ML24320A013

DG-1434 (RG 1.259 Rev 0) Addressing the Consequences of Fuel Dispersal in LWR Loss-of-Coolant Accidents - ACRS Version

ML24320A013
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Issue date:	11/21/2024
From:	James Corson NRC/RES/DSA/FSCB
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DG-1434 RG 1.259, Rev 0
Download: ML24320A013 (23)

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U.S. NUCLEAR REGULATORY COMMISSION DRAFT REGULATORY GUIDE DG-1434, Revision 0 Proposed new Regulatory Guide 1.259 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-1434. 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. ML24304A951. The regulatory analysis is associated with a rulemaking and may be found in ADAMS under Accession No. ML24239A776.

ADDRESSING THE CONSEQUENCES OF FUEL DISPERSAL IN LIGHT-WATER REACTOR LOSS-OF-COOLANT ACCIDENTS A. INTRODUCTION Purpose This regulatory guide (RG) describes an approach that is acceptable to the staff of the U.S.

Nuclear Regulatory Commission (NRC) to meet regulatory requirements related to fuel dispersal during light water reactor (LWR) loss-of-coolant accidents (LOCAs). It describes analytical limits and guidance for determining the amount of fuel susceptible to fine fragmentation for the full spectrum of postulated LOCA events. In addition, for LOCAs with break sizes greater than the transition break size (TBS) defined in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46a, Alternative acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, this RG identifies the amount of fuel that could be dispersed to the reactor coolant system (RCS) and containment in the event of cladding failure and identifies phenomena that should be addressed if fuel dispersal is predicted. In consideration of the extremely low frequency of LOCAs beyond the TBS expected for plants demonstrating compliance with 10 CFR 50.46a, fuel dispersal associated with such breaks may be assessed in a beyond-design-basis framework with relaxed analytical criteria.

Applicability This RG applies to applicants for and holders of construction permits and operating licenses for power reactors under 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 applications 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.

DG-1434, Revision 0, Page 2 o General Design Criterion (GDC) 14, Reactor coolant pressure boundary, requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

o GDC 16, Containment design, requires that reactor containment and associated systems be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to ensure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

o GDC 27, Combined reactivity control systems capability, requires that the reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system (ECCS), of reliably controlling reactivity changes to ensure that the capability to cool the core is maintained, under postulated accident conditions and with appropriate margin for stuck rods.

o GDC 28, Reactivity limits, requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures, or other reactor pressure vessel internals to impair significantly the capability to cool the core.

o GDC 34, Residual heat removal, requires that a system to remove residual heat be provided. The system safety function must be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

o GDC 35, Emergency core cooling, requires that a system to provide abundant emergency core cooling be provided. The system safety function must be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

10 CFR 50.46a provides a voluntary alternative to 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. Licensees that receive NRC approval to implement 10 CFR 50.46a are required to ensure the fuel remains in a coolable geometry and provide sufficient coolant to remove long-term decay heat.

Additionally, licensees are required to have NRC-approved limits that address cladding degradation phenomena, maintain fuel coolability, avoid explosive concentrations of hydrogen gas, and demonstrate long-term decay heat removal.

o 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, defines requirements for electrical systems important to safety, including post-accident monitoring systems, and the environmental conditions, including radiation, that must be considered in their qualification.

o 10 CFR 50.67, Accident source term, defines the regulatory limits on radiological consequence.

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

DG-1434, Revision 0, Page 3 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 of 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 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants (Ref. 4), provides guidance related to environmental qualification of equipment important to safety and includes consideration of the effects of postulated accidents.

RG 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance (Ref. 5), provides guidance for calculating realistic or best-estimate ECCS performance during LOCAs.

RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (Ref. 6), provides guidance for calculating radiological consequences for design-basis accidents.

o Draft Regulatory Guide (DG)-1425 (proposed Revision 2 of RG 1.183) (Ref. 7), includes methods for estimating the impact of fuel dispersal on radiological consequences.

RG 1.203, Transient and Accident Analysis Methods (Ref. 8), describes a process that the staff considers acceptable for use in developing and assessing evaluation models that may be used to analyze transients and accidents considered within the safety analysis of a nuclear power plant.

DG-1263, Revision 1, (proposed new RG 1.224), Establishing Analytical Limits for Zirconium-Based Alloy Cladding (Ref. 9), provides analytical limits to address cladding degradation phenomena and avoid explosive concentrations of hydrogen gas during a LOCA. Note that this RG focuses on ductile failure leading to fuel dispersal, which would generally tend to occur before reaching failure thresholds associated with the embrittlement mechanisms described in DG-1263, Revision 1.

DG-1434, Revision 0, Page 4 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 request for information or an information collection requirement unless the requesting document displays a valid OMB control number.

DG-1434, Revision 0, Page 5 TABLE OF CONTENTS A. INTRODUCTION.................................................................................................................................. 1 B. DISCUSSION......................................................................................................................................... 6 C. STAFF REGULATORY GUIDANCE................................................................................................. 10 D. IMPLEMENTATION........................................................................................................................... 20 REFERENCES........................................................................................................................................... 21

DG-1434, Revision 0, Page 6 B. DISCUSSION Reason for Issuance This guide describes acceptable analytical methods, assumptions, and limits for (1) determining the quantity of fuel susceptible to fine fragmentation for the full spectrum of postulated LOCAs and (2) addressing the impacts of fuel dispersal from high burnup fuel rods during a LOCA for LOCAs with break sizes exceeding the TBS defined in 10 CFR 50.46a. It is based on available experimental data from in-pile and out-of-pile tests simulating fuel rod behavior during a LOCA and on applicable analyses from several publications that examine fuel rod and reactor system performance under LOCA conditions.

Background

Tests performed on irradiated fuel rod segments have shown that under LOCA conditions, high-burnup uranium dioxide (UO2) fuel can finely fragment, relocate axially within the fuel rod, and disperse to RCS following the ballooning and burst of zirconium-alloy cladding, as described in NUREG-2121, Fuel Fragmentation, Relocation, and Dispersal during the Loss-of-Coolant Accident (Ref. 10);

NUREG-2160, Post-Test Examination Results from Integral, High-Burnup, Fueled LOCA Tests at Studsvik Nuclear Laboratory (Ref. 11); and a report on Fuel Fragmentation, Relocation, and Dispersal written by the Organisation for Economic Cooperation and Developments Nuclear Energy Agencys Working Group on Fuel Safety (WGFS) (Ref. 12). Collectively, these phenomena are commonly referred to as FFRD.

