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At least 127 records · Page 7

A Summary of the Space Shuttle Columbia Tragedy and the Use of LS Dyna in the Accident Investigation and Return to Flight Efforts

On February 1, 2003, the Space Shuttle Columbia broke apart during reentry resulting in loss of 7 crewmembers and craft. For the next several months an extensive investigation of the accident ensued involving a nationwide team of experts from NASA, industry, and academia, spanning dozens of technical disciplines. The Columbia Accident Investigation Board (CAIB), a group of experts assembled to conduct an investigation independent of NASA concluded in August, 2003 that the cause of the loss of Columbia and its crew was a breach in the left wing leading edge Reinforced Carbon-Carbon (RCC) thermal protection system initiated by the impact of thermal insulating foam that had separated from the orbiters external fuel tank 81 seconds into the missions launch. During reentry, this breach allowed superheated air to penetrate behind the leading edge and erode the aluminum structure of left wing which ultimately led to the breakup of the orbiter. In order to gain a better understanding the foam impact on the orbiters RCC wing leading edge, a multi-center team of NASA and Boeing impact experts was formed to characterize the foam and RCC materials for impact analysis using LS Dyna. Dyna predictions were validated with sub-component and full scale tests. LS Dyna proved to be a valuable asset in supporting both the Columbia Accident Investigation and NASA s return to flight efforts. This paper summarizes Columbia Accident and the nearly seven month long investigation that followed. The use of LS-DYNA in this effort is highlighted. Contributions to the investigation and return to flight efforts of the multicenter team consisting of members from NASA Glenn, NASA Langley, and Boeing Philadelphia are introduced and covered in detail in papers to follow in these proceedings.

Melis, Matthew↗

The Implications of Handling Qualities in Civil Helicopter Accidents Involving Hover and Low Speed Flight

Because of increasing accident rates in Army helicopters in hover and low speed flight, a study was made in 1999 of accidents which could be attributed to inadequate stability augmentation. A study of civil helicopter accidents from 1993-2004 was then undertaken to pursue the issue of poor handling qualities in helicopters which, in almost all cases, had no stability augmentation. The vast majority of the mishaps studied occurred during daylight in visual meteorological condition, reducing the impact of degraded visual environments (DVE) on the results. Based on the Cooper-Harper Rating Scale, the handling qualities of many of the helicopters studied could be described as having from "very objectionable" to "major" deficiencies. These costly deficiencies have resulted in unnecessary loss of life, injury, and high dollar damage. Low cost and lightweight augmentation systems for helicopters have been developed in the past and are still being investigated. They offer the potential for significant reductions in the accident rate.

Dugan, Daniel C.↗

The Columbia Accident Investigation and The NASA Glenn Ballistic Impact Laboratory Contributions Supporting NASA's Return to Flight

On February 1, 2003, the Space Shuttle Columbia broke apart during reentry, resulting in loss of the vehicle and its seven crewmembers. For the next several months, an extensive investigation of the accident ensued involving a nationwide team of experts from NASA, industry, and academia, spanning dozens of technical disciplines. The Columbia Accident Investigation Board (CAIB), a group of experts assembled to conduct an investigation independent of NASA, concluded in August, 2003 that the most likely cause of the loss of Columbia and its crew was a breach in the left wing leading edge Reinforced Carbon-Carbon (RCC) thermal protection system initiated by the impact of thermal insulating foam that had separated from the orbiters external fuel tank 81 seconds into the mission's launch. During reentry, this breach allowed superheated air to penetrate behind the leading edge and erode the aluminum structure of left wing, which ultimately led to the breakup of the orbiter. The findings of the CAIB were supported by ballistic impact tests, which simulated the physics of External Tank Foam impact on the RCC wing leading edge material. These tests ranged from fundamental material characterization tests to full-scale Orbiter Wing Leading Edge tests. Following the accident investigation, NASA spent the next 18 months focused on returning the shuttle safely to flight. In order to fully evaluate all potential impact threats from the many debris sources on the Space Shuttle during ascent, NASA instituted a significant impact testing program. The results from these tests led to the validation of high-fidelity computer models, capable of predicting actual or potential Shuttle impact events, were used in the certification of STS-114, NASA s Return to Flight Mission, as safe to fly. This presentation will provide a look into the inner workings of the Space Shuttle and a behind the scenes perspective on the impact analysis and testing done for the Columbia Accident Investigation and NASA's Return to Flight programs. In addition, highlights from recent Shuttle missions are presented.

