing program at Argonne National Laboratory, as well as jointly funded programs at the Kurchatov Institute (Ref. 9) and the Halden Reactor Project (Ref. 10), to develop the body of technical information needed to evaluate LOCA regulations for high-exposure fuel. The research findings were summarized in Research Information Letter-0801, Technical Basis for Revision of Embrittlement Criteria in 10 CFR 50.46, dated May 30, 2008 (Ref. 11). Most of the detailed experimental results from the program at Argonne National Laboratory appear in NUREG/CR-6967, Cladding Embrittlement during Postulated Loss-of-Coolant Accidents, issued July 2008 (Ref. 12), and NUREG/CR-7219, Cladding Behavior during Postulated Loss-of-Coolant Accidents, issued July 2016 (Ref. 13).

The research results revealed that hydrogen, which is absorbed into the cladding from corrosion during normal operation, has a significant influence on embrittlement during a hypothetical LOCA. 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, both of which promote embrittlement. Thus, corroded cladding with significant hydrogen pickup that is exposed to high-temperature LOCA conditions can experience embrittlement before reaching the regulatory acceptance criterion of 17 percent oxidation in 10 CFR 50.46(b)(2), which was derived from experiments conducted in the early 1970s with unirradiated cladding. The research results also revealed that an embrittlement mechanism referred to as breakaway oxidation may occur during prolonged exposure to elevated cladding temperature during a LOCA.

NUREG/CR-6967 summarizes most of the cladding embrittlement experimental results from the NRCs LOCA research program. Since the publication of NUREG/CR-6967 in 2008, additional testing has been conducted, focusing on cladding materials with hydrogen contents in the range of 200-350 weight parts per million (wppm). Additional oxidation and PQD tests were conducted with cladding samples sectioned from high-burnup ZIRLOTM fuel rods. The two defueled segments used to prepare samples had a corrosion-layer thickness of 25-30 micrometers and 300-340 wppm of hydrogen in the cladding metal before oxidation. Also, the ductility data for an oxidation sample with approximately 600 wppm hydrogen were reassessed. In addition, since the publication of NUREG/CR-6967, oxidation and PQD tests were conducted with pre-hydrided cladding samples containing 200-300 wppm of hydrogen. These additional tests are summarized in NUREG/CR-7219.

Before combining the more recent data from NUREG/CR-7219 with the data reported in NUREG/CR-6967, two refinements were made in data assessment:

DG-1263, Revision 1, Page 6 The first refinement was to establish and verify the following ductility criteria:

o average permanent strain 1.0 percent o or, if permanent strain cannot be measured, the average ring compression test (RCT) offset strain 1.41 percent + 0.1082 Cathcart-Pawel equivalent cladding reacted (CP-ECR) (Ref. 14). When rounded to the nearest tenth of a percent, the output of this correlation represents the one-sigma upper bound of offset strain values from 65 RCT data sets with 1.0 to 2.3 percent permanent strain. (DG-1262 and appendix A to this RG discuss in detail a ductility criterion based on RCT offset strain.)

The second refinement was to develop and use a new methodology to determine the pretest hydrogen content in the cladding metal for corroded cladding (Ref. 13).

NUREG/CR-7219 reassessed ductility and hydrogen data presented in NUREG/CR-6967 to determine embrittlement oxidation levels versus hydrogen content for pre-hydrided and high-burnup cladding. When subsequent tests were combined with the data reported in NUREG/CR-6967, and the refinements in hydrogen content and the relationship between offset and permanent strain were made, the description of the resulting behavior of cladding embrittlement as a function of hydrogen content could be depicted as shown in figure 1.

For modern as-fabricated cladding (Zry-2, Zry-4, ZIRLOTM, and M5), embrittlement thresholds cluster at 19-20 percent CP-ECR, as compared to 16 percent CP-ECR for older Zry-4 cladding. However, this improvement relative to the as-fabricated cladding specimens used to develop the existing acceptance criteria for cladding ductility in 10 CFR 50.46(b) is negated for modern claddings with hydrogen pickup as low as 100 wppm. A bilinear function for CP-ECR versus hydrogen content was used to fit the embrittlement data for pre-hydrided and high-burnup cladding. The embrittlement rate is steep for cladding with 400-wppm hydrogen. For higher hydrogen content, the embrittlement rate is more gradual. Embrittlement is highly sensitive to both hydrogen content and peak oxidation temperature, and both of these factors should be considered when determining acceptable limits.

Note that the tests performed in the NRCs LOCA research program were conducted at PCTs less than or equal to approximately 1,200 degrees Celsius (oC) (approximately 2,200 degrees Fahrenheit (oF).

Above this temperature, oxygen solubility and diffusion in the beta zirconium layer become high enough to lead to cladding embrittlement following cladding cooldown and quench (Ref. 13). This is the basis for the limit of 2,200oF (1,204oC) in 10 CFR 50.46(b)(1), as discussed in Refs. 8 and 13.

DG-1263, Revision 1, Page 7 Figure 1. Ductile-to-brittle transition oxidation level (CP-ECR) as a function of pretest hydrogen content in cladding metal for as-fabricated, pre-hydrided, and high-burnup cladding materials. Samples were oxidized at 1,200 °C +/-10 °C and quenched at 800 °C. For high-burnup cladding with about 550-wppm hydrogen, embrittlement occurred during the heating ramp at 1,160-1,180 °C peak oxidation temperatures (Ref. 13).

Consideration of International Standards The International Atomic Energy Agency (IAEA) works with member states and other partners to promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops Safety Requirements and Safety Guides for protecting people and the environment from harmful effects of ionizing radiation. This system of safety fundamentals, safety requirements, safety guides, and other relevant reports, reflects an international perspective on what constitutes a high level of safety. To inform its development of this RG, the NRC considered IAEA Safety Requirements and Safety Guides pursuant to the Commissions International Policy Statement (Ref. 16) and Management Directive and Handbook 6.6, Regulatory Guides (Ref. 17).

The following IAEA Safety Requirements and Guides were considered in the development of the Regulatory Guide:

IAEA SSG-52, Design of the Reactor Core for Nuclear Power Plants, issued in 2019 (Ref. 18),

states that limits should be established to prevent cladding embrittlement due to oxidation at high temperature.

DG-1263, Revision 1, Page 8 IAEA SSG-56, Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants, issued in 2020 (Ref. 19), includes high-level information about emergency core cooling system requirements for loss-of-coolant accidents.