Research Information Letter (RIL) 2021-13, Interpretation of Research on Fuel Fragmentation, Relocation, and Dispersal at High Burnup, issued December 2021 (Ref. 13), discusses FFRD research that had been conducted at the time of its publication. The RIL provides evidence showing that fine fragmentation begins at 55 gigawatt days per metric ton of uranium (GWd/MTU) pellet-average burnup for standard UO2 fuel, and the fragment size distribution becomes progressively finer at higher burnups.

In addition to this burnup threshold, experimental evidence demonstrates that a minimum threshold cladding diametral strain of 3 percent is necessary for fuel pellets to fragment and relocate axially within the rod. Transient fission gas release (tFGR) resulting from fuel microcracking and pulverization contributes to the pressure differential between the rod and the coolant, which can exacerbate cladding stresses and contribute to ballooning and burst behavior. When the cladding of susceptible fuel rods bursts, pulverized fuel can be expelled to the coolant.

RIL 2021-13 focuses on FFRD and tFGR under postulated LOCA conditions where fuel rod cladding ballooning and rupture is expected to occur at elevated temperatures. The loss of mechanical constraint on the high-burnup fuel pellet (during ballooning) and subsequent rapid depressurization of the fuel rod upon rupture are primary drivers of FFRD and tFGR. Due to the rapid core-wide loss of liquid coolant and system depressurization, large primary piping breaks appear to be the postulated accident sequence with the greatest potential to generate a large population of damaged fuel rods, including fuel rods that experience cladding ballooning and rupture.

When considering the potential for fuel dispersal for LOCA break sizes exceeding the TBS in accordance with 10 CFR 50.46a, NRC regulations, including the GDC, require applicants to demonstrate that safety-related systems, structures, and components are capable of satisfying the following fundamental regulatory design and performance objectives during and following postulated design basis accidents:

Achieve and maintain safe shutdown of the reactor.

DG-1434, Revision 0, Page 7 Maintain a fuel geometry amenable to continued effective cooling and capability to remove long-term decay heat.1 Maintain radiological consequences within regulatory limits by reducing the release of radioactivity to the environment.

Together with the development of verified and validated analytical models, analytical limits should be established to demonstrate compliance with these regulatory design and performance objectives. Some recommended analytical limits follow directly from the above performance objectives.

Examples include the following:

Following reactor trip, maintain the reactor core subcritical (i.e., keff < 1.0).

Demonstrate that fuel temperatures progressively decrease until cold shutdown conditions are reached and maintained to demonstrate effective long-term cooling.

Limit public radiation exposure consistent with established dose limits (e.g., 25 rem total effective dose equivalent).

However, an analytical limit for maintaining fuel in a geometry amenable to continued effective cooling is not straightforward, particularly when considering the potential for dispersal to result in significant changes in configuration. Historically, separate-effects and integral tests conducted on both fresh and irradiated fuel rod segments, as part of laboratory, hot-cell, and in-pile research programs, have been used to establish analytical limits that preserve a core geometry deemed amenable to continued effective cooling. Due to the empirical nature of demonstrating adequate coolability, these analytical limits may, in general, be fuel-and reactor-design specific.

Fuel and cladding degradation and deformation mechanisms considered in assessing a coolable core geometry include multiple phenomena that do one or both of the following:

challenge the claddings ability to retain fission products (for assessing fuel damage and radiological source term) challenge fuel assembly components abilities to maintain as-designed dimensions, configuration, and functional capabilities.

Not all core geometric changes are prohibited or unexpected. For example, fuel rod ballooning and rupture may occur under LOCA conditions. Geometric changes due to ballooning and rupture are not precluded by regulation; however, the potential impacts of such changes (e.g., to local thermal-hydraulic conditions, cladding wall thickness) must be included in the ECCS performance demonstration to ensure adequate core cooling is maintained.

Other core geometric changes have been deemed unacceptable, and performance metrics and analytical limits have been established to preclude this behavior under postulated design-basis-accident conditions. One example of unacceptable changes in core configuration is the loss of fuel bundle array geometry under postulated LOCA conditions. For LWRs fueled with UO2 pellets within certain types of zirconium-alloy cladding, 10 CFR 50.46(b) provides prescriptive analytical limits on calculated peak 1

Both 10 CFR 50.46 and GDC 35 refer to core cooling, while 10 CFR 50.46 also mentions core geometry. The alternative ECCS acceptance criteria in 10 CFR 50.46a recognize that fuel dispersal is possible and refer more broadly to ensuring that the fuel remains coolable, whether it remains in the core or is dispersed elsewhere.

DG-1434, Revision 0, Page 8 cladding temperature, maximum local oxidation, and hydrogen generation, as well as more general requirements that changes in core geometry must be such that the core remains amenable to cooling, and that the core temperature must be maintained at an acceptably low value and decay heat must be removed for the extended period of time required by the long-lived radioactivity remaining in the core. Similarly, 10 CFR 50.46a(f) requires the development of NRC-approved limits that address cladding degradation phenomena. DG-1263, Revision 1, includes acceptable limits for complying with 10 CFR 50.46a(f) for UO2 fuel in zirconium-based alloy cladding.

The prescribed analytical limits of 2,200 degrees Fahrenheit (°F) (1,204 degrees Celsius (C))

peak cladding temperature (PCT) and 17 percent equivalent cladding reacted maximum local oxidation (MLO) in 10 CFR 50.46(b), as well as the related limits on PCT and integral time at temperature in DG-1263, Revision 1, were developed to retain post-quench ductility in zirconium-alloy cladding. Retaining this favorable material property (i.e., ductility) would provide reasonable assurance that fuel rods would not experience catastrophic failure as a result of thermal shock upon quench (i.e., core reflood). The catastrophic failure of a significant number of fuel rods in this manner could disrupt the core configuration, making it difficult to demonstrate compliance with the regulatory requirement to prevent fuel and clad damage that could interfere with continued effective core cooling.

