ML22151A117

Topfuel 2022 - NRCs Research Information Letter on Fuel Fragmentation, Relocation and Dispersal

ML22151A117
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Issue date:	05/31/2022
From:	Michelle Bales, Chung A, James Corson, Lucas Kyriazidis NRC/RES/DSA, OECD Nuclear Energy Agency (NEA)
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Corson J
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Download: ML22151A117 (9)

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NRC's Research Information Letter on Fuel Fragmentation, Relocation and Dispersal

James Corson,* Alice Chung,* Lucas Kyriazidis,* Michelle Bales

• United States Nuclear Regulatory Commission, Rockville, Maryland, USA, James.Corson@nrc.gov OECD Nuclear Energy Agency, Boulogne -Billancourt, France

Extensive research has been conducted on fuel (LOCAs). Over the last 10 or more years, research has fragmentation, relocation, and dispersal (FFRD) during a indicated that high-burnup fuel can finely fragment, loss-of-coolant accident (LOCA). This research has shown relocate axially, and disperse into the coolant under certain that FFRD phenomena are correlated with burnup. As the LOCA conditions. Transient conditions may cause trapped U.S. nuclear industry pursues the operation of plants with gaseous fission products to be released from the pellet, higher fuel burnup levels, it is important to understand and increasing rod internal pressure and impacting burst account for FFRD -related phenomena and their impact on timing. Finely fragmented fuel may easily relocate axially regulatory figures of merit (e.g., peak cladding within the fuel rod following ballooning of overheated temperature) in licensing applications. Recently, the U.S. cladding, impacting local heat distribution along the fuel Nuclear Regulatory Commissions Office of Nuclear rod, and potentially disperse through the breach in the Regulatory Research (RES) published a research cladding. If fuel disperses out of a burst fuel r od, it could information letter to communicate the staffs interpretation compromise coolable geometry, impact the accident of findings from experimental programs on FFRD and to progression, complicate the safety demonstration, and alter define conservative, empirical boundaries for FFRD - cooling for long-term decay heat removal for both the fuel related phenomena. rods in the core and the dispersed fuel particles.

The RIL provides a basis for limiting the analysis of In 2012, the U.S. Nuclear Regulatory Commission FFRD to regions of the core with specific characteristics. (NRC) published NUREG -2121 [1] summarizing research Data from experimental programs conducted to date related to fuel behavior during a LOCA. Later, it published suggests that fine fragmentation is limited to fuel above 55 SECY 0148 [2] evaluating FFRD in the context of a gigawatt days per metric ton of uranium (GWd/MTU) draft final rule on emergency core cooling system pellet average burnup. Axial fuel relocation is limited to performance under a LOCA. SECY 0148 concluded regions of the fuel rod that have a local cladding strain that immediate regulatory action was not needed to address greater than 3 percent. Relocated fuel fragments can FFRD phenomena at that time. However, this conclusion occupy between 60 percent and 85 percent of the fuel rod was closely linked with existing fuel design limits and cross-sectional area in the balloon region. The propensity assumptions on how high -burnup fuel would be operated.

for fuel dispersal is correlated with fuel fragment size and Since the publication of SECY 0148 in 2015, the burst opening size; however, cladding burst and fuel NRC has continued to participate in multilateral research relocation are prerequisites. This effectively limits fuel activities related to FFRD, including the Studsvik Cladding dispersal by the same parameters as fine fragmentation Integrity Project (SCIP) and the OECD Nuclear Energy and relocation (i.e., pellet average burnup greater than Agencys Working Group on Fuel Safety [3]. In light of 55 GWd/MTU and cladding strain greater than 3 percent). recent interest by the U.S. nuclear industry to increase rod -

Finally, data from experimental programs conducted to average burnup limits beyond 62 GWd/MTU, staff date suggests that significant quantities of fission gas may published Research Information Letter (RIL) 2021-13 to be released during a LOCA transient. Transient fission gas offer its interpretation of research on the subject. RIL 2021 -

release becomes increasingly significant with increasing 13 was subjected to an external peer review by an burnup, with re leases as high as 20 percent observed from international panel of experts who have been heavily a fuel rod segment with an average burnup of 70 involved in some of the experimental programs discussed GWd/MTU. Fission gas released during a LOCA may in the RIL. The RIL was also reviewed by NRCs Advisory impact fuel rod ballooning and burst behavior and, thus, Committee on Reactor Safeguards [4]. The final version of fuel relocation and dispersal. the RIL includes changes based on these peer reviews, as described in Appendix B of RIL 2021-13 and in Ref. [5].

