Severe Reactor Accident Research Articles

Simulation benchmark based on THAI-experiment on dissolution of a steam stratification by natural convection

Locally enriched hydrogen as in stratification may contribute to early containment failure in the course of severe nuclear reactor accidents. During accident sequences steam might accumulate as well to stratifications which can directly influence the distribution and ignitability of hydrogen mixtures in containments. An international code benchmark including Computational Fluid Dynamics (CFD) and Lumped Parameter (LP) codes was conducted in the frame of the German THAI program. Basis for the benchmark was experiment TH24.3 which investigates the dissolution of a steam layer subject to natural convection in the steam-air atmosphere of the THAI vessel. The test provides validation data for the development of CFD and LP models to simulate the atmosphere in the containment of a nuclear reactor installation. In test TH24.3 saturated steam is injected into the upper third of the vessel forming a stratification layer which is then mixed by a superposed thermal convection. In this paper the simulation benchmark will be evaluated in addition to the general discussion about the experimental transient of test TH24.3. Concerning the steam stratification build-up and dilution of the stratification, the numerical programs showed very different results during the blind evaluation phase, but improved noticeable during open simulation phase.

Using the Star CCM+ software system for modeling the thermal state and natural convection in the melt metal layer during severe accidents in VVER reactors

The possibility of using the Star CCM+ software system for analyzing the thermal state of the melt pool metal layer generated as a result of melt stratification during a severe accident in pressure-vessel nuclear reactors is considered. In order to verify and substantiate the possibility of using this software system for modeling the natural convection processes in the melt at high values of the Rayleigh number, test problems were solved. The obtained results were found to be in good agreement with the known solutions and with the experimental data. The behavior of the melt metal layer was subjected to a parametric analysis for different melt heating conditions, the results of which showed that certain parameters have a determining influence on the so-called focusing effect and on the specific features of current in this layer.

Numerical Studies on Hydrogen Distribution in Enclosures in the Presence of Condensing Steam

Hydrogen release in confined spaces is an important safety issue, which is even more important in the context of severe accident in nuclear reactors. In severe accidents, hydrogen release is associated with release of steam and the condensation of steam on walls leads to increase in concentration of hydrogen in residual gas as well as condensation induced mixing. In the present paper, a model for falling film steam condensation in the presence of noncondensable gases is presented. The steam condensation model is incorporated into an in-house developed cfd code called hydrogen distribution simulator (HDS) and is used for hydrogen distribution studies. The validation of the CFD model with steam condensation against the experimental results from the CONAN facility is discussed. Subsequently, the model is applied to a typical release situation involving release of hydrogen and steam for some time and its distribution for a long time beyond it. The heat and mass transfer aspects of the problem are highlighted. It is seen that steam condensation has an important bearing on the distribution of hydrogen and the mixing behavior of gases in the enclosure.

Development of DETAC and its application to the hydrogen detonation analysis

In this study, a numerical analysis code (DETAC, Detonation Analysis Code) for hydrogen detonation during the reactor severe accident was developed using Fortran 90 language, and the simulation was performed for the hydrogen detonation. A global-chemistry model was adopted to simulate the chemical reaction. The Euler equations were solved using third-order Runge-Kutta method with fifth-order weighted essentially non-oscillatory scheme handling the convection flux. Afterward, the hydrodynamics solver was verified by comparison of predicted results and exact solutions of four cases of shock tube problems. A hydrogen detonation in a pipe was simulated to verify this code by comparing the results with the classical C-J theory. Furthermore, this code was applied to the hydrogen detonation analysis in the compartment of BWR building. Two cases with different ignition locations were analyzed in this paper and the maximum pressure of these cases were 7.5 MPa and 8.0 MPa, respectively. The pressure and the temperature during detonation were affected by the ignition location. The results indicated that the possibility of reactor building destruction exists if the hydrogen detonation occurs.