DG-1263, Revision 1, Page 9 C. STAFF REGULATORY GUIDANCE Staff Regulatory Position 1 provides acceptable methods for establishing an analytical limit on PCT and integral time at temperature to ensure PQD for zirconium-alloy cladding materials. Applicants should use one of the four methods provided. These methods are meant to prevent cladding embrittlement and ensure PQD. All of these methods use a PCT no greater than 1,204°C (2,200°F), which would also prevent failure due to cladding degradation mechanisms that occur at higher temperatures (e.g., excessive exothermic metal-water reaction, alloy-specific eutectics, and loss of fuel rod geometry due to plastic deformation). Regulatory Position 2 provides an acceptable method for establishing an analytical limit for breakaway oxidation. Regulatory Position 3 provides an acceptable limit on hydrogen generation.

1.

Establishing an acceptable analytical limit on peak cladding temperature and integral time at temperature In 10 CFR 50.46a(f)(1)(i), the NRC requires the licensee to establish analytical limits to address cladding degradation phenomena for all fuel system designs. To prevent adverse cladding degradation for zirconium-based alloy cladding, licensees should establish limits on PCT and integral time at temperature, which correspond to the measured ductile-to-brittle transition for the zirconium-alloy cladding material.

The ductile-to-brittle threshold defined in figure 2 is an acceptable analytical limit on integral time at temperature as calculated in local oxidation calculations using the Cathcart-Pawel (CP) correlation. This analytical limit is acceptable for the zirconium-alloy cladding materials tested in the NRCs LOCA research program, which were Zry-2, Zry-4, ZIRLOTM, and M5. (Refs. 12-13). Since PQD tests above 400-wppm hydrogen were conducted at a peak oxidation temperature below 1,204°C (2,200°F), a separate PCT analytical limit must be defined that is consistent with test temperature. A limit on PCT of 1,204°C (2,200°F) below 400-wppm cladding hydrogen content and 1,121°C (2,050°F) at or above 400-wppm cladding hydrogen content is acceptable.

Figure 2. An acceptable analytical limit on PCT and integral time at temperature (as identified in local oxidation calculations using the CP correlation).

0 2

4 6

8 10 12 14 16 18 20 0

100 200 300 400 500 600 700 800 Embrittlement Oxidation Limit (% ECR)

Pre-Transient Hydrogen Content (wppm)

PCT 2200°F PCT 2050°F

DG-1263, Revision 1, Page 10 For zirconium-alloy cladding materials beyond those tested in the NRCs LOCA research program (i.e., Zry-2, Zry-4, ZIRLOTM, and M5), a demonstration of comparable performance with the database established in the NRCs LOCA research program would be necessary to establish the analytical limit provided in this guide as the limit for that alloy. DG-1262 provides an experimental technique acceptable to the NRC for measuring the ductile-to-brittle transition for zirconium-alloy cladding material through RCTs. The experimental technique in DG-1262 may be used to establish whether the analytical limit shown in figure 2 is acceptable for a zirconium-alloy not included in the NRCs research program.

Alternatively, if a zirconium-alloy cladding material experiences the transition from ductile to brittle behavior at a different level of oxidation than the established database, then this RG (i.e., RG-1.224) also describes a methodology to establish a limit specific to a zirconium alloy other than the limit provided in this guide.

The database established in the NRCs LOCA research program and the resulting analytical limit described in this guide are intended to bound the envelope of fuel conditions envisioned for existing reactor core designs and ECCS capabilities. In the test program, experiments were conducted at maximum oxidation temperatures permitted by the criteria in 10 CFR 50.46. However, some core operating limits and ECCS designs may result in maximum oxidation temperatures significantly below 1,204°C (2,200°F).

Maintaining lower cladding temperatures has been shown to increase the allowable calculated total oxidation before reaching embrittlement. Therefore, conducting tests at lower peak temperatures may provide additional margin for some zirconium-alloy cladding materials. This RG describes a method to establish analytical limits at peak oxidation temperatures less than 1,204°C (2,200°F).

In summary, this guide provides the following four options for establishing limits of peak cladding temperature and integral time at temperature. In all cases, the peak cladding temperature should not exceed 1,204°C (2,200°F). Showing that ECCS performance is such that local oxidation and PCT are calculated below the analytical limits defined using one of the following methods is acceptable to demonstrate that PQD is maintained.

a.

Apply the specified and acceptable limit defined in figure 2 of this RG for cladding materials tested in the NRCs LOCA research program.

b.

Demonstrate comparable behavior of cladding alloys not tested in the NRCs LOCA research program with the database established in the NRCs LOCA research program, in order to apply the limit defined in figure 2 of this RG.

To apply the analytical limit in Figure 2 to zirconium-alloy cladding materials not tested in the NRCs LOCA research program, a demonstration of comparable performance with the established database would be necessary. The objective of PQD testing to demonstrate consistency with the analytical limit provided in figure 2 of this RG is to confirm that the transition to brittle behavior does not take place at a lower equivalent cladding reacted (ECR) than the provided limit.

(1)

Conduct oxidation and quench testing on (1) as-fabricated material, (2) pre-hydrided material for the entire range of a cladding materials anticipated hydrogen level (i.e., testing pre-hydrided material in increments not more than every 100 wppm hydrogen), and (3) irradiated material for the entire range of a cladding materials anticipated hydrogen level (i.e., testing irradiated material with hydrogen contents within 50 wppm of the anticipated maximum hydrogen content and within 50 wppm of half of the anticipated maximum hydrogen content) at the transition ECR defined in figure 2 for each samples hydrogen level, an ECR 1 percent above this limit, and an ECR 1 percent below this limit.

Section B-2 of appendix B provides an acceptable test matrix.

DG-1263, Revision 1, Page 11 (2)

Determine the ECR at which the material transitions from ductile to brittle behavior using the results of ring compression testing conducted using the experimental procedure and the guidance in DG-1262 for each material condition in Regulatory Position 1.b.(1).1 Compare to the limit defined in figure 2 of this guide.

(3)

If the experimental results for the new fuel design measured the transition from ductile to brittle behavior to be no lower than the analytical limit defined in figure 2, then the analytical limit defined in that figure may be established for the cladding alloy not tested in the NRCs LOCA research program.

(4)

Provide details of experimental techniques, unless conducted using the guidance in DG-1262, as part of the documentation supporting the request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report).

(5)

Provide results of experiments conducted with as-fabricated, irradiated material and identify the specific analytical limit on PCT and integral time at temperature as part of the documentation supporting the request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report).

The limit should correspond to the ductile-to-brittle transition for the zirconium-alloy cladding material and the oxidation temperature of the oxidation and quench experiments.

c.

Establish a zirconium-alloy-specific analytical limit on PCT and integral time at temperature at a peak cladding oxidation temperature of 1,204°C (2,200°F).

The analytical limit described in this applies to Zry-4, Zry-2, ZIRLOTM, and M5. While the existing database and resulting analytical limit described here provide a best-estimate limit for the ductile-to-brittle transition for zirconium alloys, it is possible that some zirconium-alloy cladding materials may experience the transition from ductile to brittle behavior at a different level of oxidation than the established database.