INTRODUCTION RIL 2021-13 addresses five elements of the RES staffs interpretation of FFRD research and describes the Emergency core cooling systems, core, and fuel must technical basis for these elements:

be designed to ensure that the fuel rods maintain a coolable 1. Establish an empirical threshold at which fuel geometry following postulated loss -of-coolant accidents pellets become susceptible to fine fragmentation.

2. Establish a local cladding strain threshold below rodlets [6, 7, 8]. Rodlets were pressurized to a range of which fuel relocation is limited. pressures to induce the various ballooning characteristics.

3. Examine experimental results of the mass of The majority of tests imposed thermal -hydraulic boundary dispersible fuel as a function of burnup. conditions to simulate a large -break LOCA, including

4. Provide evidence of significant transient fission heatup that induced ballooning and burst and, in some gas release (tFGR) that may impact ballooning cases, high-temperature steam oxidation as well as reflood and burst behavior of high -burnup fuel under and quench. Tests performed in the Halden reactor utilized LOCA conditions. nuclear heating while tests performed at Studsvik and at

5. Establish the basis for a range of packing fractions Oak Ridge National Laboratory utilized furnace heating.

of relocated but nondispersed fuel in the balloon Elsewhere, analysis has been performed for different region. heating methods, nuclear - and furnace-heated, concluding This paper will summarize the evaluation of these five that the radial temperature profile in the fuel (and therefore elements presented in the RIL. But first, it will describe the the thermally induced pellet stresses) resulting from both experimental programs considered in the RIL. methods should be similar (Capps, et al., 2021).

Experimental programs on FFRD and tFGR have also EXPERIMENTAL PROGRAMS CONSIDERED IN included separate effects tests, including heating tests on RIL 2021-13 small sections of fuel rods a few pellets in height, as well as fuel pellet disks [7, 9, 10, 11].

The behavior of fuel rods under LOCA conditions has Many of the tests described above have included been studied for decades. Experiments have often focused posttest examinations to quantify the degree of fuel on the timing and degree of ballooning and burst, the fragmentation. Examinations have included sifting fuel mechanical behavior of the cladding following the LOCA fragments using a sieve stack with different mesh sizes, transient, and the cooling effectiveness around ballooned allowing for mass measurements of fuel collected for each fuel rods. fragment size group.

In 2006, the Halden Reactor Project (HRP) ran a test (IFA-650.4) on a fuel rodlet with a segment average RES STAFFS INTERPRETATION OF FFRD burnup0F1 of 92.3 GWd/MTU [6]. Following the test, RESEARCH significant fuel relocation and dispersal were observed.

Even though fuel fragmentation and relocation had Element 1: Fine Fuel Fragmentation Threshold occurred in tests before IFA-650.4 and, in some cases, minor fuel loss had even been observed, 1F2 there had been Combining Halden, NRC, SCIP, and ORNL integral little effort to quantify or specifically study the experiments, more than 35 tests were conducted on rodlets fragmentation or relocation of fuel pellets. The results of with burnups ranging from approximately IFA-650.4 were considered so significant that they c aused 45 to 90 GWd/MTU segment average, for which detailed a refocus of international LOCA research to better observations on fragmentation are available. These tests understand FFRD. Experimental methods were designed to can be examined to define an empirical threshold at which anticipate FFRD and better capture relevant experimental fuel pellets become susceptible to fine fragmentation and features. Posttest examinations were developed to quantify fuel dispersal becomes a concern.

the degree of fragmentation and relocation. In addition, The mass fractions of all mobile fuel fragments experiments began to largely focus on irradiated material smaller than 1 millimeter (mm) and 2 mm, shown in Fig.

