Severe Reactor Accident Research Articles

Skin Absorbed Dose Coefficients for Human Legs from Beta Radiation as a Function of Height

External exposure to skin from beta-emitter radionuclides following severe reactor accidents or nuclear testing can result in beta burning and other health complications. The skin absorbed dose coefficient (SADC) measures the energy deposition into the skin during such accidents. The U.S. Environmental Protection Agency has published several reports to measure the possible energy deposition into the skin in such accidents. However, the most recent SADC published by Federal Guidance Report (FGR) 12 was computed only at one meter above the contaminated surface. Therefore, it was necessary to develop a model to estimate the absorbed dose coefficients for skin at different heights. In this manuscript, Geant4, a Monte Carlo simulator toolkit, was used to estimate the absorbed dose coefficients from electron sources located on the soil surface with energies ranging from 0.1 to 4 MeV. The energy deposited from primary electrons, secondary electrons, and photons in a 50 µm thick layer of epidermis tissue (Basal Cells Layer) located at a depth of 50 µm from the skin surface was estimated at several discrete heights of human leg phantom. More than 40% of the total energy deposited comes from secondary electrons and photons in energy sources of 0.1 and 0.2 MeV on average, but for higher energies, this percentage is less than 1%, which indicates primary electrons are the main source of the deposited energy in the skin. Furthermore, the results showed the energy deposited into skin closer to the ground was 50–100% higher than the previously estimated doses for 1 m above the ground. The results from Geant4 showed a great correlation (R2 = 0.972) with the FGR 12 data at one meter height, and they were aligned with the published values from FGR 12, which validated the simulation results. Therefore, the calculated dose coefficients for different energy sources and different heights could be used in radiation protection measurements.

Chemical interaction of CsOH vapor with UO2 and Fe-Zr melt

ABSTRACT At TEPCO’s Fukushima Daiichi Nuclear Power Station, it is estimated that considerable amounts of cesium still remain in the reactors from the analysis results using the severe accident analysis codes and the reverse analysis from contaminated water. Since cesium is known to form stable compounds with uranium and zirconium, chemisorption experiments with uranium dioxide pellets and iron-zirconium melts for cesium hydroxide vapor were carried out using the similar method as for stainless steel. As the results, formations of cesium uranate, Cs2UO4, and cesium zirconate, Cs2ZrO3, were confirmed, indicating that cesium was chemisorbed on both of the uranium dioxide pellets and the iron-zirconium melts in an Ar-H2-H2O flow and an Ar-H2 flow, respectively. Therefore, it was considered that cesium released from fuel might be trapped by chemisorption with fuels and/or iron-zirconium melts during light water reactor severe accidents.

CESIUM CHEMISORPTION ONTO STAINLESS STEEL UNDER SIMULATED LIGHT WATER REACTOR SEVERE ACCIDENT

During light water reactor severe accident, source terms may interact with structural materials, generating new compounds and affecting their volatility which make the existing codes could not accurately estimate the radioactive release. Cesium is one of the source terms that can interact with structural materials and contributes greatly to the late release phenomenon. Several studies have been conducted to predict the physicochemical interactions between cesium and structural materials. However, the types of chemisorbed cesium compounds onto structural materials are still under discussion. For this reason, this study was carried out using advanced techniques, involving TEM, SEM, EDS and FIB to estimate the chemisorbed cesium compounds onto stainless steel structural material under simulated light water severe accident. This study indicates that cesium is strongly adsorbed on the oxide layer of stainless steel in the form of cesium silica, cesium aluminum silica, and/or cesium ferro silica. CsFeSiO4 and CsAlSiO4 could dominate these compounds.

