Reactor Accident Research Articles

Recall of residential history and dietary habits during pregnancy and lactation in the distant past: reliability of questionnaire-based radiation doses for persons exposed in utero and early life.

This study evaluates the reliability of information obtained by standardized questionnaires used in by personal interviews for estimation of radiation thyroid doses of 1065 individuals in the Belarusian cohort of individuals who were exposed in utero and early life following the Chernobyl accident in April 1986. Data from two interviews conducted in 2012-2017 and in 2018-2022 with mothers, who were pregnant or gave birth shortly after the Chernobyl accident, were analysed. The most reliable answers dealt with various attributes related to residential history. In contrast, the reliability of answers regarding consumption rates of milk from privately owned cows or trade network was moderate, while the agreement in responses for consumption of milk products and leafy vegetables was fair. Information from the two interviews was used to calculate thyroid doses received by the cohort members. Specifically, 'model-based' thyroid doses due to 131I were estimated using input data on individual residential history and food consumption reported during the personal interviews and ecological data (131I ground deposition in the corresponding settlements). In addition, for a subset of cohort subjects (n = 205) whose mothers were measured for 131I thyroid activity, 'measurement-based' thyroid doses were calculated by adjusting the model-based dose using a scaling factor that is defined as the ratio of measured 131I thyroid activity to model-based 131I thyroid activity calculated for the date of measurement. A moderate agreement was observed for total (prenatal and postnatal) model-based thyroid doses due to 131I intake, the arithmetic mean ± standard deviation for the Jaccard similarity coefficient ([Formula: see text]) was 0.45 ± 0.34 (median = 0.39), while measurement-based doses showed a much better agreement with a [Formula: see text] of 0.78 ± 0.29 (median = 0.93). For model-based thyroid doses from external irradiation and from ingestion of 134Cs and 137Cs, [Formula: see text] was 0.82 ± 0.23 (median = 0.90) and 0.84 ± 0.24 (median = 0.96), respectively. Measurement-based doses due to ingestion of radiocaesium isotopes resulted in an almost perfect agreement, [Formula: see text] was 0.91 ± 0.19 (median = 1.0). The present findings suggest that long-term memory recall can be reliable, if a person is asked about unique or important life events, such as pregnancy and childbirth occurring around the time of a nuclear reactor accident. However, the substantial difference (more than 10 times) observed for model-bases doses calculated using the two questionnaires represents an important source of human factor uncertainties that needs to be considered in any dose response analyses. Other lessons learned from this study are that (i) individual measurements of radionuclides in the human body are the most valuable source of information for estimating radiation doses, and (ii) whenever a radiation accident occurs, a sample of affected people should be asked to keep a diary, if at all possible.

Influence of potential contaminants on I2 and CH3I adsorption onto zeolitic imidazolate frameworks (ZIFs) using GCMC simulations

Zeolitic imidazolate frameworks (ZIFs) are one type of promising volatile radioiodine sorbents due to their chemical/thermal stability, tunable structures, and high iodine adsorption capacity. However, the inhibiting effect of contaminants on volatile iodine adsorption onto ZIFs remains unclear. Herein, the adsorption of volatile iodine (I2 and CH3I) onto five different ZIFs (SALEM-2, ZIF-8, ZIF-Cl, ZIF-90, and ZIF-65) and adsorption selectivity over ten contaminants were investigated by using grand canonical Monte Carlo (GCMC) methods. It is found that the I2 adsorption isotherm could be accurately described by the Langmuir-Freundlich model. Furthermore, the adsorption quantity of CH3I is much smaller than that of I2 at the same partial pressure. Ten potential gaseous contaminants that existed in a reprocessing plant or nuclear reactor accident, including CO, NO, H2O, CH4, C2H6, CH3OH, CH3CH2OH, Cl2, CH3Cl, and C6H6, were considered for calculation of binary adsorption. The results show that C6H6 has the largest adsorption enthalpy among all contaminants, whereas CO, NO, H2O, and CH4 have the smallest. Moreover, CO, NO, H2O, and CH4 have negligible effects on volatile iodine adsorption onto ZIFs. In contrast, except for CH3I adsorption onto ZIF-65, C6H6 has the strongest inhibiting effect since benzene could occupy the adsorption sites of iodine in ZIFs. This work not only provided the fundamental knowledge of volatile iodine adsorption on ZIFs but also proposed a new strategy to evaluate the inhibiting effect of contaminants. It could be helpful for future screening and design of better sorbents.