above 50 GWd/MTU. For these reasons, RIL 2021 -13 1, were examined to evaluate tren ds in fine fragmentation.

focuses on insights gained from experiments conducted The mobile fuel mass fractions include both the mass after 2006. dispersed during the LOCA tests and that shaken out of Experimental programs on FFRD since 2006, such as the test segments following the LOCA test. This figure those conducted in the HRP and the Studsvik Cladding shows that fuel can have a notable portion of fragments Integrity Project (SCIP), include tests on refabricated, below 1 m m and 2 mm at 60 GWd/MTU. 2F3 30- to centimeter (cm)-long, internally pressurized

1 Unless otherwise noted, the burnup of rodlets used in the 2 A loss of fuel mass equivalent to about one fuel pellet was experimental programs cited in this paper is the average observed in integral LOCA tests performed at Argonne National burnup of the rod segments. Because the test segments used Laboratory on boiling-water reactor (BWR) fuel rods with local in the cited test programs are relatively short, the variation burnup of 64 GWd/MTU. NUREG-2121 [1] contains further of burnup along the axial length was often minimal, and discussion.

3 Segment burnup values in SCIP III are determined from gamma thus segment average burnup can be used interchangeably scanning and are characterized by a relative uncertainty of +/-5%

in an analysis with pellet average burnup. [44]. Work is ongoing in SCIP IV to reduce the uncertainty of

approximately 4 percent in the lower part of the fuel segment and 5 percent in the upper part of the fuel segment at the locations where the gamma scan indicates fuel remains. A wire probe was also used to examine the extent of empty cladding following the LOCA and following the shaking. The comparison of wire probe measurements before (i.e., after LOCA) and after shaking in Fig. 2 shows that additional fuel was dispersed during shaking, and the gamma scan confirms that some fuel remained in the upper and lower halves. TABLE presents the boundary of relocated fuel, as determined by wire probe measurements from the NRCs LOCA tests at Studsvik

[12].

Similar measurements were taken in the SCIP III program on 10 segments after LOCA testing to investigate Fig. 1. Measurements of percent of fuel fragments smaller the relationship between cladding strain and relocation. In than 1 mm and 2 mm as a function of segment average SCIP III, posttest gamma-scan data were evaluated to burnup [6, 25, 12, 26, 27, 28, 29, 7, 30, 31]. determine where the fuel column was intact. The position of the intact fuel column was then compared against the The data in Fig. 1 suggest that the onset for fine local cladding strain. STUDSVIK -SCIP III-253, SCIP fragmentation may occur below 60 GWd/MTU pellet IIISubtask 1.1: Fuel fragmentation, rel ocation and average burnup; however, no tests have quantified fragment size for comparison between 45 and 60 GWd/MTU. Extrapolating from the large amount of data above 60 GWd/MTU, the data suggests a n empirical threshold for the onset of fine fuel fragmentation near a pellet average burnup of 55 GWd/MTU.

Element 2: Cladding Strain Threshold for Relocation

Another aspect of quantifying the amount of fuel dispersal associated with a burst of high-burnup fuel rods is related to the axial length of the fuel rod predicted to Fig. 2. Gamma scan, profilometry, and wire probe experience fuel relocation. Experimental results from the measurements from NRC test 192 [12].

NRCs LOCA test program at Stu dsvik, presented in Fig.

2, show that in regions of very low cladding diametrical strain, fuel does not relocate axially, even when agitated.

Fig. 3 provides an image of the fuel fragments collected after shaking, indicating that this test segment experienced fine fragmentation.

The gamma scan shown in Fig. 2 was made after the test segment was broken in half and both the upper and lower segment halves were shaken to dislodge any fuel.