Investigations on aerosol transport and deposition behavior during severe reactor accident

Investigations on aerosol transport and deposition behavior during severe reactor accident

Experimental study on the film droplet production from a bubble burst on the free surface

The aerosol released during a severe reactor accident may be retained in the liquid pool. During the venting stage of the suppression pool, bubbles generated in the pool will rise to the free surface and burst to generate droplets. The airflow re-entrains aerosols from the liquid to the air via droplets. Accurate evaluation of droplet entrainment needs to acquire the diameter and number of droplets generated by a bubble, so studying the various behaviors that affect the above parameters is necessary. In this paper, high-speed photography is used to visualize the phenomenon of a single bubble burst to generate film droplets in deionized water and TiO2 suspensions at different phase temperatures. Image processing is used to obtain the bubble film opening rate, the bubble burst position, the bubble lifetime, and the diameter and number of film droplets. The results show that the film thickness decays with time in a power law with a parameter of 2/3, and the newly defined film thickness also decays with time in a power law after considering the bubble radius and physical parameters. This rule is applicable in different liquid temperatures and solutes. Besides, the film thickness increases with liquid temperature and aerosol concentration; the bubble lifetime obeys the Weibull distribution, with the average bubble lifetime of deionized water varies non-monotonically with the liquid temperature, the average lifetime increase with aerosol concentration and liquid temperature in aerosol suspension; The average droplet diameter decreases with bubble lifetime in a power law with a parameter of 5/16, the droplet diameter distribution can be calculated by the gamma distribution of the normalized droplet diameter and the bubble lifetime, and in good agreement with the experimental data, the droplet entrainment factor can also be calculated from the critical droplet diameter and bubble lifetime, which decreases with bubble lifetime.

Parametric study of molten core concrete interaction and assessment of corium coolability in the light of MCCI experiments

Molten Corium Concrete Interaction (MCCI) is an important phenomenon observed in the late phase of reactor severe accidents. A realistic analysis of MCCI can save cost and efforts to mitigate severe accidents. The heat transfer mechanisms in MCCI, such as melt eruption and water ingression, are the primary focus of current research owing to complications in their quantification and replication in severe accident analysis codes. This study investigated MCCI using top flooding strategy with MELCOR 2.2 code for two concrete types, and three experiments that were conducted under (OECD/MCCI-1 program). The results demonstrated that the ablation process terminated in the case of Limestone Common Sand (LCS) concrete. In the simulation of siliceous (SIL) concrete, the maximum and average melt front locations confirmed the asymmetrical ablation progression. The LCS concrete showed a smaller axial ablation front in this study; however, the comparison with the concrete ablation volume (0.21 m3) showed an exact match, and the difference between the experimental and predicted ablated mass of concrete was only 0.61%. The quantitative comparison between the prediction of the present study, CORQUENCH code, and experimental results is provided. The parametric values of MCCI phenomenon are also predicated and model improvement for MELCOR 2.2 code are discussed as well.

Study on contact melting process of vertical cylindrical materials in a circular container

In view of the reactor core melting in a severe reactor accident, the contact melting process of vertical cylindrical materials inside a circular container is investigated. The melt analysis model and governing equations are established by the contact melting theory. The dimensionless analytical expressions of melting velocity and melting height were obtained by theoretical solution. The influence of temperature difference, cylinder solid size and filling rate was discussed and analyzed. It is found that the melting velocity decreases gradually during the melting process. Before all the molten liquid submerges the solid, the buoyancy of solid increases gradually and the melting velocity decreases rapidly. After all the molten liquid submerges the solid, the buoyancy and gravity decrease at the same time, and the melting velocity changes gently. Compared with the melting of an individual solid with the same height and volume, the melting velocity of multiple cylindrical solid can be effectively accelerated.

Cold Crucible Facility for Severe Accident Research at Canadian Nuclear Laboratories

Abstract The properties and behavior of corium are important factors in the progression and ultimate consequences of a severe nuclear reactor accident. Canadian Nuclear Laboratories (CNL) is developing a cold crucible facility to create and study small quantities of CANDU corium and, potentially, other molten materials. This paper describes the facility and presents preliminary results with nonradioactive charges and with radioactive, near-prototypical charges.

