Accidents In Light Water Reactors Research Articles

Studies on innovative hydrogen recombiners as safety devices in the containments of light water reactors

In order to prevent the containment and other safety relevant components from incurring serious damage caused by a detonation of the hydrogen/air-mixture generated during a severe accident in light water reactors (LWR) passive autocatalytic recombiners (PAR) are used for hydrogen removal in an increasing number of European plants. These devices make use of the fact that hydrogen and oxygen react exothermally on catalytic surfaces generating steam and heat. Experimental investigations at several research facilities indicate that existing PAR systems bear the risk of igniting the gaseous mixture due to an overheating of the catalyst elements caused by strong reaction heat generation. Innovative devices could overcome existing limitations making use of the knowledge deduced from experiments performed at the REKO facilities at Forschungszentrum Juelich (FZJ). The paper analyses the mechanisms of the thermal behaviour of catalytic plate-type recombiners and presents experimental results on existing and innovative devices for hydrogen removal introducing the modular recombiner concept.

Innovative devices for hydrogen removal

In case of a severe accident in light water reactors (LWR) a high amount of hydrogen, up to about 20 000 m n 3, might be generated and released into the containment. The mixture, consisting of hydrogen and oxygen, may either burn or detonate, if ignited. In case of detonation the generated shock wave may endanger the integrity of the containment or safety-related systems. Consequently, an effective removal of hydrogen is required. Hydrogen and oxygen react exothermally at catalytically acting surfaces, already at room temperature, and this is used in catalytic recombiners. It is recommended to combine recombiners with spark or catalytic igniters, in order to cover a broader spectrum of accident sequences. In this contribution, state of the art of hydrogen removal devices are reviewed and the possibilities for innovative methods, making use of the phenomena arising in the containments, using further components will be illustrated accordingly.

Effect of Microstructure on Failure Behavior of Light Water Reactor Coolant Piping under Severe Accident Conditions

In a severe accident of light water reactors, the reactor coolant system (RCS) piping might be subjected to thermal loads caused by the decay heat of the deposited fission products and the heat transfer from the hot gases, with an internal pressure in some accident sequences. Tests on the RCS piping failure were performed along with high temperature tensile and creep rupture tests including metallography to investigate the failure behavior. The prediction of the 0.2% proof stress by Arrhenius equation is in good agreement with the measured stress above 800°C for served RCS piping materials. The modified Norton's Law for the short term creep rupture model agrees with the experimental values between 800 and 1,150°C for type 316 stainless steel. The microstructural change was discussed with the effect of the very rapid formation and resolution of the precipitation on the strength at high temperature. The result of the piping failure tests which simulated the severe accident conditions, i.e., in short-term at high-temperature, could support the plastic limit load prediction of the flow stress model using the 0.2% proof stress.

A Probabilistic Approach for the Evaluation of Reactivity Insertion Accident Effects

Reported fuel failures at low peak enthalpies for highly exposed fuel during fast reactivity transients promoted the evaluation of reactivity insertion accidents (RIAs) in light water reactors with an approach different from the ultraconservative point-kinetics licensing evaluations performed in the past. On the basis of realistic estimates for the rod worth and plant conditions, an evaluation of the consequences of the RIAs has been performed. For the pressurized water reactor and boiling water reactor (BWR) cases, rod worth became the most important parameter affecting the severity of the accident. In BWRs high subcooling can adversely affect the consequences of the event. The RIA analyses have been performed using an estimation of the distribution for these parameters in an actual plant during startup. The results show that when assumptions consistent with operating procedures are used, the probability of a significant enthalpy increase due to an RIA is greatly reduced.

