The need to develop more economical and safer nuclear reactors has led to the emergence of small modular reactors (SMRs), most notably the NuScale design, and the use of accident tolerant fuel, e.g. chromium-coated Zircaloy (Cr-coated Zr) cladding. The former (SMR design) is of interest because of its scale, cost, and passive cooling capabilities. The latter is of interest because of its characteristics of improved accident tolerance compared to the Zircaloy (Zr-based) cladding. In fact, researchers suggested that the implementation of Cr-coated cladding has the potential to provide more economic benefits for SMRs. However, our focus here is on studies exploring its role during accident progression.This paper investigates the significance of Cr-coated cladding for a loss-of-coolant accident under severe accident conditions. This investigation is performed by developing two NuScale plant models using the MELCOR-2024v code: (1) NuScale design with Zr-based cladding and (2) NuScale design with Cr-coated Zr cladding, where the oxidation model of the Cr-coated Zr cladding is captured using a homogeneous-material model approach.The presence of Cr-coated Zr reduces H2 generation by a factor of 4. The Zr-based model displays four distinct heating phases: (1) initial slow heating; (2) accelerated heating influenced by oxidation, ending with loss of the Zr-metallic form; (3) settling behavior due to radionuclide release outside the core and constant evaporation of the coolant; (4) final slow heating after complete dryout of the cooling channel. The Cr-coated Zr model differs significantly featuring a shorter Phase II and parabolic behavior in Phase III. To summarize, Cr coating is advantageous in reducing the overall amount of H2 generation and providing a longer coping time since Zr-metallic cladding is preserved for longer duration.

The University of Wisconsin (UW) is part of a team supporting General Atomics Electromagnetic Systems (GA-EMS) in its development of a 100-MWth gas-cooled fast modular reactor. In particular, the FMR uses the reactor vessel cooling system (RVCS) to passively remove the decay heat from the reactor pressure vessel. UW has developed an RVCS test facility for the GA-EMS modular high-temperature gas-cooled reactor and has performed past experiments to demonstrate its performance. This RVCS facility has been updated, and we have performed a series of repeatability tests under ISO (International Organization for Standardization) 17025 standards to validate the existing data set and our previous work. In addition, the UW team has developed a MELCOR model of the UW RVCS facility to analyze these tests and to simulate the two-phase natural circulation flow.This paper presents the recent tests and associated analyses that demonstrate good agreement with the data, except for the larger flow oscillation. The effects of the heat loss in the loop and the water tank pressurization were investigated, but no significant impact was shown. The detailed investigation revealed that the MELCOR simulation results in a higher void fraction (10% to 70%) compared to the experiments (5% to 40%) and an earlier start of flashing. We hypothesize that the water needs to be superheated to flash in the UW RVCS facility due to the slightly colder pipe structures with heat loss.

Deformation and failure of chromium (Cr) coated Zircaloy-4 (Zry-4) were studied in loss-of-coolant accident (LOCA) conditions using the BISON fuel performance code. The BISON validation model simulating Halden research reactor experiments was extended to include Cr coatings and higher rod internal pressures to simulate high-burnup fuel. The transient simulations show Cr coatings help relieve stress in the Zry-4 substrate during the transient, delaying the onset of high-temperature creep, which leads to ballooning and bursting of the cladding. A nominal Cr coating thickness of 30 μm delays clad failure by 26 seconds and increases the clad burst temperature by 40 K. A parametric study showed that time to failure and burst temperature both increase with coating thickness, and Cr-coated cladding offers burst resistance under a wide range of rod internal pressures simulating high-burnup fuel. Results indicate that a thin Cr coating provides resistance against ballooning and bursting of cladding during LOCA events.