Effect of molybdenum release on UO2/MOX fuel oxidation under severe light water reactor accident conditions

To investigate the Mo release behavior and its influence on the oxidation of UO2/mixed oxide (MOX) fuel, the oxidation and evaporation (O&E) behaviors of Mo powders and their influencing mechanisms with respect to oxygen partial pressure around powders (pO2s) were researched by using a thermogravimetric analysis technique under Ar-0.1 % O2, Ar-1.0 % O2, and Ar-1.5 % O2 atmospheres in the temperature range from 1473 to 1673 K. The study results reveal that Mo O&E could quickly consume oxygen, substantially decreasing pO2s. This decrease in pO2s facilitates the attainment of an oxygen mass balance by enhancing the diffusion rate of O2(g) towards the powder surfaces and decelerating the O&E rates of the powders. Subsequently, the oxygen consumption rate with respect to Mo release (FO-con) and the potential oxidation rate of UO2 nuclear fuels (FO-ads) under the typical accident conditions created in VERCORS were estimated using the latest analytical model of Mo and the existing fuel oxidation model, respectively. The results demonstrate that, as accident proceeds, FO-con has the potential to surpass FO-ads, thereby retarding the oxidation of nuclear fuel.

Interaction of Cesium Hydroxide with Oxide Layers Under Simulated Light Water Reactor Severe Accident

A large amount of cesium hydroxide (CsOH) is generated during a light water reactor severe accident (SA) and transported through leaky parts to the environment. During that process, some CsOH may interact with oxidized structural materials and change their physicochemical properties. Accurate examination of this interaction is required by source term analysis to derive consistent and appropriate source term transport models, i.e., for SA, decommissioning, and dismantling work of a nuclear reactor. To obtain detailed interaction characteristics, in this study CsOH was exposed to Fe3O4/Fe2O3 and Cr2O3 under a simulated SA environment over a wide temperature range, from 300°C up to 1050°C. As a result, Cs2FeO4, CsFeO2, and Cs2CrO4 were observed at respective temperatures. Cs2FeO4 is stable only at low temperatures and decomposes to form CsFeO2 at about 591°C. However, both Cs2FeO4 and CsFeO2 could react with Cr2O3 to form more stable Cs2CrO4, which melts at 957°C and then completely evaporates at higher temperatures.

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.

Dry-out characteristics of debris bed under different water injection conditions

A porous debris bed may form through the interaction between melting corium and coolant during a pressurized water reactor severe accident, which continues to emit large amounts of decay heat. The cooling of debris bed is a critical phenomenon and serves as a potential mitigation measure to halt the accident. In this paper, an experimental facility DBC was constructed to study the cooling properties and dry-out heat flux under different water injection conditions and system pressures. The debris bed consisted of stone particles ranging in diameter from 1～10 mm, with a porosity of 0.37. In the debris bed, a total of thirty-three electrical heating elements, twelve optical fibers equipped with 204 temperature sensors, and thirty-three thermocouples were embedded. Additionally, a computer program MIDEC was developed to investigate the vapor-liquid flow characteristics and dry-out mechanism. The experimental results indicated that the cooling of debris bed can be significantly improved by increasing system pressure. Furthermore, it was observed that the dry-out heat flux under bottom water injection conditions was considerably higher compared to that of top water reflooding. Under top reflooding conditions, the counter-current flow of vapor-liquid impeded the downward flow of coolant, resulting in insufficient coolant supply and leading to a dry-out phenomenon at the bottom region. However, under bottom injection conditions, both vapor and liquid flowed in an upward direction. As a result, the void fraction reached its maximum value at the top region of the debris bed. The distinct vapor-liquid flow characteristics under top and bottom injection conditions resulted in notable discrepancies in dry-out heat flux and dry-out positions. Based on the comparison between the calculated dry-out heat flux (DHF) and experimental results, it was observed that the Reed model accurately predicted the cooling of debris bed under top-flooding conditions when compared with measured data obtained from DBC experiments. However, for very small particle sizes, the Lipinski model would be more suitable and provided better predictions.