Melis, Matthew E.↗

Pressing the Approach: A NASA Study of 19 Recent Accidents Yields a New Perspective on Pilot Error

This article begins with a review of two sample airplane accidents that were caused by pilot error. The analysis of these and 17 other accidents suggested that almost all experienced pilot operating in the same environment in which the accident crews were operating and knowing only what the accident crews knew at each moment of the flight, would be vulnerable to making a similar decision and similar errors. Whether a particular crew in a given situation makes errors depends on somewhat random interaction of factors. Two themes that seem to be prevalent in these cases are: Plan Continuation Bias, and Snowballing Workload.

Berman, Benjamin A.↗

Aircraft Loss-of-Control Accident Analysis

Loss of control remains one of the largest contributors to fatal aircraft accidents worldwide. Aircraft loss-of-control accidents are complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. To gain a better understanding into aircraft loss-of-control events and possible intervention strategies, this paper presents a detailed analysis of loss-of-control accident data (predominantly from Part 121), including worst case combinations of causal and contributing factors and their sequencing. Future potential risks are also considered.

Belcastro, Christine M.↗

Future Integrated Systems Concept for Preventing Aircraft Loss-of-Control Accidents

Loss of control remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft loss-of-control accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. This paper presents future system concepts and research directions for preventing aircraft loss-of-control accidents.

Belcastro, Christine M.↗

Orbiter Payload Bay Bucket Hoist Mishap...An Accident We Should Never Forget

This slide presentation reviews the accident that occurred in 1985 when a bay bucket hoist fell from its stowed position. This accident damaged the orbiter, injured a technician, and delayed the launch. The accident was investigated, and the cause of the accident was determined to be the practice of two-blocking. Two-blocking is the result of hoisting beyond the intended safe upper limit of hook travel to the point of solid contact between the load block and the upper block or hoist/trolley structure. The usual result is immediate failure of the wire rope, due to the ropes being cut by the grooves of the drum or sheaves. The design of the hoist, the inspection, and the operator training were all in part responsible for this failure.

Lytle, Bradford P.↗

Comparing the Identification of Recommendations by Different Accident Investigators Using a Common Methodology

Accident reports play a key role in the safety of complex systems. These reports present the recommendations that are intended to help avoid any recurrence of past failures. However, the value of these findings depends upon the causal analysis that helps to identify the reasons why an accident occurred. Various techniques have been developed to help investigators distinguish root causes from contributory factors and contextual information. This paper presents the results from a study into the individual differences that can arise when a group of investigators independently apply the same technique to identify the causes of an accident. This work is important if we are to increase the consistency and coherence of investigations following major accidents.

Johnson, Chris W.↗

Generation IV Benchmarking of TRISO Fuel Performance Models Under Accident Conditions Final Report