Introduction by the Guest Editors

In order to sustain the structural integrity of the containment and other safety relevant components i.e. to avoid a detonation of the hydrogen-air mixture generated during a severe accident in light water reactors, passive auto-catalytic recombiners (PAR) are used for hydrogen removal in many European nuclear power plants (NPP). In 1999, the German NPP Emsland (KKE) was equipped with 58 PAR of AREVA design as an internal accident management measure for a beyond-design accident. Since that time the recombiners are in a stand-by state. As the catalyst elements are exposed to various airborne substances during normal plant operation their function is controlled periodically by testing selected catalyst sheets in a specially designed device. Under the conservative test conditions during this procedure some catalyst sheets showed a delayed responding behavior. First internal analysis gave indication of a beginning fouling on the catalytic surface.The aim of a precautionary investigation performed in cooperation between KKE, Forschungszentrum Juelich and RWTH Aachen University was to characterize the composition of the fouling and to correlate it with potential sources within the containment.In the framework of the investigation the reports of the periodic inspections were analyzed and appropriate sample sheets were selected from the installation. These samples were subjected to a comprehensive chemical surface analysis in order to identify effects like thermal sintering, poisoning or a blocking of the catalytic surface (Baerns, M., 2004. Basic Principles in Applied Catalysis, Springer Verlag). Along with the chemical analysis the catalytic activity of the samples was assessed in several test series in order to correlate the chemically quantified deposition on the catalyst samples with the characteristics of the start-up and the steady-state performance of the recombination reaction. In a final step, possible sources of the fouling were analyzed with regard to their possible contribution to the phenomena. According to the results achieved, measures have been implemented at KKE in order to optimize procedures and to enhance the performance of the PARs.

Non-Empirical Analysis of the Chemical Reaction of Cesium with Steam; CS+H2O→CsOH+H, in Severe Light Water Reactor Accidents

The elementary chemical reaction of cesium which is a volatile fission product with steam (water) shown in the title are analyzed from the first principle. The activation energies for both reactions of the forward and the backward are calculated for the first time with a computer program Gamess, which is based on the non-empirical molecular orbital theory. The calculated values are expected to agree with those of experiments within an accuracy of several tens of kJ/mol. Using the values obtained above, the rate constants are also calculated in the framework of the transition state theory. The derived rate constants will allow us to predict the orders of magnitude of experimental values under high temperature conditions (1,000K or higher). Thus these rate constants are available as input data for the kinetic (non-equilibrium state) study of cesium in severe light water reactor accidents.

Non-Empirical Analysis of the Chemical Reaction of Cesium with Steam; Cs+H2O.RAR.CsOH+H, in Severe Light Water Reactor Accidents.

The elementary chemical reaction of cesium which is a volatile fission product with steam (water) shown in the title are analyzed from the first principle. The activation energies for both reactions of the forward and the backward are calculated for the first time with a computer program Gamess, which is based on the non-empirical molecular orbital theory. The calculated values are expected to agree with those of experiments within an accuracy of several tens of kJ/mol. Using the values obtained above, the rate constants are also calculated in the framework of the transition state theory. The derived rate constants will allow us to predict the orders of magnitude of experimental values under high temperature conditions (1,000K or higher). Thus these rate constants are available as input data for the kinetic (non-equilibrium state) study of cesium in severe light water reactor accidents.

Thermodynamics of the UO and Zr-systems and application to analysis of fuel liquefaction during severe accidents in light water reactors

Severe accident modeling requires knowledge of the heat effects accompanying simultaneous dissolution of UO 2 and fusion of Zircaloy to form the UZrO melt. Estimation of the enthalpy changes must rely chiefly on data from the UO and ZrO binary systems. The partial molar enthalpies of oxygen in the pure liquid metals are determined by application of Sieverts' law. The Sieverts' law constant is obtained by deriving the complete relationship between the oxygen pressure, the O M of the solid or liquid phase and temperature from the pure metal M to the dioxide MO 2. This assessment utilizes pressure p, over the entire composition range. Once the p-C-T relatnonship has been established, the enthalpy changes on dissolving UO 2 and αZr(O) into the UZrO melt are computed. The heat effects of fuel/cladding dissolution in a severe accident are applied in a liquefaction model based on kinetic control by the available heat from fission product decay. In addition to the portion of the decay heat reaching the inner cladding surface, the rate of melting of the cladding depends on heat flow to its outer surface from steam oxidation or radiation from other parts of the core. Depending on the relative importance of these two heat inputs to the cladding, time to melting is from 100 to 400 s and the fraction of the fuel dissolved varies from 5 to 20%.