The fragmentation and dispersal of any FFRD-susceptible fuel introduces new phenomena and technical complexities that may need to be addressed to demonstrate compliance with the requirement to maintain fuel (i.e., including dispersed fuel) in a coolable geometry and other applicable regulations. New or different performance metrics and analytical limits may need to be established to continue satisfying the fundamental regulatory design and performance objectives. Furthermore, addressing these phenomena may require sufficient empirical data from separate-effects and integral testing on irradiated fuel rod segments to understand sensitivities and uncertainties, and to validate analytical models capable of accurately or conservatively modeling relevant phenomena. The following gives examples of new engineering calculations that may be appropriate for assessing the potential consequences of fuel dispersal:

quantification of fragmentation-induced transient fission product releases quantification of fragmented and dispersed fuel particles size distribution of fragmented and dispersed fuel particles interaction of fuel particles with coolant and structures transport and deposition of fuel particles criticality of dispersed and deposited fuel particles long-term decay heat removal from dispersed and deposited fuel particles fuel particle oxidation, stability, and fission product releases equipment qualification (e.g., cables, instruments) impact of fuel particles on safety-related systems (e.g., residual heat removal system, sump, recirculation strainers)

To gain a better understanding of the importance and state of knowledge of phenomena that would impact the above calculations, the NRC sponsored a (PIRT) exercise on fuel dispersal, discussed in

DG-1434, Revision 0, Page 9 NUREG/CR-7307, Phenomena Identification and Ranking Tables on High Burnup Fuel Fragmentation, Relocation, Dispersal, and Its Consequences for Design-Basis Accidents in Pressurized-and Boiling-Water Reactors, issued May 2024 (Ref. 14). The expert PIRT panelists noted that to demonstrate that fuel dispersal does not negatively impact the fundamental regulatory design and performance objectives listed above, one must first understand the maximum amount of fuel that could be dispersed for a given reactor design and core loading pattern. If the mass of fuel is relatively small, the panel concluded that it should be possible to perform simple calculations to estimate the impact of fuel dispersal on the accident progression. However, the panel acknowledged that many important parameters are not well understood and are subject to large uncertainties, which makes it difficult to provide more detailed guidance on demonstrating compliance with core coolability and long-term cooling requirements given in 10 CFR 50.46 and 10 CFR 50.46a, reactivity-control requirements in GDC 27, and equipment qualification requirements in 10 CFR 50.49.

The most straightforward way to demonstrate compliance with the underlying regulatory requirements is to show that cladding integrity is maintained during a LOCA for fuel rods susceptible to fine fragmentation. For existing core and fuel designs, such a demonstration may be more readily accomplished for smaller break sizes, for which peak cladding temperatures are typically lower and the RCS pressure remains elevated as compared to large-break LOCA scenarios, where the RCS pressure rapidly decreases to a value somewhere between atmospheric pressure and the containment design pressure. With that said, it may also be possible to show that high burnup fuel rods do not burst in a large-break LOCA, particularly for plants that adopt the TBS approach in 10 CFR 50.46a that allows for a more realistic analysis of the accident progression than the 10 CFR Part 50, Appendix K, ECCS Evaluation Models, or best-estimate-plus-uncertainty approaches required by 10 CFR 50.46.

The Staff Regulatory Guidance section below provides further discussion on these issues.

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. 15) and Management Directive and Handbook 6.6, Regulatory Guides (Ref. 16).

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. 17),

states that the core should be designed such that it remains coolable under accident conditions.

IAEA SSG-56, Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants, issued in 2020 (Ref. 18), includes high-level information about ECCS requirements for LOCAs.

DG-1434, Revision 0, Page 10 C. STAFF REGULATORY GUIDANCE This section contains regulatory positions that establish methods acceptable to the NRC staff for addressing the impact of high burnup fuel dispersal during LOCAs on the coolability and long-term cooling requirements of 10 CFR 50.46a.

1.

Limits on Applicability

a.

The FFRD thresholds (section C.2) and the methods for estimating the mass of dispersed fuel (section C.4) apply to undoped UO2 fuel in zirconium-alloy cladding. Justification for the applicability of the FFRD thresholds and dispersed fuel methods in sections C.2 and C.4 to mixed-oxide (MOX) fuel or UO2 fuel containing dopants (e.g., gadolinia, chromia, alumina, and/or silica) will be considered on a case-by-case basis.

b.

The methods for addressing the impacts of fuel dispersal are applicable to all UO2 or MOX fuel designs, including fuels with dopants, unless otherwise noted.

2.

Fuel Fragmentation, Relocation, and Dispersal Thresholds

a.

UO2 fuel becomes increasingly susceptible to fine fragmentation above a pellet-average burnup of 55 GWd/MTU, as discussed in RIL 2021-13.

b.

Fuel axial relocation during simulated LOCA tests has only been observed above a certain cladding strain threshold of approximately 3 percent, as discussed in RIL 2021-

13.

c.

Fuel dispersal under LOCA conditions requires fuel rod cladding failure due to ballooning and burst.

d.

Given the thresholds defined above, fuel rods with peak pellet-average burnups above 55 GWd/MTU that balloon and burst during a LOCA are susceptible to fuel dispersal.

3.

Analytical Limits for Fuel Dispersal

a.

LOCAs with breaks below the TBS (1)

Accident analyses should demonstrate that there is no fuel dispersal.

i.

This can be accomplished by demonstrating that the cladding does not balloon and burst for rods containing fuel above the fine fragmentation threshold defined in position C.2.a.

ii.

Cladding ballooning and burst should be determined using NRC-approved analytical models and methods.

iii.

The evaluation model should consider the impact of tFGR on fuel rod behavior. Section C.4 discusses this topic in more detail.

b.

LOCAs with break sizes above the TBS (1)

The simplest approach to addressing the consequences of fuel dispersal is to demonstrate that there is no fuel dispersal.

DG-1434, Revision 0, Page 11

i.

Again, this can be accomplished by demonstrating that the cladding does not balloon and burst for rods containing fuel above the fine fragmentation threshold defined in position C.2.a.

ii.

Cladding ballooning and burst should be determined using NRC-approved analytical models and methods. Note that 10 CFR 50.46a allows for the use of more realistic assumptions in the accident analysis methodologies. This may make it easier to demonstrate that there is no dispersal even for breaks above the TBS. The evaluation should still consider the impact of transient fission gas release on fuel rod behavior.

(2)

Alternatively, an applicant may propose an acceptable limit on the mass of fuel dispersed for LOCAs with break sizes above the TBS.

i.