The Generation IV International Forum (GIF) is a co-operative international endeavor of fourteen members organized to carry out the research and development needed to establish the feasibility and performance capabilities of the next generation nuclear energy systems. GIF selected six reactor technologies, amongst which is the Very High Temperature Reactor (VHTR) that is primarily dedicated to the cogeneration of electricity and hydrogen. The technical basis for VHTR is the tristructural isotropic (TRISO)-coated particle fuel, the graphite as the core structure, helium coolant, as well as the dedicated core layout and lower power density to removal decay heat in a natural way. At the heart of safety features of the VHTR concept lie the TRISO fuel particles that are designed to keep their structural integrity and retain fission products at temperatures up to 1600°C. As part as the design and future operation of VHTRs, a key aspect is the accurate prediction of fuel performance under irradiation and accident conditions. Modeling and simulation allow prediction of TRISO fuel behavior when subject to neutron flux and in high temperature accident scenarios. The refinement of the fuel performance models and codes is performed by comparison to in-pile and out-of-pile experimental data that reproduce the expected irradiation conditions in high temperature gas-cooled reactors (HTGRs). Historically, the International Atomic Energy Agency (IAEA) developed a benchmark dedicated to the validation of predictive methods for fuel and fission product behavior through the Coordinated Research Program CRP-2 (IAEA, 1997). CRP-2 was later updated to cover fuel fabrication, quality assurance, irradiation performance, safety testing, and spent fuel. The scope of the resulting CRP-6 benchmarks focused on HTGR fuel performance and fission product release (IAEA, 2012). Taking advantage of additional TRISO fuel fabrication, irradiation, and safety testing campaigns, GIF launched a Generation IV Benchmarking of TRISO Fuel Performance Models under Accident Conditions in late 2015. This GIF benchmark is a three-year program steered by Idaho National Laboratory (INL, USA). The other participants include the Japan Atomic Energy Agency (JAEA, Japan) and the Korea Atomic Energy Research Institute (KAERI, Korea). The objectives of the benchmark are to: follow on the IAEA CRP benchmarks, (2) model fission product release under accident conditions, (3) compare results obtained by the fuel performance modeling codes of the benchmark participants, and (4) compare these code predictions to experimental data. Safety tests chosen for modeling include the first and second experiments of the Advanced Gas Reactor program (AGR-1 and AGR-2) and the High Flux Reactor (HFR) EU1bis experiment. This report presents the results obtained by the three research institutions using their respective fuel performance modeling codes. Comparisons of the corresponding fission product release predictions are made with experimental data from AGR-1, AGR-2, and HFR-EU1bis. The benchmark results show good agreements between all participants but also show a general trend of over-prediction of the experimental release data, which is mainly attributed to the use of over-estimated diffusion coefficients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

PNNL Measurement Results for the 2018 FlatTop Criticality Nuclear Accident Dosimetry Exercise at the Nevada National Security Site (IER-253)

The Pacific Northwest National Laboratory (PNNL) participated in the FlatTop exercise IER-253 at the National Nuclear Security Site (NNSS) during the week of May 21, 2018. This report summarizes the measurements of the personal nuclear accident dosimeters (PNAD). Additionally, portable survey instrument readings and measurements of the simulated biological samples are included. The FlatTop exercise (IER-253) tested only PNADs and did not include fixed nuclear accident dosimeters (FNAD). Previous PNAD results obtained in 2016 at the Godiva-IV exercise IER-148 were summarized in PNNL Measurement Results for the 2016 Criticality Accident Dosimetry Exercise at the Nevada National Security Site IER-148 (PNNL-26497). The results clearly indicated that the previous dose conversions were a historical legacy of the Hanford Site and corresponded closer to kerma dose (or first collision dose, and not to Hp(10) dose). The technical basis for PNAD dose was based on the Godiva-IV exercise and documented in A New Dose Calculation Methodology for New PNAD and FNAD Designs at PNNL (PNNL-27023). The configuration of the PNAD was described in these two documents. The FlatTop exercise was the first time PNNL tested this new dose calculation methodology. Results were generally favorable with some results conservatively over reported that fell above the performance criteria. The reported gamma dose results fell non-conservatively low because only the default parameters obtained from the Godiva-IV exercise were used, and because the OSL gamma dosimeter has an energy dependency. The results from the simulated biological samples were also favorable, and in several cases conservatively over reported.

61 RADIATION PROTECTION AND DOSIMETRY↗

Accident-Mitigation-Based Safety Adequacy Evaluation at MSRs [Slides]

Accident mitigation based safety adequacy evaluation is anticipated to be less costly and time consuming than accident prevention for molten salt reactors due characteristics of the reactor class. The presentation provides an overview of the technical basis for developing accident mitigation-based regulatory guidance for future commercial molten salt reactor nuclear power plants.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