Non-Empirical Analysis of the Chemical Reactions of Iodine with Steam; I+H2O→HI+OH and I+H2O→IOH+H, in Severe Light Water Reactor Accidents

The elementary chemical reactions of iodine which is a volatile fission product with steam (water) shown in the title are analyzed from the first principle. The activation energies of these reactions are calculated for the first time with a computer program Gaussian94, which is based on the non-empirical molecular orbital theory. The calculated values are expected to agree with those of experiments within an accuracy of a few kcal/mol. Using the values obtained above, the rate constants are also calculated in the framework of the transition state theory. The derived rate constants will allow us to predict the orders of magnitude of experimental values under high temperature conditions (1,000K or higher). Thus these rate constants are available as input data for the kinetic (non-equilibrium state) study of iodine in severe light water reactor accidents.

Non-Empirical Analysis of the Chemical Reactions of Iodine with Steam; I+H2O.RAR.HI+OH and I+H2O.RAR.IOH+H, in Severe Light Water Reactor Accidents.

The elementary chemical reactions of iodine which is a volatile fission product with steam (water) shown in the title are analyzed from the first principle. The activation energies of these reactions are calculated for the first time with a computer program Gaussian94, which is based on the non-empirical molecular orbital theory. The calculated values are expected to agree with those of experiments within an accuracy of a few kcal/mol. Using the values obtained above, the rate constants are also calculated in the framework of the transition state theory. The derived rate constants will allow us to predict the orders of magnitude of experimental values under high temperature conditions (1,000K or higher). Thus these rate constants are available as input data for the kinetic (non-equilibrium state) study of iodine in severe light water reactor accidents.

Non-Empirical Study of Chemical Reactions Including Fission Products in Severe Light Water Reactor Accidents

Non-Empirical Study of Chemical Reactions Including Fission Products in Severe Light Water Reactor Accidents

Fission Product Releases at Severe Light Water Reactor Accident Conditions: ORNL/CEA Measurements Versus Calculations

During the 1970s, reactor safety authorities developed increasing interest in methods for accurately predicting the extent of hazards associated with severe accidents in light water reactors (LWRs). In response to these concerns, out-of-pile experimental projects were initiated by the U.S. Nuclear Regulatory Commission and the French Nuclear Protection and Safety Institute, at Oak Ridge National Laboratory (ORNL) and the Commissariat a l'Energie Atomique (CEA), respectively. Both experimental efforts were designed for source term characterization of the fission products (FPs) released from LWR fuel samples under test conditions representative of severe accidents, i.e., in oxidizing or reducing atmospheres at temperatures up to 2700K (at ORNL) and 2570K (at CEA). The experimental devices, procedures, and parameters are described. The combined database of available results is summarized and related to experimental conditions. Using Booth diffusion theory, diffusion coefficients of the FPs were calculated, and their evolution with temperatures in the 1070 to 2700 K range were plotted. The results show the good agreement between the independently determined ORNL and CEA FP diffusion coefficient values. By plotting the data in Arrhenius fashion, it has been possible to do the following : 1. quantify the thermal activation coefficients for both volatile and low volatile FPs 2. identify classes of FPs whose release behavior does not follow a purely diffusional mechanism, but rather depends on chemical interactions with the environment ; i.e., they exhibit mixed diffusional and transport-dependent mechanisms. A model supported by these experimental results and based on the CORSOR-Booth code is proposed and compared with CORSOR calculations. To facilitate the comparison, a model based on FP diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model, as well as CORSOR-Booth calculations are compared with the ORNL experimental data in a blind test. Additional experimental work is in progress to determine the releases of nonvolatile FPs and transuranics at temperatures up to fuel melting as well as FP deposition mechanisms under both oxidizing and reducing conditions. An effort is made to quantify the dependence of the temperature transient δT/δt as well as the oxygen potential effects on the release kinetics. Those data will allow further verification and extension of the field of application for our model.