In doing so, the applicant should apply the methods in section C.4 to determine the mass of fuel dispersed to the RCS.

ii.

The proposed limit on the mass of fuel dispersed should address the positions in section C.5.

4.

Methods for Estimating the Dispersed Fuel Mass The amount of fuel that could be dispersed is governed by several factors, including the rod internal pressure at the time of failure, the burst opening size and shape, and the fuel particle size distribution. While existing fuel performance and accident analysis models can predict some of these parameters (e.g., the rod internal pressure, cladding failure timing) with reasonable accuracy, other parameters are less well understood and subject to considerable uncertainty. In particular, the expert panel convened for the fuel dispersal PIRT identified the particle size relative to the burst opening size as a parameter with a high significance, low state of knowledge, and high uncertainty, as discussed in NUREG/CR-7307. This highlights the difficulty in developing a mechanistic model to predict the amount of dispersed fuel.

Given these difficulties in developing a mechanistic model, the NRC staff proposed six empirical fuel dispersal models in Appendix A, A Model for Predicting Dispersal, of RIL 2021-13. The NRC staff then applied those models to the measured masses of dispersed fuel and mobile fuel from seven integral LOCA tests performed as part of the third phase of the Studsvik Cladding Integrity Project (SCIP-III). (The mass of mobile fuel was defined as the amount of fuel that was collected after breaking the fuel rod and shaking out the loose fuel after the LOCA test.)

Two of the models deserve further discussion. Model A is derived under the assumption that all fuel fragments smaller than 1 millimeter (mm) in the length of the rod with greater than 3 percent hoop strain would disperse. The mass of fuel fragments smaller than 1 mm was based on available data from tests performed at Studsvik Nuclear Laboratory in Sweden, at the Halden Boiling Water Reactor in Norway, and at the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory:

mass fraction =

0, BU < 55 0.04BU 55, 55 < BU < 80 1, BU > 80 Equation 1 Model C assumes that all fuel in the length of the rod with greater than 3percent hoop strain would disperse:

DG-1434, Revision 0, Page 12 mass fraction = 0, BU < 55 1, BU > 55 Equation 2 In these equations, BU is the pellet-average fuel burnup in GWd/MTU. In light of significant uncertainties concerning, among other things, the size of cladding rupture openings, this RG uses Model A simply to determine a mass of dispersed fuel that is consistent with applicable test data, irrespective of the actual particle size distribution.

Compared to the available data from SCIP-III, Model A overpredicts the mass of fuel dispersed for four of seven tests and slightly underpredicts the dispersed mass for the remaining tests, while Model C overpredicts the dispersed fuel mass for all cases. (In most cases, Model C overpredicts the dispersed mass by more than a factor of 2.) However, as noted previously, the amount of fuel that disperses during a test is governed by a number of factors, some of which may vary stochastically (such as the burst opening size). Furthermore, the test conditions may not account for all the conditions that may be present during a LOCA at a commercial power reactor. For these reasons, in RIL 2021-13, the NRC argued that it is important to compare the models to the amount of mobile fuel in the rod, since some of the mobile material that remained in the rod following the test could have been dispersed had the burst opening size been larger, or had the rod been subjected to forces greater than those experienced in the test. Compared to the available data, Model A underpredicts the amount of mobile fuel, while Model C was fairly accurate in its predictions of the mass of mobile fuel. Ultimately, in the RIL, the NRC recommended Model C as a sufficiently conservative model for fuel dispersal.

Since that time, additional research has been performed to study fuel dispersal. Tests with surrogate particles have shown that dispersal from the rod occurs immediately following burst and that little additional material is released when the rod is shaken to simulate potential forces on the rod during a LOCA (Ref. 19). This behavior was observed even when the burst opening width was greater than the largest particle size. High burnup UO2 fuel would likely exhibit similar behavior in post-burst shaking tests, though work is ongoing at Oak Ridge National Laboratory on this topic. Such conclusions suggest that the blowdown of gases following burst is the primary force leading to dispersal and that post-burst shaking of the rod would have little impact on the amount of fuel released.

Additional research has also contributed to a better understanding of the burst opening characteristics, which play an important role on the amount of fuel dispersed, as discussed in NUREG/CR-7307. Notably, Capps and Sweet (Ref. 20) analyzed publicly available data on cladding burst tests and developed models correlating the burst opening length with peak cladding hoop strain and correlating burst width with burst length. Historical models from NUREG-0630, Cladding Swelling and Rupture Models for LOCA Analysis, issued April 1980 (Ref. 21), suggest peak cladding hoop strains for high burnup rods would range from approximately 20 percent to 90 percent, assuming cladding burst in the 700-800°C temperature range. (Note that the burst temperature would depend significantly on the fuel and cladding operating characteristics, so the range here is only a crude estimate for demonstration purposes.) Applying this hoop strain range, the limiting Capps and Sweet models would predict a burst length ranging from roughly 15-35 mm and a burst opening width ranging from 70-100 percent of the cladding outer diameter. (Cladding outer diameter is approximately 10 mm for LWR fuel rods). In comparison, the average Capps and Sweet models would predict a burst length ranging from roughly 10-20 mm and burst opening width ranging from 30-60 percent of the cladding outer diameter. In other words, the limiting models would predict large burst opening sizes that could result in significant dispersal, more in line with the Model C dispersal model in RIL 2021-13, while the smaller burst opening predicted by the average model would presumably result in more modest dispersal more in line with Model A from the RIL. Note that the historical cladding hoop strain models in NUREG-0630 are meant to be conservative models, there is significant scatter in the strain data reported in NUREG-0630, and few data are available for burst at lower temperature (below about 725°C). Applying an average correlation

DG-1434, Revision 0, Page 13 for peak cladding hoop strain as a function of burst temperature would decrease the burst opening length and width predicted by the Capps and Sweet model even further.