Chromium (Cr) coated cladding is the most promising candidate for near-term accident-tolerant fuel (ATF) for current light water reactors (LWRs) because of its excellent oxidation resistance at high temperatures. For deployment of the Cr-coated cladding as ATF, the performance should be evaluated under various conditions, including the loss-of-coolant accident (LOCA). In this research, a single-rod reflood facility was constructed to investigate the quenching behavior and performance of Cr-coated cladding under bottom reflooding at high temperature conditions up to 1200 degrees C. Quench tests were performed using an uncoated, cold spray Cr-coated, and PVD Cr-coated Zr-alloy at various initial temperature ranges from 600 degrees C to 1200 degrees C with 20 K of water subcooling and 4.5 cm/s of reflood velocity. The quench temperatures were obtained at four different axial locations based on the maximum curvature method developed in this study. The bottom reflood quench test results indicate that the quench temperature increases with increasing initial cladding temperature and decreasing axial distance from the bottom. The quench temperature was not notably different for uncoated Zralloy and Cr-coated Zr-alloy, regardless of the coating deposition method. However, above 1000 degrees C, the Crcoated Zr-alloy showed a consistent and predictable quench temperature in contrast with the uncoated Zralloy for which the quench temperature data were severely scattered. The underestimation by the existing quench temperature model was observed with the current test using typical cladding thickness (0.57 mm). The thin cladding thickness was suggested as the reason for the underestimation. The modified Kim and Lee's correlation is proposed for better prediction of the thin-walled Zr-alloy cladding. The Cr-coating appears to mitigate the degradation of the mechanical integrity of the underlying Zr-alloy tube under severe oxidation and repeated thermal cycling, by virtue of superior oxidation resistance. It indicates that the coated cladding provides better integrity and performance in power transient conditions such as load-following operation. Post-test characterization showed the Cr-coating oxide layer thickness to be an order of magnitude lower than that of the uncoated cladding. The change in wettability was minimal in all test cases despite the trend of increasing roughness with the increasing initial temperature.