An integrated approach to source term uncertainty and sensitivity analyses for nuclear reactor severe accidents

Large-scale computer programs simulate severe accident phenomena and often have a moderate-to-large number of models and input variables. Analytical solutions to uncertainty distributions of interested source terms are impractical, and influential inputs on outputs are hard to discover. Runs of such integral codes for complex severe accidents are generally time-consuming and hence computationally expensive. This article presents an integrated approach to uncertainty and sensitivity analyses for nuclear reactor severe accident source terms, with an example which simulates an accident sequence similar to that occurred at Unit 2 of the Fukushima Daiichi Nuclear Power Plant using an integral code, MELCOR. Monte-Carlo-based uncertainty analysis has been elaborated to investigate the released fractions of representative radionuclides, Cs and CsI. In order to estimate the sensitivity of inputs, which have a substantial influence on the core melt progression and the transportation process of radionuclides, a variance decomposition method is applied. Stochastic process, specifically a Dirichlet process, is applied to construct a surrogate model in sensitivity analysis as a substitute of the code. The surrogate model is cross-validated by comparing with corresponding results of MELCOR. The analysis with the simpler model avoids laborious computational cost/load, so that the importance measures for input factors are obtained successfully.

Investigation of severe accident scenario of PWR response to LOCA along with SBO

Severe accident analysis is important for the safety evaluation of a nuclear power plant (NPP). In this paper, analysis has been performed for a Chinese three-loop Pressurized Water Reactor (PWR) severe accident induced by loss of coolant accident (LOCA) along with Station- Block-Out (SBO) using the MIDAC code. The simulation results show the influence of different break sizes in the cold leg on the severe accident progression.Important parameters, such as primary coolant system (RCS) pressure, core melting time, pressure vessel rupture time. are analyzed with three different break sizes in the cold leg. With the break size of 0.002 m2, the reactor core starts to melt at 1481 s and the lower head fails at 10,317 s after the accident occurs. With a larger break size of 0.05 m2, the reactor core starts to melt at 1066 s and the lower head fails at 2473 s. If the break size is further enlarged to 0.2 m2, the reactor core starts to melt at 422 s and the lower head fails at 1757 s. Another case in which a hot leg break is with the size of 0.002 m2 is performed. The reactor core starts to melt at 3641 s and the lower head fails at 14,744 s. It is less severe than the cold leg break accident.The severe accident prevention and mitigation measures are summarized based on the computation results. The results are helpful to develop the management strategies and guideline for the severe accident of the similar types of PWRs.

Transient stratification modelling of a corium pool in a LWR vessel lower head

In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters.

New set of convective heat transfer coefficients established for pools and validated against CLARA experiments for application to corium pools

During an hypothetical Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) severe accident with core meltdown and vessel failure, corium would fall directly on the concrete reactor pit basemat if no water is present. The high temperature of the corium pool maintained by the residual power would lead to the erosion of the concrete walls and basemat of this reactor pit. The thermal decomposition of concrete will lead to the release of a significant amount of gases that will modify the corium pool thermal hydraulics. In particular, it will affect heat transfers between the corium pool and the concrete which determine the reactor pit ablation kinetics.A new set of convective heat transfer coefficients in a pool with different lateral and horizontal superficial gas velocities is modeled and validated against the recent CLARA experimental program. 155 tests of this program, in two size configurations and a high range of investigated viscosity, have been used to validate the model. Then, a method to define different lateral and horizontal superficial gas velocities in a 0D code is proposed together with a discussion about the possible viscosity in the reactor case when the pool is semi-solid. This model is going to be implemented in the 0D ASTEC/MEDICIS code in order to determine the impact of the convective heat transfer in the concrete ablation by corium.