Based on the preceding discussion, the NRC staff believes that, while Model C in RIL 2021-13 provides a conservative model with margin to account for existing uncertainties regarding fuel dispersal, Model A in the RIL likely provides a better point estimate of the mass of fuel that could be dispersed during a LOCA. Model A is more consistent with the philosophy in 10 CFR 50.46a that breaks above the TBS can be analyzed using more realistic assumptions and methods without allocating margin for uncertainties. Therefore, the staff finds that Model A (equation 1 above) can be used to estimate the mass of fuel dispersed during a beyond-TBS LOCA, combined with core physics calculations of the fuel rod axial burnup profile, thermal-hydraulics calculations to determine the fuel rod boundary conditions during a LOCA, and fuel performance models to calculate the cladding burst behavior and axial strain profile, as demonstrated by NRC staff from the Office of Nuclear Regulatory Research in NRCs methodology to estimate fuel dispersal during a large break loss of coolant accident (Ref. 22), which was published in Nuclear Engineering and Design.2 Note that the NRC staff calculations described in Ref. 22 were only meant for demonstration purposes, not to provide a definitive estimate of the mass of fuel that would be dispersed during a LOCA for any operating reactors.

Many fuel performance codes do not model the impact of fuel assembly spacer grids on fuel rod behavior. However, evidence suggests that spacer grids can significantly impact the local cladding strain due to the combined effects of lower local power and cladding neutron fluence resulting from parasitic absorption of neutrons by the spacer grid material, lower temperature due to improved heat transfer around the grids, and mechanical restraint provided by the grids. For example, the IFA-650.14 test performed in the Halden Boiling Water Reactor showed a depression in the cladding strain profile where a spacer grid had been when the rod had been irradiated in a commercial nuclear power plant (Ref. 23).

Note that the grid was no longer present during the test, so the observed behavior is not a result of mechanical restraint. Furthermore, analysis performed with the BISON fuel performance code (Ref. 24) suggests that the spacer grids may limit strain to below the 3percent threshold identified in RIL 2021-13.

For this reason, it is reasonable to assume that fuel dispersal would generally be limited to a single grid span.

Existing LOCA methodologies already account for the impacts of fission gas release during normal reactor operating conditions, but experiments have shown that additional burst releases of fission gas may occur under simulated LOCA conditions. This tFGR will increase the rod internal pressure and may impact the cladding ballooning and burst behavior.

RIL 2021-13 also discusses the impact of tFGR under LOCA conditions. The tFGR results summarized in the RIL ranged from near zero to about 18 percent, based on fuel samples ranging from about 50 GWd/MTU to 75 GWd/MTU segment-average burnup. Generally, tFGR increased with burnup, but there is considerable scatter in the data. However, many of the tFGR tests reported in the RIL were conducted with peak temperatures between 1,000 and 1,200°C, which is significantly above the expected burst temperature for high-burnup fuel (roughly 600-800°C, although it depends on the fuel operating history). Temperature significantly impacts fission gas release in general and burst release more specifically. For example, in the French GASPARD program, fuel at a pellet-average burnup of approximately 72 GWd/MTU showed two burst release regimes: an initial burst release at lower temperature (approximately 600-800°C) and a larger burst release at temperatures above 1,000°C (Ref. 25). Most likely, the second burst release would occur after cladding burst, and hence the tFGR data 2

As discussed further below, the NRC staff's recommendation of Model A for estimating the dispersed mass of fuel for LOCAs beyond the TBS does not constitute acceptance of an assumption that all dispersed fuel particles would have a diameter of 1 mm or less.

DG-1434, Revision 0, Page 14 points reported in the RIL overstate the importance of tFGR on ballooning and burst behavior.

Nevertheless, the available data do show that some tFGR occurs at and below the temperatures associated with cladding burst, and data also show that burst releases occur at lower temperatures as burnup increases (Refs. 25 - 26). Licensees need to consider these effects when determining whether high-burnup rods fail during a LOCA.

Refs. 26 - 27 describe models that were developed to calculate tFGR. However, these models require additional validation before they can be relied upon in fuel performance or systems codes used for LOCA safety analysis.

To summarize, the staff finds the following methods acceptable for estimating the mass of fuel dispersed during a LOCA with break size larger than the TBS.

a.

For all fuel rods that are predicted to balloon and burst, the mass of fuel that disperses from that rod should be calculated using equation 1.

b.

The mass of fuel susceptible to dispersal should be limited to the grid span containing the burst location.

c.

The calculation should be performed using the average fuel burnup in the grid span containing the burst. If more detailed burnup information is available, the calculation can consider the pellet-average burnup as a function of axial position within the grid span.

5.

Impacts of Fuel Dispersal

a.

Fuel particle transport and deposition If fuel particles are released from the rod following ballooning and burst, they may be carried by the fluid flowing through the core to other locations within the RCS and, eventually, the containment. This transport will be governed by characteristics of the particles and of the flow, which may in general be two phase (i.e., vapor and liquid).

The PIRT exercise documented in NUREG/CR-7307 discusses important parameters that impact the transport of fuel particles through the RCS. The expert panelists concluded that the thermal-hydraulic response of the reactor to a LOCA is well understood and that methods exist to predict the transport of regularly shaped particles (e.g., spheroids) in single-phase turbulent flow.

However, much less is known about the transport of irregularly shaped particles (such as UO2 fragments) in a two-phase steam-water mixture. The situation is complicated by the presence of structures like spacer grids and reactor vessel internals that will impact the multiphase steam and water-droplet flows and particle transport and deposition. Furthermore, as noted previously, there are large uncertainties about the mass of fuel that may be released and the size distribution of the released particles, which in turn would have an important effect on particle transport. Finally, it is possible that the size distribution of dispersed fuel particles may change during the transport process, either due to further fragmentation of entrained or suspended fuel particles, or due to agglomeration with other debris particles. For these reasons, detailed tracking of the transport of heated fuel particles through the reactor vessel, RCS, and containment is beyond the capabilities of current reactor systems or subchannel thermal-hydraulics codes. Significant efforts would be required to develop and validate appropriate models.

Instead, the expert panelists suggested that coolability concerns could be addressed through simplified, bounding calculations, using engineering judgment about where particles may deposit. One potential location for particle accumulation is on the spacer grids, either directly

DG-1434, Revision 0, Page 15 below or above the burst location: larger particles could fall and collect on a lower grid, while smaller particles could be carried upward and become trapped in the grids above the burst location. Again, predicting how much material would be trapped is beyond current modeling capabilities. Instead, it would be appropriate to assume that all released material would collect at the grid below the burst and to assess the impact of the resulting debris bed on coolability, as discussed in the next subsection.

However, the spacer grids are not the only location at which fuel debris could collect.