An experimental investigation to detect cavitation in the nozzles of working Diesel injectors was conducted. Cavitation behavior of Diesel injectors was characterized by a non-dimensional cavitation parameter and a coefficient of discharge. Transient behavior in Diesel injectors with different needle opening pressures and different numbersof nozzle holes was observed and measured. The behavior of sharp-edged single and multi-hole injector tips was found to be reasonably consistent with established characteristics of cavitating nozzles, as observed in steady-state experiments and as predicted by a one-dimensional model. The measurement of flow through a rounded multi-hole tip was consistent with the known behavior of non-cavitating nozzles. INTRODUCTION One important method of reducing emissions in Diesel engines is to improve fuel injector spray breakup, producing smaller and more disperse droplets. The flow inside the fuel injector nozzle is known to have a significant effect on the spray, but researchers have not discovered the exact nature of this effect [1]. Recent investigations have suggested that cavitation occurring within the fuel injector nozzle significantly affects spray breakup [2, 3]. However, much of what we know about cavitating nozzles has come from scaled-up models, with precisely determined geometry. Real fuel injector nozzles may have minute imperfections which can cause significant changes in the flow [4]. This investigation uses an experimental technique to indirectly detect the existence of cavitation in a variety of real injectors. The results of this technique should prove especially interesting in more complicated, multi-hole injectors. This experimental method will be applied to a single hole pump line injector and a multihole hydraulic electronic unit injector with sharp and rounded nozzle inlets. A One-Dimensional Model of Cavitating Nozzles Cavitation bubbles form because of the very low static pressure that occurs in high speed nozzle flow near a sharp inlet corner. This low static pressure is predicted by incompressible potential flow theory, which indicates that flow around a sharp corner, (e.g. a corner with a zero radius of curvature), will have infinite negative pressure. This physically impossible result is a direct consequence of the constant density restriction. In real injectors the fuel density decreases with decreasing pressure, most likely leading to a change in phase. The sharper the corner and the higher the velocity, the more likely cavitation is to occur. Sac Volume: Point 1 Contraction: Point c Cylinder: Point 2 Figure 1. Schematic of nozzle flow In the case of a sharp inlet, where the flow separates at the corner, the flow experiences a vena contracta. A diagram of the sharp entrance flow is shown in Fig. 1. Point 1 would be downstream of the injector needle and yet far enough upstream of the nozzle that the local velocity would be small, such as in the sac of the injector. Point c is downstream of the inlet, where the vena contracta effect is a maximum. In the case of a sufficiently rounded nozzle this point is nonexistent, in which case this analysis may not be useful. For convenience a ratio between the area at the contraction and the nominal nozzle area, known as the coefficient of contraction, is defined: Cc ≡ Ac A {1} Ac represents the effective flow area through the contraction and A represents the nominal nozzle area. The value of the contraction coefficient varies with the nozzle geometry and cavitation characteristics. For a very rounded entrance, the flow will not separate and the coefficient of contraction will be unity. For a short nozzle with a sharp entrance, conformal mapping by von Mises predicts a coefficient of contraction of 0.611 [5]. Experimental data seems to suggest that the coefficient of contraction is a constant with respect to Reynolds number, upstream pressure, and downstream pressure [6, 7]. Interestingly, the steady state coefficient of contraction for a sharp entrance seems to be around 0.61 for both cavitating and non-cavitating nozzles, as measured by Numachi [7]. When and how much the contraction area varies is very important and not well known. At increasingly high injection pressures the vapor region which bounds the contraction has been observed to elongate, apparently without further constricting the flow [8]. Another relevant integral property of the flow is the coefficient of discharge, Cd. The coefficient of discharge represents the efficiency of the nozzle between points 1 and 2 and thus is a measure of whatever losses occur in the nozzle The definition of the coefficient of discharge is: Cd ≡ ṁ A 2ρ P1 − P2 ( ) {2} In order to get closure for the one-dimensional model, an important assumption is made. We assume that the pressure at the point of contraction in a cavitating nozzle is equal to the vapor pressure, Pv. In this simplified view of the nozzle, all the losses are assumed to occur between c and 2. The mass flow rate behaves quite peculiarly under these assumptions. If the contraction pressure, Pc, is fixed at the vapor pressure, Pv, then the mass flow rate becomes independent of back pressure: ṁ = ACc 2ρ P1 − Pv ( ) {3} This peculiar behavior is similar to compressible choking, in that the mass flow rate depends only on the upstream pressure, but not on downstream pressure. This behavior was observed by Randall in cavitating venturi nozzles [9]. It is still unknown whether this flow is actually sonic because of the complexity of the twophase flow. We can combine the definition of Cd, continuity, and Bernoulli's equation to obtain the following expression for the coefficient of discharge of a cavitating nozzle: Cd = Cc P1 − Pv P1 − P2     1 2 {4} The pressure ratio in the right side of Equation 4 turns out to be a very useful cavitation parameter and is referred to as K in the remainder of this paper. K ≡ P1 − Pv P1 − P2 {5} Nurick plotted the coefficient of discharge versus the cavitation parameter, K on log-log axes in order to verify the square root dependence of Cd onK [6]. He observed that the data from the cavitating region lay on a straight line with a slope of one-half, where Cc is the value of the Y-intercept. At some point, the value of K is high enough that the nozzle no longer cavitates. The higher values of K occur when the difference between the upstream and downstream pressure is small. At high values of K the coefficient of discharge stays fairly constant or decreases with increasing K and thus falls to the right of the cavitating line. This variation occurs because the coefficient of discharge is no longer a function of K, but depends on the Reynolds Number instead. In order to further validate Nurick's findings, we have collected more data for sharp nozzles with an L/D ratio of about 4. The data come from real-scale experiments as well as scaled-up experiments and spans sixty years of research [8, 10, 11, 12, 13]. As shown in Fig. 2, this wide variety of sources tends to confirm Nurick's hypothesis. In an actual injector P1 represents the sac pressure. Unfortunately, it is extremely difficult to measure the sac pressure of a working injector. Instead, fuel injectors may be equipped with a pressure transducer just upstream of the needle. Because the exact sac pressure is not known, we must contend with the fact that our measured coefficient of discharge will actually include strong needle effects. During periods of low needle lift there will most likely be a large pressure loss across the needle [14]. For the beginning and ending portions of the injection, the needle will dominate the behavior of the coefficient of discharge. Due to needle effects, the measured value of K is only credible during large needle lift. For this reason, the bulk of the previous analysis should be applied to the portion of injection where the needle is nearly fully open. 0.5 0.6 0.7 0.8 0.9 1 1 2 3 Knox-Kelecy Hi royasu Reitz O h r n B e r g w e r k Gelalles T h e o r y C d

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO2) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA (R)) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. The reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of-forced cooling accident, which showed the peak cladding temperature at similar to 1600 degrees C during the transient, which is below the current design limit of 1800 degrees C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