PQD testing to establish an alloy-specific limit should characterize a cladding alloys embrittlement behavior through the entire spectrum of conditions expected during operation. A diverse matrix of material conditions can provide a complete characterization, and repeat testing can be used to address expected variability in oxidation behavior.

(1)

Conduct oxidation and quench testing on (1) as-fabricated material, (2) pre-hydrided material for the entire range of a cladding materials anticipated hydrogen level (i.e., testing pre-hydrided material in increments not more than every 100 wppm hydrogen), and (3) irradiated material for the entire range of a cladding materials anticipated hydrogen level (i.e., testing irradiated material with hydrogen contents within 50 wppm of the anticipated maximum hydrogen 1

The term each material condition refers to the range of as-fabricated, pre-hydrided, and irradiated material in the Discussion section of this RG. For a zirconium alloy with an anticipated, end-of-life hydrogen content, the range of material conditions within the Discussion section of this RG would include (1) as-fabricated material, (2) pre-hydrided material at 100 wppm hydrogen, (3) pre-hydrided material at 200 wppm hydrogen, (4) pre-hydrided material at 300 wppm hydrogen, (5) pre-hydrided material at 400 wppm hydrogen, (6) irradiated material with a hydrogen content of 200+/-50 wppm, and (7) irradiated material with a hydrogen content of 400+/-50 wppm. See also appendix B to this guide for a high-level overview of an acceptable test matrix.

DG-1263, Revision 1, Page 12 content and within 50 wppm of half of the anticipated maximum hydrogen content) at four oxidation levels for each material condition (see footnote 3), in increments not greater than 2 percent ECR. Section B-1 of appendix B provides an acceptable test matrix.

(2)

With the results of four oxidation levels, for each material condition in (a) above, determine the ECR range in which the transition from ductile-to-brittle behavior occurs and conduct three repeat oxidation and quench tests at each ECR level within this range, using the guidance in DG-1262.

(3)

Determine the ECR at which the material transitions from ductile to brittle behavior, using the results of ring compression testing conducted using the experimental procedure and the guidance in DG-1262 for each material condition in Regulatory Position 1.c.(1).

(4)

Provide details of experimental techniques, unless conducted using the guidance in DG-1262, as part of the documentation supporting the request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report).

(5)

Provide the results of experiments conducted with as-fabricated, irradiated material and identify the specific analytical limit on PCT and integral time at temperature as part of the documentation supporting the request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report). The limit should correspond to the ductile-to-brittle transition for the zirconium-alloy cladding material and the oxidation temperature of the oxidation and quench experiments.

d.

Establish an analytical limit on PCT and integral time at temperature at a peak oxidation temperature less than 1,204oC (2,200°F).

The existing database and resulting analytical limit described in figure 2 of this guide is intended to bound the envelope of fuel conditions envisioned for existing reactor core designs and ECCS capabilities. In the test program, experiments were conducted at maximum oxidation temperatures 1,200°C +/-10°C (2,200°F +/-20°F) and quenched at 800°C (1,472°F).2 However, some core operating limits and ECCS designs may result in maximum oxidation temperatures significantly below 1,204°C (2,200°F). Oxidation at lower temperatures has been shown to increase the allowable calculated oxidation before embrittlement. Therefore, conducting tests at lower peak temperatures may provide additional margin for zirconium-alloy cladding materials.

PQD testing to establish a limit at a PCT lower than 1,204°C (2,200°F) should characterize a cladding alloys embrittlement behavior through the entire spectrum of conditions expected during operation. A diverse matrix of material conditions can provide a complete characterization, and repeat testing can be used to address expected variability in oxidation behavior.

2 These test conditions were selected to be relevant and bounding for current light-water reactor core operating limits and ECCS designs. However, it may be necessary to evaluate and possibly modify the conditions accordingly for advanced reactor designs. In addition PQD measurements were made at 135°C (275°F), which is approximately equal to the saturation temperature at the expected post-LOCA containment pressure for current light-water reactors (Ref. 13), but it require modification for advanced designs.

DG-1263, Revision 1, Page 13 (1)

Conduct oxidation and quench testing on (1) as-fabricated material, (2) pre-hydrided material for the entire range of a cladding materials anticipated hydrogen level (i.e. testing pre-hydrided material in increments not more than every 100 wppm of hydrogen), and (3) irradiated material for the entire range of a cladding materials anticipated hydrogen level (i.e., testing irradiated material with hydrogen contents within 50 wppm of the anticipated maximum hydrogen content and within 50 wppm of half of the anticipated maximum hydrogen content) at four oxidation levels for each material condition (see footnote 3), in increments not greater than 2 percent ECR. Section B-1 of appendix B provides an acceptable test matrix.

(2)

With the results of four oxidation levels for each material condition in Regulatory Position 1.d.(1), determine the ECR range in which the transition from ductile to brittle behavior occurs and conduct three repeat oxidation and quench tests at each ECR level within this range using the guidance in DG-1262.

(3)

Determine the ECR at which the material transitions from ductile-to-brittle behavior, using the results of ring compression testing conducted using the experimental procedure and the guidance in DG-1262 for each material condition in Regulatory Position 1.d.(1).

(4)

Provide details of experimental techniques, unless conducted using the guidance in DG-1262, as part of the documentation supporting the request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report).

(5)

Provide the results of experiments conducted with as-fabricated, irradiated material and identify the specific analytical limit on PCT and integral time at temperature as part of the documentation supporting the request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report). The limit should correspond to the ductile-to-brittle transition for the zirconium-alloy cladding material and the oxidation temperature of the oxidation and quench experiments.

2.

Establishing analytical limits for breakaway oxidation The purpose of the requirements in 10 CFR 50.46 is to ensure core coolability during and following a LOCA. If breakaway oxidation occurs, then the embrittlement process is accelerated.

Therefore, the PQD analytical limits established in accordance with 10 CFR 50.46 are no longer effective in precluding embrittlement, and core coolability may not be maintained even if the analytical limits on PCT and local oxidation (surrogate for time at temperature) are not exceeded.

In 10 CFR 50.46a, the NRC requires applicants to address cladding degradation phenomena, which, for licensees using zirconium-based alloys, includes breakaway oxidation.

Based on data reported by Leistikow and Schanz (Ref. 15), zirconium alloys have been shown to be susceptible to the breakaway oxidation phenomenon at temperatures as low as 650°C (1,202°F). At 650°C (1,202°F), it took more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (beyond LOCA-relevant times for conventional reactors) for Zry-4 to accumulate 200 wppm hydrogen, while at 800°C (1,472°F), the time to accumulate 200 wppm hydrogen was only 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (within LOCA-relevant times). Thus, time spent in steam at 650°C (1,202°F) was benign with regard to breakaway oxidation and hydrogen accumulation because of the very low oxidation rate. Because NUREG/IA-0211, Experimental Study of Embrittlement of Zr-1%Nb VVER

DG-1263, Revision 1, Page 14 Cladding under LOCA-Relevant Conditions, issued March 2005 (Ref. 9), did not present hydrogen-accumulation data for temperatures between 650°C and 800°C (1,202°F and 1,472°F), there is no basis for not including time spent at temperatures >650°C (1,202°F) in establishing the analytical limit for transient time.