MECHANISTIC MULTIPHYSICS MODELING OF CLADDING RUPTURE IN NUCLEAR FUEL RODS DURING LOSS-OF-COOLANT ACCIDENT CONDITIONS

The Loss of Coolant Accident (LOCA) is a design basis accident that is included as part of the safety analysis of nuclear power plants. As the nuclear industry desires to increase the cycle length and discharge burnup of existing nuclear power plants they must demonstrate safe operation during a LOCA on high burnup fuel. During a LOCA transient on high burnup fuel rods, the rods may undergo a process known as fuel fragmentation, relocation, and dispersal (FFRD). To permit dispersal, the cladding encapsulating the fuel must rupture with an opening size large enough to allow the fragmented fuel particles to release. Current licensing tools used by industry and the United States Nuclear Regulatory Commission are limited in geometric fidelity and materials that can be analyzed. These simulation tools generally employ a quasi-two-dimensional (1.5D or Layered1D) or 2D-RZ axisymmetric geometric representations exclusively. While a valid approach under some instances, there are times when important physics have an asymmetric behavior in the fuel rod. Examples include fuel fragmentation, thermal-hydraulic boundary conditions, and cladding rupture, all of which are important for LOCA analysis. As industry pursues burnup extensions it must be demonstrated that fuel dispersal can be mitigated or eliminated. To do this, an understanding of the rupture opening after cladding failure is required. This work presents the development of a model for predicting the size and location of the rupture opening in failed fuel rods during LOCA conditions using advanced modeling and simulation tools. In order to supply the rupture model with appropriate boundary conditions, improvements to fuel fragmentation, axial relocation and oxidation modeling were required. First, the eXtended Finite Element Method (XFEM) is used to mechanistically predict the number of fuel fragments that form due to material strength randomization, criteria for strength randomization, mesh density, power ramping rates and irradiation effects. These predictions with associated uncertainty were compared to empirical correlations developed for UO2 verifying that they can be used with increased confidence in subsequent axial relocation analyses. Secondly, a new first-of-its-kind Layered2D computational framework was developed that provides the ability to apply azimuthally varying boundary conditions while providing discrete layers to track fuel movement during the LOCA. An existing fuel axial relocation model developed for Layered1D was extended to work within the Layered2D framework. A large sensitivity study was performed on the initial version of the model to identify modeling parameters of particular importance, with the emissivity used for radiation after blowdown being the primary source of uncertainty. Then, a simplistic approach to incorporate mechanical degradation of the cladding due to oxidation was also developed to investigate the impact of reduced cladding thickness on predictions of the time to failure of cladding tubes. It was found that the cladding will typically rupture prior to a reduction in thickness that is sufficient to impact the rupture behavior. A model was then developed for predicting cladding rupture that transfers the cladding surface temperatures, rod internal and external pressures, fast neutron fluence, and fast neutron flux from a more detailed Layered1D, Layered2D, or 2D-RZ analysis to a 3D cladding only analysis. Comparisons of the rupture model to a few experiments indicate reasonable predictions. The rupture model was then applied to two accident tolerant fuel concepts (FeCrAl and Cr-coated Zircaloy) where it predicted that both ATF concepts would have smaller rupture openings and delayed rupture times than the standard Zircaloy-4 cladding material under identical loading conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Effects of Cr/Zircaloy-4 coating qualities for enhanced accident tolerant fuel cladding

Cr-coated zirconium alloys represent a modern approach to enhance cladding safety during accident scenarios. Two high-power impulse magnetron sputtered Cr-coated Zry-4 systems were subjected to simulated loss-of-coolant accident conditions to investigate cladding performance. The first Cr-coating (4.8 µm thick) was deposited onto Zry-4 cladding and exhibited through-thickness cracking while the second Cr-coating (6.8 µm thick) was deposited with improved deposition parameters onto polished Zry-4 and exhibited no cracking. During burst testing, the coating with a higher density of defects failed to consistently reduce oxidation and exhibited similar burst behavior as Zry-4. In contrast, the second Cr-coating reduced ZrO 2 formation through formation of Cr 2 O 3 and displayed enhanced burst temperatures by ~80 °C compared to Zry-4. Utilizing an empirical relation for burst behavior of zirconium alloys, the 6.8 µm Cr/Zry-4 system displayed enhanced burst temperatures equivalent to an effective 0.464 mm increase in Zry-4 wall thickness, highlighting the value of continuous Cr coatings for accident scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

The response of accident tolerant fuel cladding to LOCA burst testing: A comparative study of leading concepts