Information obtained from Fukushima Daiichi Nuclear Power Station (Daiichi) is required to inform future Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. commercial nuclear power plants. This document summarizes results from the Fiscal Year 2023 (FY2023) U.S. effort to review Daiichi information and extract insights to enhance the safety of existing and future nuclear power plant designs. This U.S. effort, which was initiated in 2014 by the Department of Energy Office of Nuclear Energy, is completed by a group of experts in reactor safety and plant operations that identify examination needs and evaluate recent Daiichi examination data to address these needs. Fukushima-related information and associated discussions during these meetings benefit operating, new, and advanced reactors. Significant safety insights have been and are continuing to be obtained in several areas: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties related to severe accident modeling progression, these insights have and continue to be used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, Daiichi-related activities, such as code modeling improvements and analysis, testing, and new technology deployment efforts, have the potential to offer additional benefits to the operating fleet and new LWR and non-LWR designs. U.S. evaluations of obtained examination information and input regarding future Daiichi examinations are of interest to several organizations within Japan. Since its inception, the U.S. has provided consensus input for high priority time-sequenced examination tasks and supporting research activities. In their Mid-to-Long-term Examination Plan for 1F investigations, TEPCO included all remaining U.S. consensus information requests and additional information requests they identified. TEPCO periodically provides reports on the status of these requests (reflecting D&D priorities, new insights from investigations, and new technologies that become available). Hence, U.S. experts agreed that it was appropriate for TEPCO to track and prioritize these information requests as D&D progresses. U.S. experts will continue to review and comment on the information obtained from examinations and, as needed, provide additional details and relevant background material to support future examinations. As documented in this report, several other items, such as additional details on information requests pertaining to ex-vessel examinations, relevant references from prior research, additional documents to provide insights regarding recent investigation findings, and reviews of recently released documents, were agreed to during the FY2023 meeting.

Information obtained from Fukushima Daiichi Nuclear Power Station (Daiichi) is required to inform future Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. commercial nuclear power plants. This document summarizes results from the Fiscal Year 2022 (FY2022) U.S. effort to review Daiichi information and extract insights to enhance the safety of existing and future nuclear power plant designs. This U.S. effort, which was initiated in 2014 by the Department of Energy Office of Nuclear Energy, is completed by a group of experts in reactor safety and plant operations that identify examination needs and evaluate recent Daiichi examination data to address these needs. Fukushima-related information and associated discussions during forensics meetings benefit operating, new, and advanced reactors. Significant safety insights are being obtained in several areas: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties related to severe accident modeling progression, these insights have and continue to be used to update guidance for severe accident prevention, mitigation, and emergency planning. As discussed in this document, revised operator guidance was successfully used to improve operator response during a loss of off-site power event at the Duane Arnold Energy Center plant. Reduced uncertainties in modeling the events at Daiichi improve the realism of reactor safety evaluations that inform future D&D activities. U.S. evaluations of information from Fukushima and input regarding future examinations are of interest to several organizations within Japan. Meeting presentations by Japan describe how comments and recommendations documented in prior U.S. forensics effort reports, including consensus information requests developed by forensics effort participants, are considered in future Fukushima D&D activities. As discussed in this report, TEPCO Holdings considered these information requests in their D&D planning activities. An updated list of consensus information requests is included in this FY2022 report.

Since the accident at Fukushima, one major goal of reactor safety research has been the development of more accident tolerant technologies that can mitigate or delay fuel degradation during a Beyond Design Basis Accident (BDBA). One major effort has been focused on increasing the capability of the fuel to be more tolerant of damage during an accident, i.e., Accident Tolerant Fuel (ATF) materials. In this work, we present the development of a generic BWR plant model, the modification of MELCOR to model ATF materials and the use of ATF materials (specifically FeCrAl alloy) as a coating on Zircaloy cladding or as a substitute material for cladding and fuel assembly canister material and its effect on severe accident progression, specifically, a Station Blackout accident. The analysis indicates that significant fuel degradation via fuel heat-up, clad oxidation, and hydrogen generation was delayed up to an hour if FeCrAl alloy was used as a clad and canister material. And, combined with the passive safety systems (i.e., the Reactor Core Isolation Cooling system, RCIC), the extended operation of these systems delayed fuel degradation further. However, an adverse effect should be emphasized for the monolithic FeCrAl design-it generated more hydrogen than the designs based on the Zircaloy due to the high reaction rate at a high temperature of FeCrAl. The design of the FeCrAl-coated-Zircaloy avoids this defect. Therefore, it is a promising choice to combine some of the beneficial traits of both materials.