To establish a zirconium-alloy-specific limit for a new or existing fuel design, the applicant would provide experimental results for testing for breakaway oxidation behavior as part of the documentation supporting its request for NRC approval of the new or existing fuel design (e.g., through a license amendment request or vendor topical report). DG-1261 provides an experimental technique to measure the onset of breakaway oxidation to establish a specified and acceptable limit on the total accumulated time that a cladding may remain at high temperature. The applicant would provide details of the experimental technique (unless the experiments were conducted in accordance with DG-1261) and the results of experiments conducted. Applicants would establish the time limit for the total accumulated time that the cladding may remain above 650 °C (1,202°F) as part of the documentation supporting its request for NRC approval of the new or existing fuel design (e.g., through a license amendment request or vendor topical report).

Applicants may elect to establish the analytical limit for breakaway oxidation with conservatism relative to the measured minimum time (i.e., reduce the time) to the onset of breakaway oxidation. This approach may reduce the likelihood of reassessing small-break LOCA cladding temperature histories in the event of a minor change in measured time to breakaway oxidation. For example, the minimum time to breakaway oxidation may be demonstrated to occur at 975°C (1,787°F) at a time of 4,000 seconds. An applicant may elect to establish an analytical limit of 3,000 seconds for the total accumulated time that the cladding may remain above 650°C (1,202°F) (1,202°F).

Upon NRC approval of the fuel design, an acceptable method to demonstrate avoidance of breakaway oxidation is to ensure 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 this phenomenon is not greater than the proposed limit.

In summary, applicants should take the following steps to add establish an analytical limit for breakaway oxidation in zirconium-based alloy cladding.

a.

Follow the procedures in DG-1261 to establish the shortest time observed to lead to breakaway oxidation for a zirconium-alloy cladding.

b.

Provide the results of the testing as part of the documentation supporting the request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report).

c.

Establish an analytical limit for the total accumulated time the cladding may remain above 650°C (1,202oF), which is less than or equal to the shortest time observed to lead to breakaway oxidation.

d.

Provide the analytical limit for breakaway oxidation as part of the documentation supporting the request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report).

3.

Establishing an analytical limit on combustible gas generation The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam should not exceed 0.01 times the hypothetical amount that would be

DG-1263, Revision 1, Page 15 generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4.

Applying analytical limits

a.

Qualification of hydrogen pickup models A licensee will need an alloy-specific cladding hydrogen uptake model if the licensee chooses to use the hydrogen-dependent embrittlement threshold provided in this RG. To establish an alloy-specific cladding hydrogen uptake model for a new or existing fuel design, the applicant would provide post-irradiation examination hydrogen measurement data and a hydrogen uptake model as part of the documentation supporting its request for NRC approval of the new or existing fuel design (e.g., through a license amendment request or vendor topical report). The documentation should include a cladding-specific plot of predicted-versus-measured cladding hydrogen content. The post-irradiation examination data supporting the hydrogen uptake model should include values for multiple burnup levels, encompass all applicable operating conditions and reactor coolant chemistry conditions, and should quantify axial, radial, and circumferential variability. (See the next section for further details.)

b.

Accounting for uncertainty and variability in hydrogen content Many investigators have observed variation of hydrogen content across the radius of the cladding (hydride rim effect) and over short axial distances (pellet-pellet interface effect). Studies using pre-hydrided Zry-4 with dense hydride rims have demonstrated that the homogenization of hydrogen across the radius of the cladding is very rapid at >900°C (1,652°F) because of the affinity of the beta phase for hydrogen, as well as the high solubility of hydrogen in this phase. In the NRCs LOCA research program, significant circumferential variation (+/-100-140 wppm) in hydrogen content was measured and observed for high-burnup cladding alloys. For oxidation test times at 1,200°C (2,200°F) up to the embrittlement CP-ECR level, no significant diffusion of hydrogen in the circumferential direction was observed. Hydrogen concentration variations of 450 to 750 wppm measured for cladding quarter segments before LOCA testing remained after LOCA testing.

The uncertainty in models of the cladding hydrogen distribution can be characterized and quantified by supporting the model with post-irradiation examination that includes values for multiple burnup levels; encompasses all applicable operating conditions and reactor coolant chemistry; and quantifies axial, radial, and circumferential variability.

To apply the analytical limit in figure 2 to an individual fuel rod (or fuel rod grouping),

the allowable CP-ECR should be based on predicted peak circumferential-average hydrogen content for the individual rod (or fuel rod grouping).

c.

Post-quench ductility analytical limits Based on the approved ECCS evaluation models and methods, the applicant should identify the limiting combination of break size, break location, and initial conditions and assumptions that maximize predicted PCT and local oxidation (surrogate for time at temperature).

Appropriate combinations of initial conditions and uncertainties will vary for different design conditions as well as the regulatory requirements under which the evaluation model has been developed, including evaluation models that satisfy the following:

DG-1263, Revision 1, Page 16 the required and acceptable features of Appendix K, ECCS Evaluation Models, to 10 CFR Part 50 the evaluation model requirements for realistically considering the behavior of the reactor system during a LOCA, with consideration of uncertainties, specified in 10 CFR 50.46(a)(3)(i), or for postulated LOCAs smaller than the transition break size, in 10 CFR 50.46a(e)(2) the requirements for realistic methods (i.e., with no consideration of uncertainties), as described in 10 CFR 50.46a(e)(3), which are intended for postulated LOCAs larger than the transition break size where licensees have demonstrated such breaks may be addressed as beyond-design-basis events Separate cases may be necessary to identify the limiting scenario for PCT relative to local oxidation and vice versa. The applicant should demonstrate that the predicted PCT remains below the lesser of 1,204 °C (2,200°F) and the maximum oxidation PQD temperature for the relevant cladding hydrogen content. The applicant should also demonstrate that the maximum predicted local oxidation remains below the established PQD analytical limits.

Because of the strong function of allowable local oxidation with cladding hydrogen content (see figure 1), the applicant may elect to subdivide the fuel rods within the core based on characteristics such as cladding hydrogen content, burnup, fuel rod power, or a combination thereof. For example, PCT and local oxidation calculations could be performed on three representative sets of fuel rods (e.g., 0-30 gigawatt-days per metric ton of uranium (GWd/MTU),

30-45 GWd/MTU, and 45-62 GWd/MTU) using bounding power histories and other attributes for each fuel rod grouping. The predicted PCT and local oxidation would then be compared to the analytical limits for that range of burnup/hydrogen.

d.