While the shaking action was not designed to represent any particular load experienced during a LOCA, the observation that fuel remained in the test segment after shaking is an indication that fuel pellets in low strain Fig. 3. Fuel fragments collected from the top end of regions away from the burst location tend to resist axial rod 192 after gentle shaking and just before gamma relocation, even in fuel rods that have experienced fine scan [12].

fragmentation. The local strain from this test was

burnup values derived from gamma scan measurements; however, the findings remain within the +/-5% previously stated uncertainty.

dispersal, Final Summary Report, issued 2019 [7], reports size with respect to burnup (Capps, et al., 2021; NRC, and discusses the results. When the results of the 2012).

10 SCIP III tests are combined with the NRCs 6 LOCA Finally, the mass of the short fuel rod segments used tests presented above, the data indicates an average value in these experiments and the relatively short balloon of a strain threshold for relocation is 3. 7 percent, with a (resulting from a relatively steep temperature gradient standard deviation of 1.7 percent [13]. induced by furnace heating) may not be representative of The observations discussed above suggest that fuel the mass dispersed in a full -length rod with a different relocation is limited in regions of the fuel rod experiencing strain profile.

less than 3-percent cladding strain. Based on the se observations, it is reasonable to assume that all fuel above a burnup of 55 GWd/MTU in the length TABLE I. Estimates of relocation strain thresholds from of the rod with greater than 3-percent cladding strain could the NRCs LOCA tests at Studsvik [12]. disperse.

Test Strain Strain Element 4: Transient Fission Gas Release Number threshold, top threshold,

(%) bottom (%) The amount of fission gas released during normal 189 6.0 3.0 operating and accident conditions is important to 191 6.0 4.0 understanding the behavior of a nuclear fuel rod. FGR 192 5.0 4.0 introduces adverse fuel performance effects that include 193 1.0 4.0 the degradation of the thermal conductivity within the 196 3.0 5.0 fuel-clad gap and an increase in cladding hoop strains when 198 4.5 9.0 rod internal pressure exceeds the reactor coolant system pressure [14]. During steady-state normal operation, Element 3: Mass of Dispersible Fuel fission gas release (FGR) into the rod void volume is governed by diffusion. Modern fuel performance codes Another objective of RIL 2021-13 is to use available predict normal-operation FGR well. Howe ver, research to document insights that could be used to develop observations in experimental programs, such as the HRP,

a model quantify the amount of fuel dispersal associated SCIP, and the French GASPARD program, indicate that with burst of high-burnup fuel rods. The Halden, NRC, FGR can be exacerbated by LOCA-like transients. This SCIP, and ORNL experiments were examined to inform phenomenon is termed transient fission gas release the model. (tFGR).

In developing the RIL, the staff considered various Fission gas released during a transient may further approaches to interpret the available data for an empirical increase rod internal pressure, which may lead to cladding model. Six approaches are discussed in an appendix to the failure that would not have been expected if tFGR was RIL, but ultimately the most conservative approach was neglected [15].

proposed in the RIL in recognition of three sources of To initiate tFGR in the experiments referenced below, uncertainty. a fuel pellet or segment is subjected to a temperature First, the posttest examination of several rodlets transient. The tFGR tests generally consist of three phases:

revealed that fuel in the segment was finely fragmented, 1. a thermal equilibrium phase even when limited dispersal was observed during the test. 2. a temperature transient phase Handling of these rods showe d that this finely fragmented 3. a cooling phase fuel could readily relocate within the rod and fall out after Most of the temperature ramp rates observed in the the test. It would be difficult to rule out the possibility that experiments varied between 0.2 oC per second and 20 oC forces acting on the fuel in a design -basis LOCA c ould per second. Once the target temperature is reached, the fuel result in greater dispersal than observed du ring these segment is either held at temperature for a specified time experiments. followed by cooling, or the fuel segment is immediately Second, the burst opening size is a key determinant of cooled by turning off the furnace. To simulate the the amount of fuel that can disperse (i.e., smaller burst blowdown phase of a LOCA, some experiments were openings could limit the particle size and total mass of performed in a steam environment or with water introduced dispersed fuel, while a larger burst opening could allow within the test environment [16].

larger particles and more mass to disperse). B urst opening Fig. 4 presents a compilation of more than 15 tFGR size can vary stochastically with respect to fuel rod tests from several experimental programs. tFGR results characteristics such as rod internal pressure. Data collected presented in Fig. 4 exclude fission gas released during base through various LOCA and cladding balloon-burst test irradiation and account only for the fission gas released programs also indicate wide variability in burst opening during the LOCA -like transient. This is because experiments are conducted on refabricated fuel rod LOCA conditions and did not burst, yet significant FGR of 18.6 percent was observed during the test. Thus, it is not clear whether results from single-pellet and furnace tests are truly conservative compared to in-pile LOCA conditions.