Study of contact melting of plate bundles by molten material in severe reactor accidents

In a severe reactor accident, a crust will form on the surface of the molten material during the core melting process. The crust will have a contact melting with the internal components of the reactor. In this paper, the contact melting process of the molten material on the austenitic stainless steel plate bundles is studied. The contact melting model of parabolic molten material on the plate bundles is proposed, and the rule and main effect factors of the contact melting are analyzed. The results show that the melting velocity is proportional to the slope of the paraboloid, the heat flux and the distance between two plates D. The influence of melt gravity and the plate width on melting velocity is negligible. The thickness of the molten liquid film is proportional to the heat flux and plate width, and it is inversely proportional to the gravity. With the increase of D, the liquid film thickness decreases at first and then increases gradually. The liquid film thickness has a minimum against D. When the width of the plate is small, the width of the plate is the main factor affecting the thickness of the liquid film. The parameters are coupled with each other. In a severe reactor accident, the wider internal components of reactor, which can increase the thickness of the melting liquid film and reduce the net input heat flux from the molten material to the components, are the effective measures to delay the melting process.

The Effect of Fuel Aerosol Particle Size on Aerosol Release Potential in Energetic Core Disruptive Accident Studies

In nuclear reactor accident safety studies, the radiological source term is a metric that quantifies the release of radiological material from the reactor to the environment. The present work evaluates heat transfer between high-temperature vapor bubbles and the surrounding coolant and the effect these interactions have on the source term for postulated core disruptive accident scenarios associated with an oxide-fueled, liquid metal–cooled fast reactor class. It is shown that aerosol particle size can influence heat transfer, and it is suggested that the extent of the influence depends on the fineness of the particles in the aerosol. The results are consistent with legacy experiments conducted in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory and offer a more comprehensive assessment of vapor condensation by treating the bubble constituents, in the context of radiation heat transfer, as participating media. The model, which couples classical scattering theory to the equation of radiative transfer and the energy equation, provides a means for estimating size-affected radiative cooling times. Solutions are obtained via the P-1 method of spherical harmonics with improved, higher-order boundary conditions. Outcomes include the development of an “extinction-time ratio” criterion for assessing whether ejection of aerosol from the bubble to the cover region is likely. Aerosol release from the coolant pool is evaluated using this criterion with the potential to extend this work to reactor-scale accidents. A baseline evaluation is provided that shows that omission of participatory effects could lead, in a relative sense, to cooling time offsets in excess of 14%. In addition to enhancing previous evaluations of FAST results, these modeling outcomes contribute to knowledge management efforts aimed at developing a more mechanistic assessment of the source term while suggesting potential enhancements to severe accident safety analysis through the use of more comprehensive radiative heat transfer models.

Lumped parameter model for the safety assessment of the IAEA 10MW benchmark reactor during a protected loss of flow accident

The loss of flow accident in the IAEA 10Mw benchmark reactor is treated in this study through a coupled solution of the point kinetic and thermal hydraulic models. By employing the lumped parameter technic, a one-dimensional thermal hydraulic model is derived and solved beside the point kinetic model with the continuous reactivity feedback effect of the coolant and fuel temperatures. This coupled solution, allows us, to determine, the transient variation of the maximum fuel and coolant temperatures, furthermore, to the safety criteria (ONBR, OFIR, CHFR and BOR) which must be greater than the imposed safety limits, to avoid any nuclear reactor safety issue. Finally, the obtained results of our simulation were validated after a satisfactory comparison with the results of other codes system and developed computer codes.

Numerical Study on Relocation Process of Al, Fe, and Pb by Using the Moving Particle Semi-Implicit Method During Severe Accident of Reactor

Numerical Study on Relocation Process of Al, Fe, and Pb by Using the Moving Particle Semi-Implicit Method During Severe Accident of Reactor

Hierarchical Bayesian modeling to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