Application in the Rupture Region During a postulated LOCA, fuel rods may be predicted to balloon and rupture as a result of elevated cladding temperature and differential pressure (between rod internal pressure and system pressure, which is decreasing because of a break in pressure boundary). The regions of the fuel rod near the ballooned and ruptured location will thus be exposed to oxidation from the inside surface of the cladding. Combined with oxygen diffusion from the cladding outside diameter (OD), oxygen diffusion from the cladding inside diameter (ID) would further limit integral time at temperature to reach the analytical limit in Figure 2. In addition, local regions above and below the rupture opening will absorb significant hydrogen because of the steam oxidation on the ID, which may result in locally brittle regions above and below the rupture.

Finally, the balloon region will experience wall thinning, which impacts the calculation of ECR because the value is taken to be a percentage of the pre-oxidation cladding thickness.

The LOCA acceptance criteria that limit peak oxidation temperature and maximum oxidation level versus hydrogen content are based on retention of ductility. As discussed above, ductility will not be retained everywhere in the balloon region.

To investigate the mechanical behavior of ruptured fuel rods, the NRC conducted integral LOCA testing, designed to experience ballooning and rupture, on as-fabricated and hydrogen-charged cladding specimens and high-burnup fuel rod segments exposed to high temperature steam oxidation followed by quench (Ref. 13). The integral LOCA testing confirms that continued exposure to high-temperature steam environments weakens the already-flawed region

DG-1263, Revision 1, Page 17 of the fuel rod surrounding the cladding rupture. Hence, limitations on integral time at temperature 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 preexisting cladding hydrogen content.

Therefore, in regions of the fuel rod where the calculated conditions of transient pressure and temperature lead to a prediction of cladding swelling, an acceptable approach would be to (1) define the cladding thickness as the cladding cross-sectional area divided by the cladding circumference, taken at a horizontal plane at the elevation of the rupture, and (2) calculate two-sided oxidation using the CP correlation and apply the analytical limit in figure 2 (or an alternative specified and acceptable analytical limit).

e.

Accounting for Double-Sided Oxidation Due to the Fuel-Cladding Bond Layer The NRCs LOCA research program identified that, for high-burnup fuel, oxygen can diffuse into the cladding metal during a LOCA from the ID as well as from the OD, even when no steam oxidation is occurring on the ID (Refs. 10 and 11). The ID oxygen diffusion phenomenon was discovered in the United States in 1977, confirmed by tests in Germany in 1979, and is seen in the present results.

Combined with oxygen diffusion from the cladding OD, oxygen diffusion from the cladding ID would further limit integral time at temperature to nil ductility. To account for the observation that oxygen can diffuse into the cladding metal during a LOCA from the ID, one acceptable approach would be to calculate two-sided local oxidation for fuel rods with a local (nodal) exposure beyond 30 GWd/MTU. The NRC notes that there would be no metal-water-reaction heat associated with this process on the ID, in contrast to the situation in a rupture node.

A licensee may propose a threshold for the onset of this inside surface oxidation source other than 30 GWd/MTU and provide it as part of the documentation supporting its request for NRC approval of the new or existing fuel design (e.g., license amendment request or vendor topical report). A threshold other than 30 GWd/MTU could be supported by metallographic images of bonding layers as a function of burnup.

f.

Breakaway Oxidation Analytical Limits Based on the approved ECCS evaluation models and methods, the applicant should identify the limiting combination of break size, break location, and initial conditions and assumptions that maximize 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 this phenomenon. The applicant should demonstrate that this time interval remains below the established alloy-specific breakaway oxidation analytical limit.

The applicant may credit operator actions to limit the duration at elevated temperatures provided that credit for these actions is consistent with the plant design basis and emergency operating procedures, and the timing of such actions is validated by operator training on the plant simulator.

DG-1263, Revision 1, Page 18 D. IMPLEMENTATION Licensees generally are not required to comply with the guidance in this regulatory guide. If the NRC proposes to use this regulatory guide in an action that would constitute backfitting, as that term is defined in 10 CFR 50.109, Backfitting, and as described in NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests (Ref. 20); affect the issue finality of an approval issued under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants; or constitute forward fitting, as that term is defined in Management Directive 8.4, then the NRC staff will apply the applicable policy in Management Directive 8.4 to justify the action.

If a licensee believes that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may inform the NRC staff in accordance with Management Directive 8.4.

DG-1263, Revision 1, Page 19 GLOSSARY breakaway oxidation The fuel-cladding oxidation phenomenon in which the 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 loss of cladding ductility.

corrosion The formation of a zirconium oxide layer resulting from the reaction of zirconium with coolant water during normal operation.

loss-of-coolant accident (LOCA)

A hypothetical accident that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system.

offset strain The value determined from a load-displacement curve by the following procedure: (1) linearize the initial loading curve, (2) use the slope of the initial loading curve to mathematically unload the sample at the peak load before a significant load drop (30-50 percent) indicating a through-wall crack along the length of the sample, and (3) determine the offset displacement (distance along the displacement axis between loading and unloading lines). This offset displacement is normalized to the outer diameter of the preoxidized cladding to determine a relative plastic strain.

oxidation The formation of a zirconium oxide layer resulting from the reaction of zirconium with high-temperature steam during LOCA conditions.

permanent strain The difference between the posttest outer diameter (after the sample is unloaded) and the pretest outer diameter of a cladding ring, normalized to the initial diameter of the cladding ring.

DG-1263, Revision 1, Page 20 REFERENCES 3

1.

U.S. Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, Energy.

2.

CFR, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Chapter I, Title 10, Energy.

3.

U.S. Nuclear Regulatory Commission (NRC), NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Washington, DC.

4.

NRC, Regulatory Guide (RG) 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, Washington, DC.

5.

NRC, RG 1.203, Transient and Accident Analysis Methods, Washington, DC.

6.

NRC, Draft Regulatory Guide (DG)-1261 (proposed new RG 1.222), Conducting Periodic Testing for Breakaway Oxidation Behavior, Washington, DC.

7.

NRC, DG-1262 (proposed new RG 1.223), Testing for Post-Quench Ductility, Washington, DC.

8.

U.S. Atomic Energy Commission, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power ReactorsOpinion of the Commission, December 28, 1973. (ML20236U832)

9.

NRC, NUREG/IA 0211, Experimental Study of Embrittlement of Zr-1%Nb VVER Cladding under LOCA-Relevant Conditions, U.S. Nuclear Regulatory Commission, Washington, DC, March 2005. (ML051100343)

10.