Overview on hydrogen risk research and development activities: Methodology and open issues

During the course of a severe accident in a light water nuclear reactor, large amounts of hydrogen can be generated and released into the containment during reactor core degradation. Additional burnable gases [hydrogen (H2) and carbon monoxide (CO)] may be released into the containment in the corium/concrete interaction. This could subsequently raise a combustion hazard. As the Fukushima accidents revealed, hydrogen combustion can cause high pressure spikes that could challenge the reactor buildings and lead to failure of the surrounding buildings. To prevent the gas explosion hazard, most mitigation strategies adopted by European countries are based on the implementation of passive autocatalytic recombiners (PARs). Studies of representative accident sequences indicate that, despite the installation of PARs, it is difficult to prevent at all times and locations, the formation of a combustible mixture that potentially leads to local flame acceleration. Complementary research and development (R&D) projects were recently launched to understand better the phenomena associated with the combustion hazard and to address the issues highlighted after the Fukushima Daiichi events such as explosion hazard in the venting system and the potential flammable mixture migration into spaces beyond the primary containment. The expected results will be used to improve the modeling tools and methodology for hydrogen risk assessment and severe accident management guidelines. The present paper aims to present the methodology adopted by Institut de Radioprotection et de Sûreté Nucléaire to assess hydrogen risk in nuclear power plants, in particular French nuclear power plants, the open issues, and the ongoing R&D programs related to hydrogen distribution, mitigation, and combustion.

Numerical Modelling of Debris Bed Water Quenching

Debris beds may be formed during a nuclear reactor severe accident and coolability of these beds is important to avoid release of radioactive materials into the environment. However, debris bed water quenching is challenging to understand, and model, because of the complex multi-phase flow and heat transfer physics involved which may include boiling. This paper develops a modelling method for boiling and demonstrates its abilities with some applications. The model is based on a multi-fluid approach, in which one phase represents the liquid, the second phase the gas phase the third phase represents the solid debris bed. In each fluid phase a set of conservation equations for mass, energy and momentum are solved with appropriate inter-phase exchange terms coupling the fluid phases.

ICONE23-1944 ON THE FEASIBILITY OF USING EX-CORE NEUTRON INSTRUMENTATION FOR REACTOR STATE DIAGNOSIS DURING ACCIDENTS

In the case of a severe reactor accident, knowledge of the coolant level and the state of the core, including the progress of a possible core melt, would be of crucial interest. In such a scenario, the in-core instrumentation will most likely not be available. In this work, we explore the possibility of using the ex-core neutron detectors to gain information about the state of the reactor pressure vessel (RPV) inventory for a light water reactor. These detectors, which are typically implemented as ionization chambers, are located inside the biological shield and might still be operational during a severe accident. Stationary Monte Carlo calculations using the radiation transport code MCNP were performed to simulate the transport of neutrons outside the RPV and the reactions of the ionization chambers. These detector signals are computed for different model reactor states which might occur during a severe accident with core meltdown. The reactor model is based on data from a typical German Pressurized Water Reactor. The results indicate that a change in coolant level should be detectable. Due to the core's neutron self-shielding, deformations in the inner core region, such as the formation of a cavity, do not yield different signal rates in the ex-core neutron chambers. Changes not confined to the centre of the core, such as the relocation of corium into the lower head, are detectable by their change in the ionization chambers' reaction rates.

Steel oxidation phenomena during Molten Corium siliceous Concrete Interaction (MCCI)

The VULCANO facility at CEA Cadarache is a Molten Corium Concrete Interaction (MCCI) installation for testing material reactions representative of the late stages of a nuclear reactor severe accident. The objectives of the VBS-U3 test were to study ablation phenomena and oxidation of the metallic phase when two liquid phases are present: oxide phase and metallic phase (steel). In this paper we describe the materials post-test analysis of the VULCANO VBS-U3 test performed at the Institute for Transuranium Elements in Karlsruhe (JRC-ITU) with the focus on the metallic phase oxidation of the corium. Post-test analyses show that the remaining metallic phase of the corium is under two forms: drops discontinuously dispersed in the oxide phase forming an emulsion and a continuous metallic ingot clearly separated from the oxide phase. In average, taking into account or not the metallic phase dispersed in the oxide phase, between 60% and 70% of the steel has been oxidized. The size of the drops and their proportion in the oxide phase is depending on their distance from horizontal and vertical walls of the concrete test section. Oxidation mechanisms are mainly depending on two parameters: nature of the metallic interface and localization in the test section. Calculations at thermodynamic equilibrium show that the only product from steel oxidation is (Fe,Cr)3O4, Cr2O3 is never formed. Moreover taking into account the two gaseous species coming from the concrete (CO2 and H2O), considered up to now as being the only sources of oxidation of the metallic phase, the steel oxidation proportion is estimated at only 14% at thermodynamic equilibrium meaning that complementary phenomena are involved during MCCI.