Accident tolerant fuel claddings seek to improve safety margins during loss-of-coolant accident (LOCA) scenarios by reducing cladding steam oxidation rates, hydrogen production, delaying or mitigating cladding burst, and reducing cladding deformation. After rapid and extensive progress in recent years, Cr coated Zr-based alloys, ferritic FeCrAl, and SiC/SiC ceramic matrix composites (CMCs) have emerged as the most promising cladding concepts, but there are few studies comparing the accident tolerance of all three concepts simultaneously. Here, in this study, LOCA burst testing has been conducted on Cr high impulse power magnetron sputtered Zry-4 (Cr/Zry-4), C26M (FeCrAl), and SiC/SiC SiGA (CMC). Experimental observations indicate Cr/Zry-4 burst at higher temperatures than Zry-4 however the effectiveness of the coating on mitigating ZrO 2 formation via Cr 2 O 3 formation depended on physical proximity to the burst opening. No dramatic improvement in diametrical strain nor opening geometry was observed. C26M burst at higher temperatures than Zry-4 at lower pressures, exhibiting small amounts of strain and smaller openings, and formed slow-growing Al 2 O 3 regardless of location. The CMCs did not burst or show any macroscopic signs of deformation, all the way up to an internal cladding pressure of 15.2 MPa. High temperature expanding Nb plug testing indicated that CMC room temperature strength may persist up to 1650°C, with an approximately 40% loss of strength at 1900°C. CMC steam oxidation was also compared to reference chemically vapor deposited SiC up to 1700°C, and CMCs were found to oxidize at slightly accelerated rates that were still orders of magnitude lower than bare Zry-4. Projected H 2 generation during LOCA burst testing showed that when accounting for cladding burst exposing inner diameters to steam, Cr coated Zry-4 reduced H 2 generation by a factor of two compared with bare Zry-4, while both C26M and SiC/SiC reduced H 2 generation by orders of magnitude relative to bare Zry-4.

36 MATERIALS SCIENCE↗

Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

The QUENCH-19 experiment was a first-of-its-kind full-bundle test simulating accident conditions followed by water quench on accident-tolerant fuel (ATF) cladding. Here, a type of FeCrAl(Y) alloy, B136Y3, was developed at Oak Ridge National Laboratory and tested at the Karlsruhe Institute of Technology using Kanthal APM corner rods, a shroud, and Kanthal AF spacer grids. Testing conditions were similar to those in QUENCH-15—which tested ZIRLO cladding behavior—so that B136Y3 and ZIRLO cladding could be compared. QUENCH-19 consisted of an initial pre-oxidation heating followed by a transient. Then, a maximum power hold, which was not present in QUENCH-15, was executed to extend the heating period for the FeCrAl(Y) rods. Finally, a rapid water quench was executed that was similar to emergency core coolant system (ECCS) actuation. Compared with the ZIRLO rods in QUENCH-15, the bundle in QUENCH-19 released significantly less H 2 (9.2 g vs. 47.6 g) and achieved a much lower maximum temperature (1455°C vs. 1880°C). Furthermore, no breakaway oxidation was observed in QUENCH-19. Metallographic mounts revealed that despite the symmetry of the setup, at elevations near the maximum temperature, cladding and thermocouples were heavily damaged, substantial melting and oxidation occurred, and the cladding underwent chemical interaction with the thermocouple sheaths. Additionally, the ZrO 2 spacers detrimentally interacted with the cladding, leading to mixed oxide debris and the full destruction of some rods. Additional failure was found in certain cooler rods that may have risen due to the high thermal expansion coefficient of FeCrAl alloys. This paper presents an analysis of this work, which suggests that FeCrAl cladding can chemically survive anticipated loss-of-coolant accident events followed by rapid ECCS quench if the correct geometry and core design are present.