Non-Empirical Study of Chemical Reactions Including Fission Products in Severe Light Water Reactor Accidents.

Elementary chemical reactions including volatile (gaseous) fission products, such as iodine and cesium, are analyzed from the first principle, so as to enable us to study the kinetic behavior (non-equilibrium state) of fission products in severe light water reactor accidents without experimental rate constants. Activation energies are calculated with a computer program Gaussian94, which is based on the non-empirical (ab initio) molecular orbital theory. Calculated values are expected to agree with those of experiments within an accuracy of a few kcal/mol. Using the values obtained above, rate constants are also calculated in the framework of the transition state theory. It is concluded that the derived rate constants can predict the orders of magnitude of experimental values under high temperature conditions (1,000 K or higher) as in severe accidents.

KESS—a modular program system to simulate and analyze core melt accidents in light water reactors

The research and development work program performed at the Institute for Nuclear Technology and Energy Systems and related to nuclear reactor safety includes, amongst other things, the development of detailed, mechanistic models for the analysis of severe accidents in light water reactors (LWRs). It is intended to describe the most relevant physical and chemical processes likely to occur in the reactor core, during heatup and degradation, degraded core behavior and fission product release. The models have been developed mainly as independent modules and are implemented into the KESS code system by means of the RSYST informatic structure, RSYST provides module sequences, initial and boundary conditions for the simulation, handles data interfaces between modules and includes pre- and postprocessing tools. Works has been focused on competition of model development and partial verification for early-phase accident phenomena and selected late-phase processes, such as debris bed behavior. The code version kess-III/mod1.1 has been applied to analyze selected integral experiments and to simulate reactor core behavior for specific pressurised water reactor (PWR) accident sequences. Selected kess modules have also been implemented into the system code athlet-cd (GRS, Germany) and the ester code package (European development). kess will be further extended in national and European research programs, including e.g. improvement of existing early-phase models and new model development for late-phase phenomena.

Breakup of melt jets as pre-condition for premixing: Modeling and experimental verification

As a consequence of severe light water reactor accidents with core melting, jets of melt are considered to flow into the remaining water in the lower plenum. The breakup of these jets above and in the water determines the further accident progression, especially the mixing of the molten core material with the water and thus the possibility of strong vapor explosions or the mode of contact of core material with the lower plenum structures. Stripping at the melt surface due to relative flow, supported by interface instabilities, is considered here as a major mechanism of the jet breakup. Based on this, correlations are analysed in comparison with experimental results. Results from the more detailed model Ikejet developed at the IKE yield strong differences between the Kelvin-Helmholtz (KH) description of the interface instabilities and the Jeffreys-Miles (JM) approach. Calculations for experiments with simulant materials under non-boiling as well as film boiling conditions support the latter description whereas much too strong breakup results with KH. Correspondingly, applications to corium jets yield much longer coherent jets with the JM approach. Comparisons with corium experiments at the Argonne National Laboratories (ANL) and with first FARO-LWR tests at the JRC Ispra also still suffer from experimental uncertainties.

Effect of External Cooling on the Thermal Behavior of a Boiling Water Reactor Vessel Lower Head

Flooding of the drywell of a boiling water reactor (BWR) is one of many accident management strategies being proposed to manage severe accidents in light water reactors. The effect of external cooling on the thermal behavior of the BWR vessel lower head containing molten core material has been numerically investigated using a two-dimensional implicit finite difference scheme. Results have been obtained for the vessel shell temperature, the molten pool temperature, and the crust thickness for steady-state conditions. For each equilibrium state, the thermal behavior of the vessel lower head has been investigated by parametrically changing the emissivity of the pool free surface, the vessel wall, the baffle plate, the core shroud, and the upper structure; and the temperature of the upper structure. For a certain set of parameters, nucleate boiling on the outer surface of the vessel wall is found to be effective in lowering the temperature of the inner wall of the vessel below the melting temperature of the steel. For most cases, failure of the BWR baffle plate would occur.