For realizing a highly reliable fracture limit evaluation of fuel cladding tubes during loss-of-coolant accidents (LOCAs) in light-water reactors, we developed a method to quantify the fracture limit uncertainty of high-burnup advanced fuel cladding tubes. This method employs a hierarchical Bayesian model that can quantify uncertainty even with limited experimental data. The hierarchical Bayesian model was an extended version of the fracture probability estimation model developed in our previous study. The fracture limit uncertainty was quantified as a probability using the amount of oxidation (Equivalent cladding reacted: ECR) and the initial hydrogen concentration (the hydrogen concentration in the fuel cladding tubes before the LOCA-simulated tests) as explanatory variables. We divided the regression coefficients of this model into a hierarchical structure with an overall average term common to all types of fuel cladding tubes and a term representing differences among various types of fuel cladding tubes. This hierarchical structure enabled us to quantify the fracture limit uncertainty through the effective use of prior knowledge and data, even for high-burnup advanced fuel cladding tubes with a small number of data points. The fracture limits representing a 5% fracture probability with 95% confidence of the high-burnup advanced fuel cladding tubes evaluated by the hierarchical Bayesian model were higher than 15% ECR for the initial hydrogen concentrations of up to 700–900wtppm and restraint loads below 535N. These fracture limits were comparable to the limit of the unirradiated Zircaloy-4 cladding tube, indicating that the burnup extension and use of the advanced fuel cladding tubes do not significantly lower the fracture limit of fuel cladding tubes. Further, we proposed a method to reduce the fracture limit uncertainty by using non-binary data, instead of the binary data, depending on the condition of the fuel cladding tube specimens after performing the LOCA-simulated test, thereby increasing the amount of information in the data.

Challenges to the integrity of calandria, the CANDU PHWR moderator vessel, under serious accident loads and consequences of its failure

This paper summarizes the technical and analytical basis for a more advanced examination of structural integrity of a CANDU PHWR Calandria vessel under certain reactor accident conditions. Of foremost interest is the early Boiling Liquid Expanding Vapour Explosions (BLEVEs) induced challenge to vessel structural integrity following a spontaneous in-core rupture of a fuel channel. A sudden, flashing jet discharge of over 100 kg/s of the 10 MPa, 300 °C saturated primary coolant into its 70 °C subcooled near atmospheric pressure moderator water inventory will cause an explosive, shock wave impact. This results from rapid nucleation and vaporization of the discharged superheated liquid. Consequential pressure pulse, pipe-whip and bundle debris impact will induce damage to adjacent channels and cause widespread failure of fuel channel calandria tube rolled joints. These thin tubes inevitably deform to create serious challenges to vessel integrity. On the other extreme of reactor accidents, is the matter of an in-vessel-retention (IVR) of high temperature molten fuel debris in the reactor vessel after a severe core damage accident, such as one caused by an unmitigated station blackout. This scenario has understandably drawn greater attention after Fukushima and potential IVR in the > 200 mm thick PWR vessels when partly submerged in water is viable.It is highly unlikely that we will see reactor debris melt retention in this 19 to 29 mm thick, stepped shell with a tapestry of welds. This is because partial shell melting, diminished strength beyond 650 °C across 2/3 of the thin shell thickness and severe thermal stresses adds to underlying residual stresses near the welded seams. The result is inevitable catastrophic off-site consequences as energetic interactions upon vessel breach of water with hot debris can cause additional steam & hydrogen explosions, containment failures, and large off-site releases.This paper presents technical arguments to illustrate the problem in accepting faulty assumptions conveniently made in licensing safety assessments to simplistically claim survival of the thin-walled welded structure under these two illustrative accidents. I propose pathways to further investigations. Therefore, the issue of Calandria vessel survival under a wide range of accident conditions must be given a second, thoughtful look. Extending boundaries of a safe operating envelope as pressure tubes age needs to be avoided. We need to be more realistic in our severe accident analysis and consequence assessments. Fukushima taught us that the consequences of continuing to rely on convenient but technically faulty rationale, along with hype of eternal safety, can be very unforgiving.

Spatio-temporal evolution of sodium aerosol and temperature distribution during sodium pool fire in a confined environment