36 MATERIALS SCIENCE↗

Effect of ATF Cr-coated-Zircaloy on BWR In-vessel Accident Progression during a Station Blackout

The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal and transient. In this study, the performance of the proposed ATF concept of Cr-coated-Zircaloy is assessed using a generic Boiling Water Reactor MELCOR plant model considering a Short-term Station Blackout (STSBO) scenario. Simulation results indicate that the use of Cr-coated-Zircaloy as cladding and canister material mitigates the core degradation process as compared to the traditional Zircaloy cladding and canister design. The onset of fuel degradation and collapse is delayed by over thirty minutes, and the extent of fuel degradation is reduced. Specifically, the gross in-vessel hydrogen generation decreased by almost a factor of three. Although the eutectic reaction between Cr-coating and Zircaloy could cause an early failure of the coating, the improvement in the delay of fuel degradation is still notable. In addition, a thicker coating is found helpful to obtain additional coping time and to decrease hydrogen generation. In addition to the eutectic formation that may compromise Cr-coated Zr, a different failure mode is identified for the Cr-coated-Zr when compared to Zircaloy; i.e., a complete melt of base material leads to component collapse before the coating is oxidized and consumed. These findings can help the industry focus on productive areas of research and development for accident-tolerant fuel concepts and enhancement of core safety margins.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Development of the reactor core isolation cooling system model for the extended station black-out accident

In order to accurately predict the behavior of the Reactor Core Isolation Cooling System (RCIC) in the context of the FLEX strategies under the extended station black-out accident (SBO), we present a mechanistic model of the RCIC in this work. The control volume formulation of angular momentum conservation is employed to model the Terry turbine in the RCIC system. The two-phase nozzle flow model developed in the previous work is utilized to simulate the convergent-divergent nozzles in the Terry turbine. A nozzle flow table is generated to cover the operation range of the RCIC during beyond design basis accidents (BDBAs). The mechanistic RCIC turbine-pump model and the nozzle flow table are incorporated in a TRACE BWR model. Simulations of short-term and long-term SBO transients are performed to demonstrate dynamic RCIC responses. A system analysis code with the proposed mechanistic model of the RCIC system will have a better prediction of system safety response during SBO accidents. With the addition of such more realistic models, the system code will be capable of providing expanded understanding of safety margin, which can help help to form the technical basis for the successful implementation of the FLEX and SAMG measures under Extended Loss of AC Power (ELAP) conditions for BWR or PWR reactor designs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Material Interactions in Severe Accidents – Benchmarking the MELCOR V2.2 Eutectics Model for a BWR-3 Mark-I Station Blackout: Part II – Uncertainty Analysis

Single case comparisons between severe accident simulations can provide detailed insights into severe accident model behavior, however, they cannot offer insights into model uncertainty, sensitivity to uncertain parameters, or underlying model biases.Here in this analysis, the single case benchmark comparison of the MELCOR material interaction models for a station blackout (SBO) scenario of a boiling water reactor (BWR) using representative Fukushima Daiichi Unit 1 boundary conditions is expanded to include an uncertainty analysis. As part of this uncertainty analysis, 1200 simulations are performed for each material interaction model (2400 total), with random sampling of 14 uncertain MELCOR input parameters. Input parameters are selected for their impact on models representing core degradation processes. These include candling, fuel rod failure, debris quenching and dryout. The analysis performed here is not a traditional “best-estimate” uncertainty analysis that uses best-estimate parameters or identifies best-estimate figure of merit distributions. Instead, it is an exploratory uncertainty analysis that identifies and interrogates underlying model form biases of the two material interaction models (eutectics and interactive materials models). Uniform distributions are applied to all uncertain parameters to ensure coverage of the model parameter uncertainty space. Key findings from this study include underlying model form biases exhibited by material interaction models, and notable differences in accident progression outcomes between the material interaction models. This uncertainty study extends and confirms the conclusions from the first part of this study, which compared the impact of material interaction modeling on simulation of a short-term station blackout scenario with representative Fukushima Daiichi Unit I boundary conditions. In particular, this study confirms that the eutectics model generally exhibits accelerated degradation and failure of fuel components, the core plate, and the lower head. The eutectics model also has a tendency to exhibit a greater degree of core degradation, greater debris mass formation, and larger debris mass ejection. Finally, the eutectics model exhibits higher maximum temperatures for fuel, cladding, particulate debris, oxidic molten pool, and metallic molten pool components than the interactive materials model; interactive materials model simulations exhibit a soft “limitation” on maximum temperatures that is related to the temperature at which material relocation occurs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