Other potential locations to which fuel fragments could transport and deposit include the assembly inlet debris filters, the core support plate, the lower head, the upper core plate, the upper plenum, RCS piping, and containment. Because the two-phase flow conditions (e.g., quality, phase velocities) are expected to change substantially during the LOCA event, certain transport and deposition mechanisms may only be active during certain phases of the event (e.g.,

blowdown, refill, reflood, long-term cooling). Note that fuel particles would need to navigate multiple obstacles to reach the above locations, so it is likely that only a fraction of the particles released from the fuel could reach any given location. Nevertheless, without reliable tools for predicting fuel particle transport, some engineering judgment is required to estimate how much material would be transported far from the burst location to confirm that the dispersed fuel would not compromise coolability, long-term cooling, or safety equipment performance, or lead to other undesirable outcomes, as discussed below.

Regardless of where the particles end up, the resulting debris bed should be assumed to include particles ranging in size from about 4 mm to less than 0.125 mm (i.e., the smallest sieve size used in the NRC-sponsored tests at Studsvik and in the SCIP-III tests, as discussed in NUREG-2121, NUREG-2160, and RIL 2021-13). The resulting porosity of the bed is unknown but would likely range from about 20 percent to 40 percent, as described in the WGFS report on FFRD (Ref. 12). The lower end of this range is based on measurements of the packing fraction of fuel particles in the lower portion of the balloon region in rods tested in the SCIP-III program, as described in RIL 2021-13, while the upper end of the range represents loose random packing of equally sized spheres.

b.

Fuel coolability and long-term cooling One of the primary concerns with fuel dispersal is that it could lead to a situation that compromises the coolability and long-term cooling of the fuel. As noted previously, the existing limits on PCT and MLO in 10 CFR 50.46 and the proposed PCT and time at temperature limits in DG-1263, Revision 1, are meant to prevent the possibility of zircaloy becoming embrittled by steam oxidation and shattering during quench, allowing the fuel pellets to fall into a heap, as described in the 1973 U.S. Atomic Energy Commission opinion paper in the matter of the rulemaking hearing on acceptance criteria for ECCSs for light-water cooled nuclear power reactors (Ref. 28). Dispersed fuel may impact calculated PCT, MLO, and cladding time-at-temperature enough to challenge existing and proposed regulatory limits, potentially leading to more widespread fuel and cladding damage. Substantial amounts of fuel dispersal may also lead to the situation that the cladding embrittlement limits are meant to prevent, i.e., a heap [of uranium dioxide fuel pellets] that would be difficult to cool (Ref. 28). Finally, fuel dispersal may create flow blockages that challenge cooling, similar to the concern expressed in Ref. 28 about cladding ballooning blocking flow.

As noted previously, predicting the transport of fuel particles during a LOCA is beyond the capabilities of existing reactor systems codes, but simplified analyses based on engineering judgment may be useful in addressing the impact of dispersal on coolability and long-term

DG-1434, Revision 0, Page 16 cooling. Experts from the WGFS, as documented in its report on FFRD (Ref. 12), and from the NRC-sponsored PIRT panel, in NUREG/CR-7307, identified spacer grids as a potential location for debris accumulation. The fuel coolability scenario described in the WGFS report on FFRD considered a bed of fuel particles 10 centimeters in height accumulating in the space between fuel rods. In plant-specific applications, the debris bed height would depend on various factors, including the amount of fuel released from burst rods, the assumed porosity of the bed (likely in the 20-40 percent range, as noted previously), and the ratio of the flow cross-sectional area to the fuel cross-sectional area.

Explicitly analyzing this scenario is challenging, even if the debris bed characteristics (e.g., bed height, porosity, and particle size) are set. It involves flow through a porous medium (i.e., the bed) within an array of cylindrical fuel rods, with heat generated both within the bed and within the cylindrical fuel rods due to decay heat from dispersed particles and from fuel remaining within the rod, respectively. The expert panelists from the NRC-sponsored PIRT exercise noted in NUREG/CR-7307 that it may be possible to address this situation using a subchannel thermal-hydraulics code coupled to a fuel performance code, but more work would be needed to develop and validate such a tool to reduce uncertainties and provide more confidence in the result.

The WGFS report on FFRD mentions a simpler approach of addressing coolability based on calculating the dryout heat flux of a debris bed using the 0-D Lipinski model. This model shows that the dryout heat flux decreases (i.e., keeping the bed cool becomes more challenging) for increasing bed height, decreasing bed porosity, and decreasing particle size. (See figures 8.1-1 and 8.1-2 of the WGFS report on FFRD.) The 0-D Lipinski model, described in Swedish Nuclear Power Inspectorate (SKI) Report 02:17, A Review of Dryout Heat Fluxes and Coolability of Particle Beds (Ref. 29), compares well to experimental data from tests meant to simulate the behavior of debris beds formed in nuclear reactor severe accident scenarios. While the experiments differ from the situation of interest in this RGfor example, the severe accident experiments often involve either a uniform or stratified bed with top-cooling only, whereas debris beds formed on spacer grids would be impacted by the cylindrical rods through the bed and may involve both top and bottom coolingthe range of particle sizes and bed porosities in the experiments overlaps with the expected characteristics of debris beds that are postulated to form on spacer grids in the WGFS report. Therefore, the 0-D Lipinski correlation should give a reasonable estimate of whether a debris bed would be coolable, although it does not address debris bed quenching or the impact of the debris on the surrounding cladding before the bed is quenched.

The NRC-sponsored PIRT (NUREG/CR-7307) identified deposition of coagulants as a phenomenon with high importance and low knowledge level that impacts coolability. These coagulants would change the porosity of the debris bed, potentially lowering the dryout heat flux and making it more difficult to keep the debris bed cool. The WGFS report on FFRD considered this impact, including a limiting situation in which the bed porosity approaches zero and the bed is cooled by conduction only. The effects of the deposition of coagulants and other suspended post-LOCA debris should be considered when addressing long-term cooling for dispersed fuel.

Of course, it is possible that fuel particles would end up in other parts of the RCS or in the containment. The resulting situation would more closely resemble the conditions considered in the severe accident debris bed experiments and in the 0-D Lipinski correlation. Thus, it is reasonable to assess coolability and long-term cooling using the 0-D Lipinski correlation for debris beds that form in the RCS or containment. Once again, the challenge is in defining bounding scenarios in terms of the mass of the debris bed.