Study on aerosol thermophoretic force and thermophoresis deposition model in reactor building

During nuclear reactor severe accident, the input of external passive containment cooling system will result in significant aerosol thermophoresis deposition in reactor building, which plays important factor affecting the safety assessment of nuclear reactor. Ratio of aerosol particles radius and gas molecule mean free path (Kn number) in reactor building is not greater than 2, corresponding to the near-continuous medium zone and transition zone. Firstly the typical thermophoretic force equations for these zones are evaluated. Combined with key influencing factors analysis, the correction factor connected to Kn number is introduced to address the deviation in the prediction for the transition zone of the current equations. The modified thermophoretic force equation obviously improves the underestimation of the original equation and is then used to obtain the thermophoresis deposition coefficient. According to experimental validation, the obtained coefficient is more accurate than the model that is widely employed in analysis codes in predicting the thermophoresis deposition of aerosol particles. Finally, fluid continuity, momentum, energy and particle continuity equations are established with deduced thermophoresis deposition coefficient and solved numerically. The thermophoresis deposition model considering natural convection (CT model) is established by theoretical derivation and fitting with the numerical solutions. By introducing the effects of natural convection on temperature distribution and aerosol concentration distribution the CT model could accurately describe the aerosol thermophoresis deposition behavior under natural convection conditions for Kn number is not greater than 2, which can be applied to aerosol thermophoresis deposition prediction in reactor building.

Transient heat transfer and crust evolution during debris bed melting process in the hypothetical severe accident of HPR1000

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m2. The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20–50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

Experimental and numerical study on the oxidation and melting behaviors of annular fuel rod under high temperature steam condition

In the safety analysis of nuclear reactor severe accident, the oxidation and melting behaviors of fuel rods under high temperature steam condition is a crucial issue that need to be concerned. As a new type of high-performance light water reactor nuclear fuel, the annular fuel is composed of outer cladding, annular fuel pellet, inner cladding and two coolant channels. At present, the oxidation and melting behaviors research of annular fuel under high temperature steam condition is insufficient. Thus, this paper builds a visual experimental platform for out-of-pile analysis, and carries out the high temperature oxidation experimental research of annular fuel rod based on real cladding material-Zircaloy. The transient temperature characteristic curve and visualization process of annular fuel rod under the conditions of high temperature melting, steam oxidation and re-flooding are obtained. In addition, this paper establishes a seven steps numerical analysis method for oxidation and melting behaviors analysis of annular fuel rod. Based on the three tests results, the accuracy of the model is verified, and the temperature prediction error is within 10%, which lays a foundation for the further calculation of the whole reactor core.

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.

Study on the safety characteristics of molten corium in-vessel retention of annular fuel core

In the process of reactor severe accidents, maintaining the integrity of reactor pressure vessel to achieve IVR (In-Vessel Retention) of molten corium plays an important role in accidents mitigation and suppression. The ERVC (External Reactor Vessel Cooling) is an important technology to realize IVR. As a new economical annular fuel, the increase of zirconium content and power density in the core may threaten the integrity of lower head under severe accidents. In this paper, the IVRASA (In-Vessel Retention Analysis code in Severe Accident) is used to analyze the heat transfer characteristics of lower head molten pool corresponding to annular fuel core. The components of two-layer and three-layer configurations of molten pool are determined based on the geometric structures of annular fuel and verified heavy metal layer model. In terms of the integrity of lower head, the power density of the annular fuel core can be increased to about 110%.