Finally, it is worth noting that these tests have been performed on short (about 30- cm) segments or single pellets. It is unclear how extensive tFGR would be in a full-length rod during a LOCA.

Researchers have developed tFGR models applicable to LOCAs that account for these known dependenc ies [18, 15]. However, these models have received little validation to date and are therefore not ready for regulatory applications.

Element 5: Packing Fraction of Relocated Fuel Fig. 4. Measured tFGR as a function of segment average burnup. Circle symbols represent out -of-pile LOCA tests Axial fuel relocation and fuel packing within regions

[7, 40, 41, 16, 11]. The triangle represents an in-pile LOCA of a fuel rod that experience ballooning can significantly test [39], and crosses represent single pellet-clad affect LOCA analyses. When fuel redistributes axially nonwelded samples [10]. within the rod, it changes the axial power distribution and segments and samples, meaning the gas re leased during local cladding temperature. Pulverized fuel in a packed normal operation is no longer present. crumbled configuration will have an increased void Fig. 4 shows that tFGR tends to increase with fraction when compared to its undamaged state, impacting increasing fuel segment burnup. However, the simple plot the overall heat removal from the fuel rod. This will, in of tFGR versus burnup does not account for many test turn, affect temperatures in the fuel and cladding, variables that may significantly impact tFGR behavior. For potentially driving microstructural changes, FGR, instance, the Studsvik tests were performed with a low -fill differences in cladding ductility, ballooning and burst pressure (i.e., low hydrostatic pressure and constraint), behavior, and cladding oxidation. Fig. 5 illustrates axial while the single-pellet tests were unpressurized. Studies fuel relocation and packing.

have shown that tFGR decreases with increasing hydrostatic pressure [17, 9]. Thus, performing tFGR tests at low-fill pressures may be conservative (i.e., little to no hydrostatic pressure).

Furthermore, the terminal temperature in many of the tests shown in Fig. 4 is greater than 1,000 oC. This may be higher than best estimate predictions of peak temperatures for high-burnup fuel rods, and it is almost certainly higher than the temperature at which high -burnup rods would be expected to burst. 3F4 The GASPARD program showed that tFGR occurred in two temperature regimes: a burst release at lower temperatures (~600-800 oC) and a larger release at high temperatures (>1,000 oC) [10]. Only the lower temperature burst release would be expected to influence ballooning and burst behavior based on observed burst temperatures for high-burnup fuel rods. This suggests that the Studsvik and single -pellet (i.e., GASPARD) data in Fig. 4 may be conservative when considering the impact of tFGR on ballooning and burst behavior. On the other hand, Halden LOCA test IFA-650.14 (i.e., the Halden (In-Pile) Fig. 5. Illustration of fuel relocation and packing in the point in the figure) was subjected to more prototypical ballooned region.

4 However, PCTs for high-burnup rods are significantly impacted by pre-transient linear heat generation rates, so predicted PCTs are heavily influenced by the fuel loading pattern.

direction. Measurements were made of fuel rods that burst, as well as fuel rods that ballooned but did not burst.

The Cs-137 measurement was normalized so that the non-fragmented and nonrelocated fuel at the top end of the fuel column has a value of 1. This was then divided by the cross-sectional area at the given position, yielding the packing fraction, shown in Fig. 7. For this test, the average packing fraction in the lower portion of the balloon varied between 0.7 and 0.85, with an average value of approximately 0.78. In the upper part of the balloon, the packing fraction is lower (0.4- 0.7), likely because of a lack of fuel available to pack this region.