DG-1434, Revision 0, Page 17 To summarize, the staff holds the following positions regarding fuel coolability and long-term cooling.

(1)

Coolability

i.

Accident analyses should consider the impact of fuel relocation and dispersal on the accident progression and should verify that the limits defined in DG-1263, Revision 1, are not exceeded. These analyses should assume that all fuel dispersed from an assembly accumulates at the spacer grid immediately below the burst location.

ii.

In addition to addressing the impact of dispersal on the accident progression, accident analyses should demonstrate that the fuel particle beds that form on the spacer grids do not exceed the dryout heat flux calculated using the 0-D Lipinski model, as described in SKI Report 02:17.

iii.

Analyses should consider the accumulation of dispersed fuel particles in other locations within the RCS or containment. For these debris beds, the analyses should demonstrate that the dryout heat flux is not exceeded using the 0-D Lipinski model.

iv.

Calculations should consider a range of fuel particle sizes from 0.125 mm to 4 mm and a range of dispersed fuel debris bed porosities from 20 percent to 40 percent.

(2)

Long-term cooling

i.

Analyses should demonstrate that adequate coolant flow is provided to remove decay heat both from within the core and from fuel dispersed out of the core.

ii.

Analyses should demonstrate that the dryout heat flux is not exceeded for particle debris beds. These calculations should consider the impacts of coagulants and other post-LOCA debris that may reduce the porosity of the debris beds below 20 percent.

c.

Re-criticality GDC 27 requires that reactivity control systems be designed to shut down the reactor and maintain a subcritical configuration under postulated accident conditions. Existing LWRs accomplish this through a combination of control rods (pressurized-water reactors (PWRs)) or control blades (boiling water reactors (BWRs)) and soluble boron injection (primarily PWRs, but BWRs include emergency boron injection systems that can be used if other reactivity control systems are unavailable). Design analyses are performed assuming an intact rod bundle geometry.

Thus, the impact of fuel dispersal on the ability to keep the fuel subcritical should be considered.

Once again, it is difficult to predict the transport of fuel particles within the RCS.

Fortunately, when it comes to assessing the potential for re-criticality, defining a bounding scenario is straightforward. The WGFS report on FFRD (Ref. 12) describes one such analysis that used a homogeneous mixture of fuel and unborated water to calculate the infinite multiplication factor (k) (i.e., the effective multiplication factor (keff) for an infinite system; any real system

DG-1434, Revision 0, Page 18 would have some neutron leakage that would reduce the multiplication factor, such that keff < k).

The isotopic composition of the fuel mixture was calculated based on a 17x17 PWR assembly with fuel enriched to 4.8 weight percent uranium (U)-235 and 24 rods containing 10 weight percent gadolinia (Gd2O3) enriched to 3.2 weight percent U-235. The analysis showed that k is less than 1 for fuel with burnup greater than 50 GWd/MTU, regardless of the ratio of fuel to moderator. Given that the fine fragmentation threshold is above this burnup, there is no chance of re-criticality for fuel with this initial enrichment.

To assess the impacts of increasing the enrichment on the potential for re-criticality, the NRC staff performed additional analyses using the SCALE code suite. For these analyses, documented in the NRC memorandum Evaluation of Re-criticality within the Lower Plenum due to Fuel Fragmentation, Relocation and Dispersal (Ref. 30), the staff used SCALE-ORIGAMI to calculate depletion to a burnup of 55 GWd/MTU for fuel initially enriched to 8 weight percent U-235, then used SCALE-Shift to compute the effective multiplication factor (keff). Initial calculations showed that the infinite multiplication factor exceeded 1, so the analysis was refined to account for neutron leakage effects. The staff performed these calculations using a simplified representation of the lower plenum of a large PWR. These calculations assume a homogeneous mixture of fuel and unborated water at room temperature with a steel reflector (i.e., the lower head) below the particle bed and unborated above the debris bed. The staff looked at a range of water volume fractions and debris bed heights. The analysis showed that the configuration would remain subcritical for all the scenarios analyzed, which included dispersed masses as high as 7 metric tons with a water volume fraction of 32 percent. Note that for that scenario, which involves a quantity of dispersed fuel the staff considers highly conservative relative to expectations associated with envisioned strategies for reactor operation with high-burnup fuel, the calculated keff was less than 0.85, which is far below the keff = 1 requirement for criticality.

The staff believes that the calculations described in its memorandum on re-criticality (Ref. 30) may bound realistic fuel dispersal scenarios currently envisioned for many plants. First, the maximum fuel mass used in the calculations far exceeds estimated masses of fuel dispersal included in previous analyses, such as those performed in by the NRC staff and documented in Bielen et al. (Ref. 22). Second, the calculations assume that all dispersed fuel would have a burnup of 55 GWd/MTU. More realistically, much of the fuel dispersed would have a burnup greater than 55 GWd/MTU. Since the fissile content decreases with increasing burnup, assuming all dispersed fuel has a burnup of 55 GWd/MTU is conservative. However, uncertainty remains concerning whether other fuel dispersal configurations may be more or less limiting than the lower head from a re-criticality perspective. Most configurations would have significantly more neutron leakage than the lower head configuration, resulting in a lower keff for all other parameters remaining the same. However, different reactor designs may have different internal geometries (e.g., flow skirts, guide tubes, instrument tubes) that may affect the configuration of accumulated fuel particles, resulting in different quantitative results than those calculated in the staffs analyses. Additionally, one geometry not explicitly considered in the calculations is an accumulation of dispersed fuel on spacer grids within the core. From a general neutronics perspective, the critical geometry with fuel accumulation on spacer grids may resemble the configuration with intact cladding and no dispersal; however, the quantitative results may be different. Locally, particularly in the vicinity of high-burnup assemblies, some of the fuel would be outside the rods rather than inside, which would result in a different critical geometry.

However, the total fuel mass would be the same, and the fuel to moderator ratio would be similar for the intact and dispersed fuel configurations. In addition, while not credited for some reactor designs for the large-break LOCA event, licensees may be able to demonstrate that the control elements (i.e., PWR control rods or BWR control blades) may be fully inserted into the core well before fuel is dispersed for breaks about the TBS.