In recent years, micro-reactor concepts have attracted increasing attention in the nuclear industry due to the market demand for flexible, reliable, and sustainable power and heat on-site for industrial or federal installations or remote communities. To help demonstrate and validate these innovative reactor concepts, the Micro-reactor AGile Non-nuclear Experimental Test-bed (MAGNET) is being constructed at Idaho National Laboratory (INL) with an initial focus on the thermal and structural performance of heat pipe cooled micro-reactors. At this time, the preliminary design parameters of the MAGNET facility have been specified. In this work, a simulation using the System Analysis Module (SAM) code is performed for the prototypical 37-heat-pipes test article to predict its experimental facility performance and associated uncertainties. We first carry out a benchmark demonstration of our modeling method with an example provided by Argonne National Laboratory (ANL). Then, we predict the thermal performance of the MAGNET facility under steady-state operation. Moreover, several sensitivity parameters are analyzed to investigate their impact on facility thermal performance. This MAGNET experiment simulation provides valuable information for researchers to validate the facility's initial design and steady-state operation.

Recently, microreactor designs have been receiving significant attention in the nuclear industry due to their potential advantages in certain applications. These nuclear reactor designs have been considered to provide reliable and sustainable power for on-site installation and operation. Microreactors may be utilized to provide heat and power to hydrogen production, remote communities, and industrial facilities such as military installations, disaster relief zones, and are being considered for underwater and deep space operation as well. However, these designs and concepts remain largely untested and unproven in the commercial industry. Further research and development are still required to prove microreactor designs are safe and reliable for commercial use. Different cooling technologies have been taken into consideration for microreactor concepts since the 1960s, mainly for federal space reactor projects such as LEGOLRCS, HOMER, and KRUSTY. This work provides thermal hydraulics and analysis for the Idaho National Laboratory's MAGNET (Micro-reactor Agile Non-nuclear Experimental Testbed) facility. The MAGNET facility is currently being developed to duplicate a microreactor design using heat pipe cooling technology. Our main goal is to examine the response of the test facility under steady-state and transient operation conditions. We constructed our own estimated model of the MAGNET geometry using a software coupling method made up of MOOSE/SAM software systems. The steady state results of this work have been published in a former article. The new results mainly focus on the transient states. By communicating with the Idaho National Laboratories, we upgraded the geometry of MAGNET heat pipes. This not only verifies the design of the facility under such conditions but also benchmarks the modeling capability of the MOOSE/SAM code system that can be potentially used to model other microreactor concepts in the future.

Much is still not known about the end-state of core materials in each unit that was operating on March 11, 2011 at the Fukushima Daiichi Nuclear Power Station (Daiichi). Information obtained from Daiichi is required to inform Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. commercial nuclear power plants. This document summarizes results from the Fiscal Year 2021 (FY2021) U.S. effort to review Daiichi information and extract insights to enhance the safety of existing and future nuclear power plant designs. This U.S. effort, which was initiated in 2014 by the Department of Energy Office of Nuclear Energy (DOE-NE), is completed by a group of experts in reactor safety and plant operations that identify examination needs and evaluate recent Daiichi examination data to address these needs. Since its inception, annual reports were issued that document significant safety insights being obtained in areas of special emphasis: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties related to severe accident modeling progression, these insights have and continue to be used to update guidance for severe accident prevention, mitigation, and emergency planning. Reduced uncertainties in modeling the events at Daiichi improve the realism of reactor safety evaluations that inform future D&D activities. A key aspect of prior U.S. efforts, the updated list of information requests, is included in this FY2021 report to ensure that they are transmitted to organizations within Japan. This report also continues to emphasize how information obtained from the affected reactors at Daiichi has been and will continue to be used to update severe accident management strategies and reduce uncertainties in systems analysis code models. In addition, recommendations are included that would expand the use of this information to provide insights regarding maintenance, radiation protection, design, and siting activities for existing and new reactors.