Fig. 6. Post-test gamma scans of IFA-650.9, a high-burnup PWR fuel rod subjected to a LOCA simulation at the Halden reactor [12].

This phenomenon has been observed in multiple Fig. 7. Packing fraction of 3V5-Q13 [7].

programs and facilities such as Halden, SCIP, and the Fig. 8 shows the average packing fraction of many of Power Burst Facility [1, 19, 6, 20]. In the various programs, the SCIP III LOCA tests calculated as described above.

the packing fraction, sometimes referred to as the filling The packing fraction reported by SCIP III is based on the ratio, is defined as the ratio of the volume of fuel to the total densely packed region typically below the burst in the available local volume. Early tests performed at the Power lower balloon area. Packing fraction in the upper balloon Burst Facility and at Forschungsreaktor 2 (Research area is typically lower due to fuel mass limitations. Fig. 8 Reactor 2 or FR2) in Germany on unirradiated or shows that there is a slight correlation between segment low-burnup fuels (up to 35 GWd/MTU) showed packing burnup and packing fraction. This may be due to the fractions in a range from roughly 60 to 80 percent [21, 20]. increase in fine fragmentation at higher burnups. It is Axial fuel relocation and packing were also observed possible that the finer fragments relocate more easily and during Haldens IFA -650.9 test, which consisted of a high-increase the packing fraction in the balloon region just burnup PWR rod, subjected to LOCA conditions. Fig. 6 below the burst location while decreasing the packing shows the posttest gamma scans of IFA -650.9. As can be fraction above the burst location. This is consistent with seen, a significant portion of the fuel stack was missing due recent discrete element modeling work [22]. However, the to axial fuel relocation and dispersal. The relocated fuel effect of burnup on packing fraction is not large; most had dropped to the lower portion of the rod near the burst measured packing fractions are near the average packing opening, where the diameter nearly doubled. In this fraction of 0.78. The exception is the lowest burnup test, ballooned area, the cesium (Cs)-137 and the ruthe nium which has a packing fraction of approximately 0.6. This is (Ru)-103 count rates were respectively consistent with discrete element modeling calculations for 30- 70 percent and 20- 30 percent higher than the general cases with no fine fragmentation [22].

level of the rod [1]. Later work at Halden on test IFA-650.12 estimated the average packing fraction in the balloon region to be approximately 55 percent based on cladding strain measurements and a fuel mass balance [6].

In SCIP, the packing fraction has been estimated from posttest gamma scans and profilometry measurements.

After the LOCA test, gamma scans we re performed on the fuel segment, measuring the Cs-137 signal in the vertical Previous tests on lower burnup fuel showed lower ballooning. Furthermore, ballooning behavior may be packing fractions (as low as approximately 0.6), which is sensitive to the method used to heat the rod, which varies consistent with the lower burnup SCIP III test. It is among the experiments described in the RIL. For these reasonable to use packing fraction values in this range for reasons, it may be necessary to make conservative LOCA calculations. In general, a larger packing fraction assumptions about the circumferential strain as a function will increase the local decay heat, which may in crease the of axial position when estimating the mass of fuel that may local cladding temperature. At the same time, a smaller be dispersed during a LOCA.

packing fraction may reduce local heat transfer and increase fuel temperatures, which in turn would impact CONCLUSIONS AND FUTURE WORK FGR and thus ballooning and burst behavior. I t is important to examine a range of packing fractions to account for these RIL 2021-13 summarizes the RES staffs competing effects on integral rod behavior. interpretation of research related to FFRD and present s a conservative, empirical threshold for when significant fuel fragmentation begins as 55 GWd/MTU pellet average burnup for standard UO 2 fuel. The research described in the RIL indicates that below this burnup value, fine fragmentation is not a concern.

The RIL also summarizes the RES staffs interpretation of available research to define empirical thresholds that could be used to quantify the amount of fuel dispersal associated with burst of high-burnup fuel rods. It defines a cladding strain threshold of 3 percent as a value below which fuel relocation is not a concern. The staff considered multiple empirical models to quantify fuel dispersal and concluded that predicting all fuel above 55 GWd/MTU in the region of the fuel rod with cladding strain above 3 percent represents a conservative prediction of fuel dispersal.