DG-1434, Revision 0, Page 19 Licensees should demonstrate that the potential for recriticality is addressed for their plant configuration, which may be feasible through comparisons to the staffs calculations. Note that the calculations are based on a maximum fuel enrichment of 8 weight percent U-235 and make assumptions concerning the geometric arrangement. Licensees considering fuel accumulation with significantly different geometries from those considered in the staffs calculations or using higher fuel enrichments, doped UO2, or MOX fuel, or with other important design differences not considered by the staff should provide additional justification to address the impacts of re-criticality. For licensees that are bounded by the NRC staff's calculation with significant margin, the NRC staff expects that qualitative engineering arguments would suffice to demonstrate acceptable results for avoiding recriticality.

d.

Radiological consequences and environmental qualification Finally, fuel dispersal could potentially challenge accident dose limits and the performance of safety-related equipment during the accident. DG-1425 includes one acceptable approach to address the potential impacts of fuel dispersal on the accident source term. This accident source term may be used in conjunction with RG 1.89 to assess the potential impacts of fuel dispersal on the environmental qualification of safety-related equipment.

DG-1434, Revision 0, Page 20 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. 31); 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-1434, Revision 0, Page 21 REFERENCES 3

1.

U.S. Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, 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) 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, Washington, DC.

5.

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

6.

NRC, RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Washington, DC.

7.

NRC, Draft Regulatory Guide (DG)-1425 (proposed Rev. 2 of RG 1.183), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Washington, DC.

8.

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

9.

NRC, DG-1263, Revision 1, (proposed new RG 1.224), Establishing Analytical Limits for Zirconium-based Alloy Cladding, Washington, DC.

10.

NRC, NUREG-2121, Fuel Fragmentation, Relocation, and Dispersal during the Loss-of-Coolant Accident, Washington, DC, March 2012. (ML12090A018)

11.

NRC, NUREG-2160, Post-Test Examination Results from Integral, High-Burnup, Fueled LOCA Tests at Studsvik Nuclear Laboratory, Washington, DC, August 2013. (ML13240A256)

12.

Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA), NEA/CSNI/R(2016)16, Report on Fuel Fragmentation, Relocation, Dispersal, Paris, France, October 2016.4 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 email 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 email to pdr.resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

4 Copies of OECD NEA documents may be obtained through their website, https://www.oecd-nea.org/, or by writing the Nuclear Energy Agency, 2, rue André Pascal, 75775 Paris Cedex 16, France.

DG-1434, Revision 0, Page 22

13.

NRC, Research Information Letter (RIL) 2021-13, Interpretation of Research on Fuel Fragmentation, Relocation, and Dispersal at High Burnup, Washington, DC, December 2021.

(ML21313A145)

14.

NRC, NUREG/CR-7307, Phenomena Identification and Ranking Tables on High Burnup Fuel Fragmentation, Relocation, Dispersal, and Its Consequences for Design-Basis Accidents in Pressurized-and Boiling-Water Reactors, Washington, DC, May 2024. (ML24155A058)

15.

NRC, Nuclear Regulatory Commission International Policy Statement, Federal Register, Vol. 79, No. 132, pp. 39415-39418 (79 FR 39415), Washington, DC, July 10, 2014.

16.

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

17.

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

18.

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

19.

Doyle, P., M. Ridley, J. Harp, Y. Yan, M. White, and N. Capps, Dispersal of high-burnup fuel fragment surrogate particles during and after loss-of-coolant accident tests, Nuclear Engineering and Design, 424: 113261, 2024.6

20.

Capps, N., and R. Sweet, Model for determining rupture area in Zircaloy cladding under LOCA conditions, Nuclear Engineering and Design, 401: 112096, 2023.

21.

NRC, NUREG-0630, Cladding Swelling and Rupture Models for LOCA Analysis, Washington, DC, April 1980. (ML053490337)

22.

Bielen, A., J. Corson, and J. Staudenmeier, NRCs methodology to estimate fuel dispersal during a large break loss of coolant accident, Nuclear Engineering and Design, 426: 113377, 2024.

23.

Wiesenack, W., HPR-383, Summary and Comparison of LOCA Tests with BWR Fuel in the Halden Reactor Project Test Series IFA-650, Halden Reactor Project, Halden, Norway, June 2015.

24.

Schappel, D., and N. Capps, Impact of LWR assembly structural features on cladding burst behavior under LOCA conditions, Nuclear Engineering and Design, 418: 112887, 2024.

25.

Pontillon, Y., M.P. Ferroud-Plattet, D. Parrat, S. Ravel, G. Ducros, C. Struzik, I. Aubrun, G.

Eminet, J. Lamontagne, J. Noirot, and A. Harrer, Experimental and Theoretical Investigation of Fission Gas Release from UO2 up to 70 GWd/t under Simulated LOCA Type Conditions: The 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.

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

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

DG-1434, Revision 0, Page 23 GASPARD Program, Proceedings of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, September 19-22, 2004.

26.

Capps, N., L. Aagesen, D. Andersson, O. Baldwin, W.C. Brinkley, M.W.D. Cooper, J. Harp, S.

Novascone, P-C.A. Simon, C. Matthews, and B.D. Wirth, Empirical and mechanistic transient fission gas release model for high-burnup LOCA conditions, Journal of Nuclear Materials, 584:

154557, 2023.7

27.

Jernkvist, L.O., Modelling of fine fragmentation and fission gas release of UO2 fuel in accident conditions, European Physical Journal (EPJ) Nuclear Sciences & Technology, 5: 11, 2019.8

28.

U. S. Atomic Energy Commission, Rulemaking Hearing: Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors, Opinion of the Commission, Washington, DC, December 28, 1973. (ML20236U832)

29.

Swedish Nuclear Power Inspectorate (SKI), SKI Report 02:17, A Review of Dryout Heat Fluxes and Coolability of Particle Beds, Stockholm, Sweden, April 2002.9

30.

Esmaili, H., NRC, memorandum to S. Krepel, NRC, Evaluation of Re-criticality within the Lower Plenum due to Fuel Fragmentation, Relocation and Dispersal, Washington, DC, November 21, 2024. (ML24319A262)

31.

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

7 Journal of Nuclear Materials 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 it website: https://www.elsevier.com/books-and-journals.

8 EPJ Nuclear Sciences & Technologies is an Open Access journal. Copies of this journal can be accessed at their website:

https://epjn.epj.org/.

9 Copies of SKI documents may be obtained through the Swedish Radiation Safety Authoritys website:

https://www.stralsakerhetsmyndigheten.se/en/.