In case of a Core Disruptive Accident in sodium-cooled fast reactor, sodium will be ejected from the reactor vessel into the Reactor Containment Building (RCB) and results in sodium fire. Liquid sodium combustion generates aerosol, which gets dispersed and remains suspended in RCB based on geometry of the building and degree of turbulence. The Environment Source Term (EST) primarily depends on the dynamics of sodium aerosols inside the containment. Hence, it is essential to carry out a safety analysis of SFR using computational tools and models shall be validated with experimental data, particularly aerosol dynamics inside the containment along with thermal hydraulics. Besides, all codes assume uniform aerosol concentration in large containment buildings and the evolution of suspended concentration was predicted. However, in actual case a non-uniform concentration profile exists, which leads to stratification of aerosol concentration inside the containment. Towards assessment of aerosol dynamics in realistic scenario, spatial–temporal characteristics of sodium aerosol and temperature distribution were studied by conducting sodium pool fire experiments in a large containment.The sodium fire experiment was conducted by using ∼2.0 kg of sodium in a pool area of ∼0.25 m2. The measured maximum pool and gas temperatures are 685 °C and 126 °C respectively, and the temperature gradient at the wall boundary layer was 7.5 °C/cm. The estimated average sodium burning rate and aerosol release rate are 17.3 kg/m2hr and 0.24 g/s, respectively. The spatial variation of aerosol characteristics was measured for both horizontal and vertical planes. The measured maximum aerosol concentration in the horizontal plane ranges from 2.76 to 5.32, 2.16–3.30 and 2.09–4.21 g/m3 for the bottom, middle and top elevations respectively. For the vertical plane, aerosol concentration varied from 2.16 to 3.65 g/m3 for central positions and 2.53–5.32 g/m3 for off-center positions. The trend of average suspended aerosol concentration at each elevation showed that the concentration was higher at the top elevation than at the middle and bottom for up to 15 min i.e., during aerosol generation period and after that concentration was found to be higher at the bottom elevation for the next 25 min. Based on the measurements, it is observed that the aerosol distribution is not well mixed within the experimental period. A significant difference in aerosol mass concentration was observed between each elevation, between estimation and observations based on a zero-dimensional model. The evolutions of aerosol median size by theoretical simulation capture the overall trend reasonably.

Comparison of Melted Corium Relocation during Severe Accident of High Temperature Reactor using Moving Particle Semi-Implicit Method

System failure in nuclear reactors can cause degradation of a reactor core, allowing melting and relocation of the corium to the lower plenum in the nuclear reactor system. In this study, a severe accident simulation was carried out using the Moving Particle Semi-Implicit (MPS) method. In this method, we model the relocation of molten corium on the reactor core (support plate) to the lower plenum for several conditions with variations: corium material, lower plenum conditions, temperature, viscosity, and density. Those treatments were carried out in order to be able to compare and analyze the characteristics of the corium melt by reviewing the velocity profiles. The formation of a corium pool and debris bed can result in significant temperature differences and high heat flux against the walls of the reactor vessel, causing a decrease in the integrity of the reactor vessel and reactor failure.Keywords: Corium, Uranium Dioxide (UO2), Zirconium Dioxide (ZrO2), Fluid Relocation, Moving Particle Semi-Implicit (MPS).

Investigation of plate fuel performance under reactivity initiated accidents with developed multi-dimensional coupled method

A reactivity-initiated accident is a typical nuclear reactor accident during which the core fission rate and power increase unexpectedly, making the temperature of the fuel element soar, which causes a decreasing strength, large deformation, and even the failure of the element. To simulate the performance of the plate-type fuel during RIA, a neutronics–thermal hydraulics–mechanics coupling method was implanted in a fuel performance code BEEs-Plates. The coupling scheme was firstly verified and validated against the MTR benchmark. Then the RIA results with the consideration of the deformation effect were compared with the benchmark case, showing that the mechanics as well as the burnup effect have little impact on the neutronic–thermal hydraulics results when using the constant thermal properties. Lastly, the thermal-mechanical analysis was conducted for the fuel plate at different burnups (0% FIMA, 10% FIMA, 20% FIMA) under the RIA. The results show that the peak stress of the cladding increases with the burnup increases while it has an opposite trend for the fuel meat. Some special features such as over-shrinkage of the fuel meat have also been noticeable and analyzed.

SIMMER modelling of accident initiation phase in sodium fast reactors

The SIMMER-III code was developed mainly for simulation of hypothetical reactor accidents after core melting, but can be applied also for accident initiation phase simulations in sodium and other liquid–metal-cooled fast reactors. New thermal hydraulic and neutronic simulation approaches and models have been developed for treatment of the initiation phase. Two examples of simulation of unprotected loss of coolant flow (ULOF) transients, in ESFR-SMART and FFTF reactors, are presented. It can be concluded that the SIMMER code has a large potential for application to initiation phase analyses.