Fig. 8: Average packing fraction in the lower portion of the The RIL also summarizes the RES staffs balloon in the SCIP III tests interpretation of research related to tFGR. Research has shown that tFGR increases with burnup, with releases as high as about 20 percent observed for fuel at a pellet LIMITATIONS OF THE EMPIRICAL DATABASE average burnup of about 70 GWd/MTU. Such releases could have a significant impact on cladding ballooning and The thresholds presented in RIL 2021-13 are purely burst behavior during a LOCA and should be accounted for empirical. All of the tests were performed on uranium in fuel performance models.

dioxide fuel in zirconium alloy cladding, so the observed Lastly, the RIL investigates and summarizes the RES behavior may not apply to new fuel or cladding designs. staffs interpretation of the research related to the packing Furthermore, the interpretations in the RIL focus on burnup fraction. Most of the data show packing fractions ranging as the factor most important to FFRD behavior. However, from 60 to 85 percent. Packing fraction also varies axially, it is likely that characteristics that evolve with burnup, such with high packing fractions in the low er portion of the as porosity, stresses within the fuel pellet, grain growth, balloon and lower packing fractions in the upper portion of and subgrain formation, are more directly correlated with the balloon. Because of competing phenomena controlling FFRD behavior. Other parameters - such as temperature, the fuel and cladding temperatures in a region of ballooned pressure, and heat-up rate - may also significantly impact cladding and packed fuel during a LOCA, it is important to observed behavior. More experimental data would be model a range of packing fractions to evaluate the effects needed to better understand the impacts of the of the packing fraction on cladding temperature and other aforementioned parameters on FFRD and transient fission fuel rod performance metrics.

gas release behavior. The thresholds identified in the RIL are based The dispersal models described in the RIL all require primarily on burnup and apply only to uranium dioxide fuel an accurate prediction of the axial strain profile along the in zirconium alloy cladding. NRC is participating in rod. However, most LOCA ballooning models were international research programs to address some of the developed to maximize the circumferential strain in order limitations identified in the empirical database, including to assess flow blockage and core coolability. The models SCIP IV. Furthermore, the NRC is participating in the have not been calibrated against the axial extent of Electric Power Research Institute (EPRI)s Collaborative Research on Advanced Fuel Technologies (CRAFT) for Light Water Reactors program. CRAFT is developing a Relocation, and Dispersal at High Burnup, Washi ngton, LOCA test plan to be conducted in the Transient Reactor DC: U.S. Nuclear Regulatory Commission Advisory Test (TREAT) facility at Idaho National Laboratory and Committe on Reactor Safeguards, December 20, 2021.

the Severe Accident Testing Station (SATS) at Oak Ridge 5. R. V. Furstenau, Memorandum to Dr. Joy L.

National Laboratory. The goal of the test plan is to address Rempe, re: Staff Response to ACRS Letter Dated knowledge gaps in fuel fragmentation and relocation December 20, 2021, in Regard to Research Information phenomenology. Letter 2021-13 on the Interpretation of Research on Fuel Defining when fuel pellets become susceptible to Fragmentation, Relocation, and Dispersal at High Burnup, fragmentation is the first step and a key piece of Washington, DC: U.S. Nuclear Regulatory Commission, information which could be used to design fuel, cladding January 20, 2022.

and operating regimes that limit or prevent FFRD. 6. W. Wiesenack, "HRP-380, Summary of the However, it is only part of unders tanding the overall safety Halden Reactor Project LOCA Test Series IFA-650,"

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developing a methodology to quantify t he mass of fuel that Terrani, N. Woolstenhulme and D. Wachs, "A Critical may be dispersed from a high burnup core during a loss of Review of High Burnup Fuel Fragmentation, Relocation, coolant accident, based on a core loading pattern developed and Dispersal under Loss-Of-Coolant Accident by the Nuclear Energy Advanced Modeling and Simulation Conditions," Journal of Nuclear Materials, 546, 152750 (NEAMS) program [23]. The methodology uses the (2021).

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