Modelling of the expansion phase of Sodium fast reactor severe accident

Improving safety is one of the current objective of the fourth generation of sodium fast reactors (SFR). In order to achieve this requirement and increase safety margins, developments are carried-out to take into account severe accidents as soon as the pre-conceptual phase. In this context, fast-running and parametric tools are used in addition to mechanistic tools for parametric studies and probabilistic margin evaluations.This paper is dedicated to MOREINa (Model Of vapouR Expansion and fuel coolant Interaction in sodium fast Reactor), a new parametric tool modelling expansion phase transients. During a severe accident in SFR integrating a low void worth core (non energetic primary phase), a meltdown of the core may occur and form a molten material pool in the vessel. After a secondary power excursion in the molten material pool, the expansion phase implies the self-vaporisation of superheated materials and their expansion towards the sodium located in the upper plenum, above the core. Fuel Coolant Interaction (FCI) will be possible during this phase when the unvaporised overheated molten materials are mixed with the coolant, implying a huge transfer of energy from the superheated materials to the coolant ; vaporising this latter. The mechanical energy released by the two aforementioned phenomena is evaluated during this phase in order to assess the potential damage on the vessel. Thus, MOREINa integrates a new expansion phase model quantifying the mechanical energy induced by this severe accident phase in order to study its consequences on the reactor vessel walls.The modelling implemented in MOREINa, based on dimensional analysis and different balance equations, is able to simulate:- fuel and steel vaporisations;- the creation of a vapour bubble with or without fission gases inside;- the expansion of the molten materials due to the pressure difference between the saturated pressure of the molten materials and the cover gas located above the sodium, at the top of the vessel;- the integration of the molten material droplet into the bubble vapour owing to Rayleigh–Taylor instabilities at the vapour–liquid interface;- the fuel coolant interaction (FCI) if the expansion leads the molten materials to contact the coolant. First, the safety context of the expansion phase study is introduced. The new parametric tool MOREINa is then presented with its main models. Validations of MOREINa against a validated tool and on available test cases are carried-out and, finally, a parametric study is presented and analysed.

Radiolytic degradation of dimethyl telluride in aqueous solutions

The formation of volatile radioactive species is a major concern in severe nuclear reactor accident scenarios. Release of radioactive material to the environment is highly governed by the volatility of the species and therefore it its crucial to understand the behavior of any such species during the accident and the days and weeks following. One of the volatile, yet highly understudied fission products is tellurium. Although tellurium has been released in significant amounts during the major nuclear accidents that have occurred, the knowledge of the behavior is still lacking. Here we present results on the radiolysis of dimethyl telluride, a highly volatile species shown to form in accident conditions. The behavior of dimethyl telluride was investigated under gamma irradiation in various aqueous solutions and conditions representative to severe nuclear reactor accident conditions. The results suggest that dimethyl telluride is relatively stable towards gamma irradiation and its degradation is highly affected by the amount of dissolved oxygen and competing species. It was found that dimethyl telluride degrades via oxidative processes by reacting with oxidizing radiolysis products e.g. •OH, O.-. In the absence of oxygen, several volatile telluride dimers were observed. The results presented here increase the interest in organic tellurides in severe accident conditions and highlight the need for further investigation of the re-volatilization and mitigation of volatile tellurium species.

Evolution of the perovskite phase in UO2-based samples under conditions representative of a severe nuclear accident by XANES

To understand the fission products behaviour during a Nuclear Severe Accident (NSA), simulated high burnup UO2 fuels produced by sintering at high temperature (SIMFUEL) were submitted to thermal treatments in conditions representative of the beginning of a Pressurized Water Reactor (PWR) severe accident. The samples made of UO2 doped with eleven fission products (FP) surrogates were thus thermally treated from 400 ​°C up to 1000 ​°C at different oxygen potentials (up to −290 ​kJ ​mol(O2)−1 at 1000 ​°C). The samples were characterized before and after the heat treatments by Secondary Electron Microscopy coupled with Energy Dispersive X-ray Spectroscopy (SEM-EDX) and X-ray Absorption Spectroscopy (XAS) at the MARS beamline, SOLEIL Synchrotron and at the FAME-UHD beamline, ESRF. Initially the SIMFUEL is constituted of a UO2 matrix, a metallic phase composed of Mo, Pd, Ru, Rh and Tc, and an oxide phase with perovskite structure depicted as (Ba, Sr)(Zr, U, RE)O3 where RE stands for Rare Earths. Around 1000 ​°C, it was found that Mo, initially contained in the metallic inclusions, oxidizes into MoO2. In the same conditions, a part of the oxidized Mo interacts with the perovskite phase containing Ba to form BaMoO4. These results will be useful to validate the theoretical models used in calculation codes and to assess thermodynamic databases, which are one of the main sources of errors when calculating a NSA sequence involving molten fuel.