Institute for Energy Technology, IFE/KR/E-2008/004, LOCA Testing of High Burnup PWR Fuel in the HBWR. Additional PIE on the Cladding of the Segment 650-5, Kjeller, Norway, April 2008. (ML081750715)

11.

NRC, Research Information Letter (RIL)-0801, Technical Basis for Revision of Embrittlement Criteria in 10 CFR 50.46, Washington, DC, May 30, 2008. (ML081350225)

12.

NRC, NUREG/CR-6967, Cladding Embrittlement during Postulated Loss-of-Coolant Accidents, Washington, DC, July 2008. (ML082130389) 3 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections// and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or 800-397-4209, or e-mail pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an e-mail to pdr.resource@nrc.gov or call 800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

DG-1263, Revision 1, Page 21

13.

NRC, NUREG/CR-7219, Cladding Behavior during Postulated Loss-of-Coolant Accidents, Washington, DC, July 2016. (ML16211A004)

14.

NRC, ORNL/NUREG-17, Zirconium Metal-Water Oxidation Kinetics IV. Reaction Rate Studies, Washington, DC, August 1977. (ML052230079)

15.

Leistikow, S., and G. Schanz, Oxidation Kinetics and Related Phenomena of Zircaloy-4 Fuel Cladding Exposed to High Temperature Steam and Hydrogen-Steam Mixtures under PWR Accident Conditions, Nuclear Engineering and Design, 103: 65-84, August 1987.4

16.

NRC, Nuclear Regulatory Commission International Policy Statement, Federal Register, Vol.

79, No. 132, pp. 39415-39418 (79 FR 39415), Washington, DC, July 10, 2014.

17.

NRC, Management Directive (MD) and Handbook 6.6, Regulatory Guides, Washington, DC.

18.

International Atomic Energy Agency (IAEA), Specific Safety Guide (SSG)-52, Design of the Reactor Core for Nuclear Power Plants, Vienna, Austria, 2019.5

19.

IAEA, SSG-56, Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants, Vienna, Austria, 2020.

20.

NRC, Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, Washington, DC.

4 Nuclear Engineering and Design is a publication of Elsevier Inc., 230 Park Avenue, 7th floor, New York, NY 10169, telephone: 212-309-8100. Copies of Elsevier books and journals can be purchased at its website:

https://www.elsevier.com/books-and-journals.

5 Copies of International Atomic Energy Agency (IAEA) documents may be obtained through its website: www.iaea.org/ or by writing the International Atomic Energy Agency, P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria.

DG-1263, Revision 1, Appendix A, Page A-1 APPENDIX A RELATIONSHIP BETWEEN OFFSET STRAIN AND PERMANENT STRAIN For as-fabricated cladding compressed at room temperature (RT) or 135 degrees Celsius (°C) and at 0.033 millimeter per second (mm/s) to a total displacement of 2 millimeters (mm), the difference between offset displacement and permanent displacement is 0.2 mm, which corresponds to a strain difference of 2 percent. As the applied displacement is decreased, the plastic deformation decreases and the deviation between offset and permanent strain also decreases. This was demonstrated by conducting a set of ring compression tests designed to result in low permanent strains of 1.0 to 2.3 percent. Table A-1 shows the results of these tests.

Table A-1. Results of Ring Compression Tests Conducted with As-Fabricated Cladding Samples at RT and 2 mm/minute Displacement Rate. Total Applied Displacements Were Chosen to Give Low Permanent Strains (dd/Do) in the Range of 1.0 to 2.3 percent and Corresponding Low Offset Strains Material (Do, mm)

Sample ID No.

Offset Displacement d, mm Permanent Displacement dd, mm Permanent Strain dd/Do, %

Strain Difference (d - dd)/Do, %

15x15 Zry-4 (10.91 mm) 101B7 101B8 101B9 101B10 0.24 0.20 0.20 0.16 0.21 0.17 0.18 0.14 1.9 1.6 1.6 1.3 0.3 0.3 0.2 0.2 17x17 ZIRLOTM (9.48 mm) 109D7 109D8 109D9 109D10 0.25 0.17 0.14 0.14 0.22 0.16 0.12 0.12 2.3 1.7 1.3 1.3 0.3 0.1 0.2 0.2 17x17 M5 (9.48 mm) 636B2 636B3 636B4 0.18 0.14 0.15 0.19 0.14 0.15 2.0 1.5 1.6 0.0 0.0 0.0 For as-fabricated and pre-hydrided cladding oxidized at 1,200°C, the difference between offset and permanent displacement depends on both the oxidation level and the magnitude of the permanent displacement. For material with high ductility, the difference in displacements can be as high as 0.5 mm.

DG-1263, Revision 1, Appendix A, Page A-2 For material with essentially no ductility, both the offset and permanent displacement values are in the noise of uncertainty, and their difference can be as low as 0.01 mm.

However, what is relevant to the determination of the ductile-to-brittle transition oxidation level is the error in offset strain as determined by the difference between offset (p/Do in percent) and permanent (dp/Do in percent) strains for permanent strains in the range of 1.0 to 2.3 percent. Figure A-1 summarizes the data reported in table A-1. The data are plotted as a function of Cathcart-Pawel equivalent cladding reacted (CP-ECR). Low values of permanent strain at low CP-ECR levels (e.g., 5-10 percent) are from pre-hydrided Zircaloy-4 (Zry-4), high-burnup Zry-4, and ZIRLOTM samples. Low values of permanent strain at intermediate CP-ECR levels (10-18 percent) are from high-burnup ZIRLOTM and M5 samples. Low values of permanent strain at high CP-ECR values (15-20 percent) are from as-fabricated cladding materials. The best linear fit to the data is given by p/Do - dp/Do = 0.25 + 0.0863 CP-ECR (Equation A1)

The one-sigma upper bound to the data is given by p/Do - dp/Do = 0.41 + 0.1082 CP-ECR (Equation A2)

Because of the large data scatter in Figure A-1, the one-sigma upper bound is used to establish the offset-strain ductility criterion. It is derived by setting the permanent strain (dp/Do) in equation A2 to 1.0 percent:

p/Do 1.41 + 0. 1082 CP-ECR (A3)

For multiple offset-strain data points at the same CP-ECR level, the average value for the data set, rounded to the nearest tenth of a percent, should be used for p/Do in equation A3. Similarly, the limit calculated from the right-hand side of equation A3 should also be rounded to the nearest tenth of a percent.

DG-1263, Revision 1, Appendix A, Page A-3 Figure A-1. Difference in offset strain and permanent strain as a function of calculated oxidation level (CP-ECR) for permanent strains near the embrittlement threshold (1.0% to 2.3%) for as-fabricated, pre-hydrided, and high-burnup cladding alloys oxidized at 1,200°C and ring-compressed at RT and 135°C and at 0.033 mm/s.