Development of reduced-order thermal stratification model for upper plenum of a lead–bismuth fast reactor based on CFD

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

CFD Uncertainty Quantification using stochastic spectral methods—Exemplary application to a buoyancy-driven mixing process

The consideration of uncertainties is of particular importance for nuclear reactor safety, where high safety standards for example ensure the integrity of the containment. By means of Computational Fluid Dynamics (CFD), buoyancy-induced mixing processes, which can take place during a severe reactor accident, are investigated. However, the CFD models contain uncertainties, which have a large impact on the present flow and have to be analyzed. The method development for the subsequent Uncertainty Quantification of a representative reactor test containment is conducted using the Differentially Heated Cavity of aspect ratio 4 with Ra=2×109 as a generic test case from the literature. In this way, methods for the future quantification of uncertainties in large-scale industrial applications are established. Results from single-fidelity models such as Unsteady Reynolds-Averaged Navier–Stokes, Large Eddy Simulation and Direct Numerical Simulation are presented and combined to three-level multifidelity models. Stochastic representations of scalar random responses are constructed by means of Polynomial Chaos Expansions and Karhunen–Loève Expansions are derived for the representation of stochastic processes. A new approach for the description of highly dynamic transient processes called Random Field Composition (RFC) is presented, which proposes stochastic model construction through combination of multiple random fields. The presented multifidelity (MF) models achieve high accuracy in combination with a justifiable computational effort. Therefore, MF modeling serves as a promising approach for large-scale applications. Furthermore, it is found, that RFC allows for the description of highly dynamic processes with a reasonable number of simulation runs, if the complexity of the stochastic process representation can be reduced by partitioning the stochasticity into single random fields.

Feeder Pipe Oxidation in the Presence of Steam During a Nuclear Reactor Accident

A Canada Deuterium Uranium (CANDU) reactor has hundreds of carbon steel feeder pipes that are connected at one end to the fuel channels at the reactor face and are connected at the other end to the reactor inlet and outlet headers above the reactor. The production of hydrogen gas from metal–water interaction at high temperatures is a major concern during a severe accident in a nuclear reactor. It is generally accepted that the main source of hydrogen gas during a severe accident is the chemical interaction of Zircaloy fuel cladding and the water coolant. However, it has recently been suggested that the amount of hydrogen produced by the oxidation of carbon steel located outside of a CANDU core could exceed that produced by zirconium oxidation. In this work, the linear and parabolic oxidation rate constants of CANDU feeder carbon steel, in the temperature range of 600–1,100°C, were measured to be W/t = 1.806 × 103 exp (− 126,337/RT) and W2/t = 1.879 × 105 exp (− 135,835/RT), respectively. For parabolic steam oxidation beginning at roughly 90 minutes after reaching steady-state heating temperatures, the latter equation should be used. These rates are about a factor of 5–10 greater than oxidation rates measured in 304L stainless steel and in zirconium specimens, meaning that the carbon steel oxidation can be significant in an accident.

Development and verification of helium–xenon mixture cooled small reaction system

Although Helium–Xenon mixture cooled small reactor has been widely concerned, there is no mature thermal calculation software that can be applied to Helium–Xenon mixture cooled small reactor in China. In this paper, a semi empirical formula used to predict the physical properties of binary inert gas mixtures is used to establish a physical property calculation model of Helium–Xenon mixture gas mixtures, and the flow and heat transfer calculation model, turbine model and compressor model are modified and improved, so as to achieve the calculation of the Brayton cycle system. The function of the original system analysis program is expanded and improved, and the simulation program for he xenon mixed gas reactor system is developed. At the same time, the closed Brayton cycle system is established by using the system analysis program developed in this paper, and the calculation results are compared with the design values. The maximum relative error is 6.9%. The steady-state operation results of the model are in good agreement with the design operation parameters, which proves the applicability of the improved simulation model. The steady-state response characteristics analysis, the introduction of positive and negative reactivity accidents and the calculation of transient conditions such as step load drop of the reactor were carried out. The overall trend is the same as the literature survey results. The results show that the system analysis program developed in this paper is suitable for the analysis of the Brayton cycle system of Helium–Xenon mixture cooled small reactor.

Evaluation of jet breakup length with a CFD code under steam generation condition in a pre-flooded cavity

When the reactor vessel is penetrated in a severe accident of light water reactor, the molten fuel-coolant interaction including the jet breakup occurs and the jet breakup length becomes one of the important parameters. Most numerical studies on jet breakup process have been carried out using dedicated computer codes. Some researchers are trying to apply commercial CFD codes to their investigations on comprehensive jet breakup process. However, the complexity of the phenomena limits the CFD application only to hydrodynamic aspects. In the present study, numerical analysis of jet breakup under vapor generation is pursued using the STAR-CCM + code. The obtained CFD prediction of the MATE09 experiment shows jet breakup progression patterns consistent to the images taken in the experiment. Further, the predicted positions of leading head, which determine the jet breakup length, are in good agreement with the MATE 09 data. The investigation of hydrodynamic effects on the jet breakup with higher jet velocity results in a stronger shear force and earlier jet breakup process even though there exists the vapor pocket around the corium jet. In future studies, the effect of vapor intensity on the jet breakup length would be investigated further by changing other parameters.