OSIRIS—Gamma-ray spectroscopy software for on-site inspections under the Comprehensive Nuclear-Test-Ban Treaty

We have designed and tested software for the acquisition and analysis of high-resolution gamma-ray spectra during on-site inspections under the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The On-Site Inspection RadioIsotopic Spectroscopy—OSIRIS—software filters the spectral data to display only radioisotopic information relevant to CTBT on-site inspections, e.g.,131I. A set of over 100 fission-product spectra was employed for OSIRIS testing. These spectra were measured where possible, or generated by modeling. The test spectral compositions include non-nuclear-explosion scenarios, e.g., a severe nuclear reactor accident, and nuclear-explosion scenarios such as a vented underground nuclear test. Comparing its computer-based analyses to expert visual analyses of the test spectra, OSIRIS correctly identifies CTBT-relevant fission product isotopes at the 95% level or better.

가압중수로형원전의 중대사고 대응능력 연구

The realistic cases causing severe core damage should be analyzed and arranged systematically for preparing an accident management of the specific nuclear power plant. The objective of this paper is to establish basic technical information for reactor safety and reactor building integrity management strategies in CANDU reactor severe accident. For the development of severe accident management strategies, plant specific features and behaviors must be studied by detailed analysis works. This analysis scope will serve to cover overall methods and analyzing results to understand the reactor building integrity status in the most likely severe accident sequences that could occur at CANDU reactor. Also analysis results could help prevent or mitigate severe accidents for the identification of any plant specific vulnerabilities to severe accidents using the probabilistic safety assessment (PSA) quantified results.

Chemical form consideration of released fission products from irradiated fast reactor fuels during overheating

Experiments simulating overheating conditions of fast reactor severe accidents have been previously carried out with irradiated fuels. For the present study, the chemical forms of the fission products (FPs) included in the irradiated fuels were evaluated by thermochemical equilibrium calculations. At temperatures of 2773 K and 2973 K, the most stable forms of Cs, I, Te, Sb, Pd and Ag are gaseous compounds, and their fractions are about 100% in an oxygen partial pressure range between 10−5 and 103 Pa. Cs and Sb detected in the thermal gradient tube (TGT) in the experiments can take gaseous chemical forms of elemental Cs, CsI, Cs2MoO4, CsO and elemental Sb, SbO, SbTe, respectively. The chemical forms of Cs and Mo are sensitive to the oxygen partial pressure. By comparing experimental results and the estimations, it is seen CsI thermochemically behaves in a manner that traps it in the TGT, while elemental Cs trends to move as fine particles. The moving behavior of the gaseous FPs will obey not only thermochemical principles, but also those of particle dynamics.

Iodine–paint interactions during nuclear reactor severe accidents

To assess the radiological consequences of a severe reactor accident, it is important to be able to predict the behaviour of iodine in containment. Some interactions between iodine and containment paint (e.g., adsorption) have been well known for a long time. However, in recent years, new phenomena have been identified that can affect the gas phase iodine concentration in the longer term (e.g., the release of molecular iodine and organic iodides from irradiated painted surfaces). Several international collaborations and organizations around the world are currently addressing different aspects of this topic, including laboratory experiments and theoretical studies (ab initio) designed to improve the mechanistic understanding of the phenomena. Knowledge of the underlying mechanisms will provide explanations for behavioural differences observed between paint types, and will support the extrapolation of laboratory results to the safety analyses of nuclear reactors.The purpose of this paper is to present a selection of recent work performed by Severe Accident Research Network (SARNET) members regarding iodine–paint interactions and paint aging in order to improve the common understanding and better define what has still to be done in this area. The Severe Accident Research Network (SARNET) provides a framework within which members can share and discuss results.