Qualification of the ARROTTA Code for Light Water Reactor Accident Analysis

Qualification efforts have been performed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research (INER) for the three-dimensional spatial kinetics code ARROTTA for light water reactor (LWR) core transient analysis. Together TPC and INER started a 5-yr project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multidimensional kinetics calculations such as rod ejection for pressurized water reactors and rod drop for boiling water reactors (BWR). To qualify ARROTTA for evaluation of the Final Safety Analysis Report licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicate that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include the Nuclear Energy Agency Committee on Reactor Physics rod ejection problem, the three-dimensional Langenbuch-Maurer-Werner LWR rod withdrawal/insertion problem, and the three-dimensional linear regression analysis BWR transient benchmark problem. The results indicate that ARROTTA'smore » accuracy and stability are excellent as compared with other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multidimensional core transients for LWRs.« less

Effect of Outside Cooling on the Thermal Behavior of a Pressurized Water Reactor Vessel Lower Head

Flooding of the reactor vessel cavity is one of many accident management strategies being proposed to manage severe accidents in light water reactors. The effect of external cooling on the thermal behavior of the vessel lower head containing molten core material is numerically investigated using a two-dimensional implicit finite difference scheme. Results are obtained for the vessel shell temperature, pool temperature, and crust thicknesses for both unsteady and steady-state conditions. For both cases, the thermal behavior of the vessel lower head is investigated by parametrically changing the emissivity of the pool surface, the vessel wall, and the upper structure and by changing the temperature of the upper structure. For a certain set of parameters, nucleate boiling on the outer surface of the vessel wall is effective in lowering the temperature of the inner wall of the vessel below the melting temperature of steel. Steady-state results are obtained by using two different heat transfer correlations for the natural convection in the molten pool, which helps in understanding how uncertainties in the modeling of physical processes can influence the evaluation of accident management strategies.

Ex-Vessel Releases of Radionuclides During Molten Core/Concrete Interactions in Severe Light Water Reactor Accidents

In this paper releases of radionuclides and the production of aerosols during the molten core/concrete interaction (MCCI) phase of degraded core accidents in light water reactors are termed ex-vessel releases. The VANESA and METOXA codes were respectively developed by the U.S. Nuclear Regulatory Commission and the Industrial Degraded Core Rulemaking (IDCOR) program to quantify ex-vessel releases. Comparison of calculations by VANESA and METOXA (under identical initial and boundary conditions) show that except for niobium and strontium species, the predicted ex-vessel radionuclide release rates are within an order of magnitude of each other. In an actual application of these two codes to the source term quantification of severe accidents, the initial and boundary conditions for the calculations could be significantly different, as demonstrated in an analysis of an anticipated transient without scram accident sequence in a boiling water reactor. For the same amount of debris, the MCCI thermal-hydraulic results provided for METOXA from a DECOMP calculation tend to drive more radioactive material from the debris pool than those provided for VANESA from a CORCON/MOD2 calculation. The MAAP code, however, predicts that less mass is involved in the MCCI.

Thermodynamic analysis of cesium and iodine behavior in severe light water reactor accidents

In order to understand the release and transport behavior of cesium (Cs) and iodine (I) in severe light water reactor accidents, chemical forms of Cs and I in steam-hydrogen mixtures were analyzed thermodynamically. In the calculations reactions of boron (B) with Cs were taken into consideration. The analysis showed that B plays an important role in determining chemical forms of Cs. The main Cs-containing species are CsBO 2(g) and CsBO 2(l), depending on temperature. The contribution of CsOH(g) is minor. The main I-containing species are HI(g) and CsI(g) over the wide ranges of the parameters considered. Calculations were also carried out under the conditions of the Three Mile Island Unit 2 accident.