In reaction to the Fukushima Daiichi accident, efforts are underway by the DOE and industry to develop fuel cladding that is more oxidation resistant and hence more accident tolerant than the presently used zirconium alloys (Zr-alloys). Two accident tolerant fuel (ATF) cladding strategies are being pursued – a long-term approach involving the replacement of Zr-alloys with other cladding materials such as SiC-SiCf (SiCf stands for SiC fiber) composite and FeCrAl, and a near-term approach that involves coating commercial Zr-alloy cladding material with oxidation-resistant materials. As a part of the ATF development efforts, it is important to understand the role of the surface characteristics of these ATF cladding designs on heat-transfer behavior. In this regard, the critical heat flux (CHF) phenomenon is an important metric as it represents the thermal limit of the high heat transfer rates during boiling and is an important figure of merit defining LWR safety margins. In the proposed research, CHF phenomenon and physics of the boiling behavior will be investigated for various ATF cladding material concepts that are at the forefront of consideration by DOE and industry. These will include materials being considered as structural replacements for Zr-alloys, namely, SiC-SiCf composite and FeCrAl. The surface coating approaches being developed at the University of Wisconsin will also be tested. These will include Cr and FeCrAl coatings deposited on a Zr-alloy by the powder spray process, and coatings deposited by the physical vapor deposition (PVD) process. Both pool boiling and quench experiments will be performed to generate key data that can be used to screen and compare new ATF materials to the presently used Zr -alloys. This will expand the knowledge-base on the interrelationships between CHF physics and materials’ surface characteristics. To quantify these importance of materials’ surface characteristics on CHF and boiling behavior, a suite of materials analysis techniques including, optical measurement for contact angles, profilometry, atomic force microscopy, and electron microscopy will be used to characterize the surfaces in order to effectively correlate materials’ surface characteristics to CHF and boiling behavior. Additionally, the thermal properties of the ATF cladding materials will be measured and correlated to CHF data. Experiments will be performed using various surface roughnesses as well as for the as-received samples and samples oxidized in prototypical water autoclave conditions. With this basic understanding, the next task is to conduct reactor prototypical tests under flowing pressurized water conditions at the facilities of our industrial partner, Westinghouse. These experiments will be important as actual heater rods clad will undergo prototypic testing at PWR flows, pressures and temperatures. This prototypic test data will then be used to model boiling behavior for various ATF cladding materials using computer codes such as TRACE and COBRA-TF.

For advanced research reactor designs that have a high core power with downward coolant flow in the core, it is important to analyze a unique postulated pipe break scenario under sub-atmospheric pressures. In order to benchmark the capability of thermal-hydraulic codes such as TRACE to predict flow behavior under these conditions, we must first compare TRACE analyses to reference experimental data for air-water flow near atmospheric pressure and temperature conditions. Given a successful comparison between predictions and test data, we then use TRACE to analyze air-water two-phase flow phenomena during a postulated pipe break accident with the initial conditions prototypic of a particular research reactor design. The normalized flow rate, pressure, and void fractions were predicted to describe the overall behavior of the pipe break phenomena. These results indicate that the amount of air ingested from the environment into the reactor coolant system through the pipe break has the dominant effect on the downstream void fraction as well as flow and pressure response, and thereby the time available (delay time) to take compensatory actions before deleterious effects are felt by the reactor cooling system. This defined delay time is the figure of merit that can be used for design of the system. Since the two-phase flow under sub-atmospheric pressure in the system can be affected by various design conditions, as well as initial and boundary conditions, the sensitivity analyses were performed to better understand their effect. A set of parameters were chosen based on their quantitative effect on the overall transient behavior; i.e., pool inventory, system design flow rate, and pipe break size. The various quantitative characteristics of the sensitivity analysis results showed that TRACE was capable of analyzing this unique air-water two-phase flow phenomena under sub-atmospheric pressure and will be utilized in the safety analysis for this research reactor system.

Direct containment heating is a phenomenon identified to occur in a light water reactor during a meltdown accident at high system pressures at the time of vessel breach. Immediate containment failure may result from this process due to overpressure and overtemperature loads. This paper focuses on the “science” of direct heating and the reactor safety criteria and issues that must be addressed to assure the threat to containment is minimized. This discussion attempts to be tutorial in nature with the broader context of melt-cool ant-atmosphere i nteracti ons being emphasized. From this discussion future research needs may be suggested.

•The uncertainty in the use of ATF Cr-coated-Zr cladding during STSBO is quantified.•The high temperature Zr oxidation rate has the major influence on the selected output parameters.•Cr-coated-Zr cladding can delay rapid clad oxidation and generation of hydrogen by 1–2 h.