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 0

2 4

6 8

10 12 14 16 18 20 22 CP-Predicted ECR (%)

Offset Minus Permanent Strain (%)

Zircaloy ZIRLO M5 Best Fit 1-Upper Bound

DG-1263, Revision 1, Appendix B, Page B-1 APPENDIX B OVERVIEW OF ACCEPTABLE TEST MATRICES Draft Regulatory Guide (DG)-1262, Testing for Post-Quench Ductility, provides a detailed test procedure that is acceptable for generating post-quench ductility data through ring compression tests. This appendix provides a simple overview of acceptable test matrices. The test matrix overviews in this appendix are consistent with the guidance in DG-1262.

This appendix provides two examples. The first series of test matrices could be used to generate ring compression test data to establish an alloy-specific limit or to establish a limit at a PCT lower than 1,204°C (2,200°F). The second series of test matrices could be used to generate ring compression test data to demonstrate consistency with the analytical limit provided in figure 2 of this regulatory guide.

B-1. Overview of Sample Test Matrix to Generate Ring Compression Test Data to Establish an Alloy-Specific Limit or to Establish a Limit at a Peak Cladding Temperature Lower than 1,204°C (2,200°F)

This guide describes how to perform post-quench ductility testing to establish an alloy-specific analytical limit or to establish an analytical limit at a peak cladding temperature lower than 1,204°C (2,200°F). Such testing should characterize a cladding alloys embrittlement behavior through the entire spectrum of conditions expected during operation. A diverse matrix of material conditions can provide a complete characterization, and repeat testing can address expected variability in oxidation behavior. The test matrix to generate ring compression test data to establish an alloy-specific limit or to establish a limit at a peak cladding temperature lower than 1,204°C (2,200°F) provided here includes testing of as-received, pre-hydrided, and irradiated material.

As-received cladding material may be used to conduct scoping tests to identify the oxidation equivalent cladding reacted (ECR) where transition behavior likely occurs. Table B-1 provides a sample test matrix for this scoping test.

Table B-1. Sample Test Matrix for Scoping Tests for As-Received Cladding Material From the scoping test, a brittle result and a ductile result will likely be identified. For example, the average of three samples at 17 percent ECR may be determined to be ductile using the ductility criterion 1.0 percent permanent strain or the offset strain criterion defined in appendix A to this regulatory guide, while the average of three samples at 20 percent ECR may be determined to be brittle.

Following the evaluation of results from the scoping tests, the next set of tests with as-received cladding material should be conducted within the range where the brittle and ductile results were observed, to identify the ECR at which the transition occurs. For example, if the average of three samples at 17 percent Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

1 2

3 Offset Strain Measurement Average of 3 RTC samples Average Offset strain Criterion?

Offset strain Criterion of 10% ECR =

2.5 Yes /

No Offset strain Criterion of 13% ECR =

2.8 Yes /

No Offset strain Criterion of 17% ECR =

3.2 Yes /

No Offset strain Criterion of 20% ECR =

3.6 Yes /

No Scoping test 10%

13%

17%

20%

DG-1263, Revision 1, Appendix B, Page B-2 ECR was determined to be ductile, while the average of three samples at 20 percent ECR was determined to be brittle, the next set of tests should be conducted at 18 percent and 19 percent ECR. In the transition region, repeat tests improve the characterization because of variability in oxidation behavior. Therefore, a sample test matrix for testing in this region includes multiple oxidation and quench tests at each oxidation level, as shown in table B-2. The transition from ductile to brittle behavior should be identified to occur at the highest Cathcart-Pawel equivalent cladding reacted (CP-ECR) at which the permanent strain is 1.0 percent.

Table B-2. Sample Test Matrix for Testing As-Received Cladding Material in the Identified Transition Region Pre-hydrided cladding material may be used to characterize the effect of hydrogen on an alloys oxidation embrittlement behavior. The entire range of a cladding materials anticipated hydrogen level should be characterized. To characterize the range of a cladding materials anticipated hydrogen content, an acceptable approach would be to determine the ductile-to-brittle transition for pre-hydrided material in increments not more than every 100 weight parts per million (wppm) of hydrogen. The test matrix in table B-4 illustrates an acceptable test matrix for a cladding material that is anticipated to have a maximum hydrogen content of 400 wppm hydrogen at end of life. The embrittlement threshold shown in figure 2 as a function of hydrogen content may be used as a guide in selecting the range of oxidation levels to be included in the test matrix. Table B-3 provides the embrittlement threshold in figure 2 in tabular form for clarity. Table B-4 shows a sample test matrix for scoping the behavior of pre-hydrided material.

Table B-3. Tabulated Values for the Embrittlement Threshold in DG-1263 Hydrogen Content (wppm)

Embrittlement ECR 0

18%

100 15.5%

200 12%

300 9%

400 6%

500 5%

600 4%

700 3%

800 2%

Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

1 2

3 1

2 3

1 2

3 Offset Strain Measurement Average of RTC samples Average Offset strain Criterion?

Yes / No Yes / No Transition Region 18%

19%

19%

18%

18%

19%

DG-1263, Revision 1, Appendix B, Page B-0 Table B-4. Sample Test Matrix for Scoping Tests for Pre-Hydrided Cladding Material Hydrogen Level (wppm)

Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

1 2

3 1

2 3

1 2

3 1

2 3

1 2

3 Offset Strain Measurement Average of RTC samples Average Offset strain Criterion?

Ductile-to-brittle transition identified?

Yes - continue tests at ECR betw een ductile and brittle level; No - conduct test additional scoping tests Yes - continue tests at ECR betw een ductile and brittle level; No - conduct test additional scoping tests Yes - continue tests at ECR betw een ductile and brittle level; No - conduct test additional scoping tests Yes - continue tests at ECR betw een ductile and brittle level; No - conduct test additional scoping tests Scoping Tests 400 1st test at 6%

If 6% was brittle, 2nd test at 4%

If 6% was ductile, 2nd test at 8%

Yes / No Yes / No 1st test at 12%

Yes / No Yes / No 300 Yes / No Yes / No If 15.5% was brittle, 2nd test at 13.5%

If 15.5% was ductile, 2nd test at 17.5%

100 Yes / No Yes / No 200 If 12% was brittle, 2nd test at 10%

If 12% was ductile, 2nd test at 14%

1st test at 9%

If 9% was brittle, 2nd test at 7%

If 9% was ductile, 2nd test at 11%

1st test at 15.5%

DG-1263, Revision 1, Appendix B, Page B-0 The objective of the scoping tests for pre-hydrided material is to identify an ECR level at which ductile behavior is observed, and an ECR level at which brittle behavior is observed, and thus identify the range in which the ductile-to-brittle behavior is observed. Once this range of ECR levels is identified, the test matrix continues with testing at an ECR level between the ECR level at which ductile behavior is observed and the ECR level at which brittle behavior is observed. Table B-5 shows a test matrix that can be used at each hydrogen level to characterize embrittlement behavior at the ECR level at which the ductile-to-brittle transition is expected to occur.