SYNTHESIS AND CHARACTERIZATION OF CESIUM SILICATE TO DETERMINE ITS DETAILED PROPERTIES AS CHEMISORBED ONTO STRUCTURAL MATERIALS OF LIGHT WATER REACTOR DURING SEVERE ACCIDENT CONDITIONS

Cesium chemisorption phenomenon strongly contributes to the source terms transport during light water nuclear reactor accidents. Large amounts of cesium silicates are identified to be chemisorbed onto structure material, reduce cesium volatility, and affect the late release and re-vaporization phenomena. Although it has been studied for a long time, several characteristics of these compounds are still under discussion. In this study, Cs2SiO3, Cs2Si2O5, and Cs2Si4O9 were synthesized through the solid-state method and the results have been confirmed using X-Ray Diffraction (XRD) measurement. Furthermore, their crystal structures have been refined based on the XRD analysis. The crystal structure refinement of these compounds proves the previous studies, but with minor distinctions in the lattice parameters. XRD patterns changing over time when measured in the open-air environment also show that Cs2Si4O9 is the most stable species among other cesium silicate species. This indicates that the chemisorbed Cs-Si-O compound onto the structural material as identified by previous studies is most likely Cs2Si4O9 rather than Cs2SiO3 or Cs2Si2O5. Therefore, detailed Cs2Si4O9 identification including its thermodynamic properties characterization could be very useful to enhance the database that is being built to improve current source terms transport codes.

Comparison of unirradiated and irradiated AGR-2 TRISO fuel particle oxidation response

The silicon carbide (SiC) coating in a tristructural isotropic (TRISO) particle acts as a barrier to fission product release during reactor operation and accident scenarios. Oxidation and subsequent failure of the SiC layer during a rare air ingress event is a proposed mechanism for fission product release in a high-temperature gas-cooled reactor (HTGR). Although previous oxidation studies have analyzed unirradiated TRISO particle response, this study compared the oxidation behavior of irradiated and unirradiated TRISO particles from the second Advanced Gas Reactor Fuel Development and Qualification Program irradiation experiment (AGR-2). Particles with exposed SiC were subjected to six varying oxidizing tests in the Furnace for Irradiated TRISO Testing (FITT), examined for failure fraction with the Irradiated Microsphere Gamma Analyzer (IMGA) and characterized with focused ion beam and scanning/transmission electron microscopy techniques to analyze the oxide layer. Uncorrelated unirradiated particle failures throughout the series of exposures suggests that external factors inherent to the experiment increased particle failure sensitivity. However, irradiated particle observations indicated an increased failure response at 400 h 1400 °C in both 2% and 21% O2 atmospheres above failure associated with external factors. Oxide thickness measurements after 400 h at 1400 °C revealed a greater oxidation rate than predicted by parabolic growth, which was attributed to the increased complexity of the oxide structure at longer exposure times. Altering the atmosphere from 21% to 2% O2 reduced the average oxide thickness by approximately 12%–14% in both irradiated and unirradiated particles at 400 h 1400 °C. Overall, the minor variations observed between irradiated and unirradiated particles in this study led to the conclusion that unirradiated TRISO particles can be used to approximate irradiated TRISO oxidation kinetics.

Heat Transfer Analysis of Core Melt Contact with a CANDU Calandria

In a severe accident in a CANDU reactor, disassembly of the core could produce a bed of coarse debris at the bottom of the calandria that would eventually transition into a pool of molten corium. During this process, it may be possible for small amounts of molten core material to contact the calandria wall. The transient heat flow through a calandria wall suddenly contacted by molten Zr or corium is analyzed with a finite element model. Ablation of the wall at its inner surface and the temporary increase in heat flux through its outer surface are calculated for various boundary conditions. Model calculations are compared to observations of the ablation and temperature of a stainless steel plate sprayed by prototypic corium in the Cesium Aerosol Generation-4 or CAGE-4 experiment.