Modeling corium jet breakup in water pool and application to ex-vessel fuel–coolant interaction analyses

In light water reactor core melt accidents, the molten fuel can be brought into contact with coolant water in the course of the melt relocation in-vessel and ex-vessel as well as in an accident mitigation action of water addition. For the last several decades, the potential risk of energetic molten fuel coolant interactions (FCIs, steam explosions) has drawn substantial attention in the safety analysis of reactor severe accidents. In this paper, an improved melt jet breakup model is presented and analyses of an energetic fuel–coolant interaction in a PWR cavity (1) partially filled (4m deep) and (2) completely filled (7m deep) with water are presented. The TRACER-II code was used in the analyses. For jet breakup model, the full dispersion equation of Kelvin–Helmholtz instability for the melt jet–vapor film–water was solved numerically and the solutions were correlated for use in the TRACER-II code. The new jet breakup model was benchmarked using FARO L28 test data. In reactor calculations the mixing calculations showed that the average melt drop size was much smaller in 4m deep pool with 3m free-fall than in 7m deep pool. The explosion calculations showed that the peak pressure at the center of mixture was ∼90MPa in 4m deep pool, ∼25MPa in 7m deep pool. It also showed that the maximum impulse at the cavity wall was found at the lower wall in both cases and it was 50kPas in 4m deep pool and 150kPas in 7m deep pool.

Hydrodynamic fine fragmentation of partly solidified melt droplets during a vapour explosion

In this paper we investigate the fine fragmentation criterion for partly solidified droplets during the explosion phase of a vapour explosion. First, a modified Weber number is proposed for the determination of the critical conditions for the fine fragmentation. Next, this introduced dimensionless number, which considers the crust stiffness as a stabilizing force acting to retain the crust in the presence of hydrodynamic forces, is compared with experimental data. It is further investigated to see whether the limitation of the fine fragmentation by the crusts can help explain the strongly reduced explosion strength observed in experiments with corium as compared to alumina. Finally, the conclusions are used to assess the limitation due to solidification on the strength of the explosions that could occur during severe accidents in nuclear reactors.

Numerical investigation on melt freezing behavior in a tube by MPS method

Given the nuclear reactor severe accident, the instrument tubes of reactor vessel bottom head may fail and create paths through which melt could leak out of the reactor vessel wall. Thus, the melt freezing behavior in an instrument tube is a key factor that concerns the failure of the reactor pressure boundary. In this study, the moving particle semi-implicit (MPS) method was adopted to analyze the melt penetration and solidification behaviors in a tube. Both surface tension and viscosity variation with temperature were taken into account in the present MPS method. The numerical results had been compared with the downward and upward melt injection experiments, respectively. The comparative results showed that the melt leading edge position histories were in good agreement with the experimental results. The typical melt freezing behaviors were successfully reproduced by MPS method. The crust formed on the surface of the tube increased the melt flowing resistance, and would also increase the thermal resistance between the melt and the tube. Meanwhile, the melt velocity also decreased due to the increase of its viscosity. The present results indicate that MPS method is applicable to simulate the melt penetration behavior in an instrument tube.

Uncertainty assessment of neutron resonance transmission analysis using the linear absorption model

Neutron resonance densitometry (NRD) has been proposed to quantify nuclear materials in particle-like debris of melted spent nuclear fuel formed in severe nuclear reactor accidents. The NRD is a hybrid technique of the neutron resonance transmission analysis and the neutron resonance capture analysis. We have studied the statistical uncertainties of the neutron resonance transmission analysis by using the linear absorption model. Particularly, the effects of impurities and sample thickness on the uncertainties were examined.