Table B-5. Sample Test Matrix for Testing Pre-Hydrided Cladding Material in the Identified Transition Region Irradiated cladding material can be used to demonstrate that a cladding alloys embrittlement behavior is accurately characterized by using pre-hydrided material. To demonstrate this, an acceptable approach would be to determine the ductile-to-brittle transition for irradiated material with hydrogen contents within 50 wppm of the anticipated maximum hydrogen content and within 100 wppm of half of the anticipated maximum hydrogen content. The test matrix below illustrates an acceptable test matrix for a cladding material that is anticipated to have a maximum hydrogen content of 400 wppm hydrogen at end of life. Table B-6 shows a sample test matrix for demonstrating that a cladding alloys embrittlement behavior is accurately characterized by using pre-hydrided material.

Table B-6. Sample Test Matrix for Testing Irradiated Cladding Material Hydrogen Level Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

Offset Strain Measurement Average of RTC samples Average Offset strain Criterion?

Yes / No Repeat for each hydrogen level Transition ECR Transition ECR Transition ECR Transition Region Hydrogen Level

(

)

Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

1 2

3 1

2 3

1 2

3 Offset Strain Measurement Average of RTC samples Average Offset strain Criterion?

Ductile-to-brittle transition comparable to pre-hydrided material?

Irradiated Testing Licensed Hydrogen Limit +/- 50 w ppm Half of Licensed Hydrogen Limit +/- 50 w ppm Yes / No Yes / No Yes / No Yes / No Yes / No Transition ECR Transition ECR + 1%

Transition ECR - 1%

Yes / No Yes / No Yes / No Transition ECR Transition ECR + 1%

Transition ECR - 1%

DG-1263, Revision 1, Appendix B, Page B-1 B-2. Overview of Sample Test Matrix to Generate Ring Compression Test Data to Demonstrate Consistency with the Analytical Limit in DG-1263 The objective of post-quench ductility testing to demonstrate consistency with the analytical limit in figure 2 of this regulatory guide is to confirm that the transition to brittle behavior does not take place at a lower ECR than the provided limit. Because of this, the matrix of material conditions and oxidation levels can be significantly reduced from the matrix outlined in the previous section. A range of material conditions can serve to provide a characterization through the spectrum of conditions expected during operation, and repeat testing can address expected variability in oxidation behavior. The test matrix provided here to generate ring compression test data to demonstrate consistency with the analytical limit given in figure 2 of this regulatory guide includes testing of as-received, pre-hydrided, and irradiated material. The transition from ductile to brittle behavior should be identified to occur at the highest CP-ECR at which the permanent strain is 1.0 percent. Consistency with the analytical limit in figure 2 of this regulatory guide would be considered to be demonstrated when the transition from ductile to brittle behavior is not lower than the provided limit.

Pre-hydrided cladding material may be used to characterize the effect of hydrogen on an alloys oxidation embrittlement behavior. The entire range of a cladding materials anticipated hydrogen level should be characterized. To characterize this range, an acceptable approach would be to determine the ductile-to-brittle transition for pre-hydrided material in increments not more than every 100 wppm of hydrogen. The test matrix in table B-7 illustrates an acceptable matrix for a cladding material that is anticipated to have a maximum hydrogen content of 400-wppm hydrogen at end of life. The analytical limit in figure 2 of this regulatory guide as a function of hydrogen content may be used as a guide in selecting the range of oxidation levels to be included in the test matrix. Table 3 of this regulatory guide provides the embrittlement threshold in figure 2 in tabular form for clarity.

Irradiated cladding material can be used to demonstrate that a cladding alloys embrittlement behavior is accurately characterized by using pre-hydrided material. To demonstrate this, an acceptable approach would be to determine the ductile-to-brittle transition for irradiated material with hydrogen contents within 50 wppm of the anticipated maximum hydrogen content and within 100 wppm of half of the anticipated maximum hydrogen content. Table B-7 illustrates an acceptable test matrix for a cladding material that is anticipated to have a maximum hydrogen content of 400 wppm at end of life.

Table B-7 provides a complete test matrix, including as-received, pre-hydrided, and irradiated material, acceptable to the NRC for using in post-quench ductility testing to demonstrate consistency with the analytical limit provided in figure 2 of this regulatory guide.

DG-1263, Revision 1, Appendix B, Page B-2 Table B-7. Sample Test Matrix for Testing As-Received, Pre-Hydrided, and Irradiated Cladding Material to Demonstrate Consistency with the Analytical Limit in Figure 2 of DG-1263 Hydrogen Level (wppm)

Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

1 2

3 1

2 3

1 2

3 Offset Strain Measurement Average of RTC samples Average Offset strain Criterion?

Ductile-to-brittle transition at or above limit of Fig.2?

Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No As-Received and Pre-hydrided Testing As-Received 100 w ppm Transition ECR from Fig.2 Transition ECR + 1%

Transition ECR - 1%

Transition ECR from Fig.2 Transition ECR + 1%

Transition ECR - 1%

Hydrogen Level (wppm)

Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

1 2

3 1

2 3

1 2

3 Offset Strain Measurement Average of RTC samples Average Offset strain Criterion?

Ductile-to-brittle transition at or above limit of Fig.2?

Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No Transition ECR - 1%

Transition ECR from Fig.2 Transition ECR + 1%

Transition ECR - 1%

As-Received and Pre-hydrided Testing 200 300 Transition ECR from Fig.2 Transition ECR + 1%

Hydrogen Level (wppm)

Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

Offset Strain Measurement Average of RTC samples Average Offset strain Criterion?

Ductile-to-brittle transition at or above limit of Fig.2?

Yes / No Yes / No As-Received and Pre-hydrided Testing 400 Transition ECR from Fig.2 Transition ECR + 1%

Transition ECR - 1%

Yes / No Yes / No

DG-1263, Revision 1, Appendix B, Page B-3 Hydrogen Level (wppm)

Oxidation Level (ECR)

Ring compression Sample 1

2 3

1 2

3 1

2 3

1 2

3 1

2 3

1 2

3 Offset Strain Measurement Average of RTC samples Average Offset strain Criterion?

Ductile-to-brittle transition at or above limit of Fig.2?

Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No Yes / No Transition ECR - 1%

Irradiated Material Testing Licensed Hydrogen Limit +/- 50 w ppm Half of Licensed Hydrogen Limit +/- 50 w ppm Transition ECR from Fig.2 Transition ECR + 1%

Transition ECR - 1%

Transition ECR from Fig.2 Transition ECR + 1%