Severe Accident Conditions Research Articles

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2 -zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

多次元ナトリウム燃焼解析コードAQUA-SFの過酷事故解析への適用

多次元ナトリウム燃焼解析コードAQUA-SFの過酷事故解析への適用

Investigation of the structure of debris beds formed from fuel rods fragmentation

This paper is a study of debris beds that can form in the core of a nuclear power plant under severe accident conditions. Such beds are formed of fragments of pellets and cladding remnants, as observed in the TMI-2 core. Many important issues are related with the morphology of those debris beds: are they coolable in case of water injection and how does molten corium progress through them if they are not coolable? The answers to those questions depend on the structure of the debris bed: porosity, number and arrangement of particles. In order to obtain relevant information, a numerical simulation of the formation of the debris bed is proposed. It relies on a granular approach of the type called “Contact Dynamics” to simulate the collapse of debris and their accumulation. Two different schemes of fuel pellet fragmentation are considered and simulations for different degrees of fragmentation of the pellets are performed. The results show that the number of axial cracks on fuel pellets strongly influences the final porosity of the debris bed. Porosities vary between 31% (less coolable cases) and 45% (similar to TMI-2 observations), with a most probable configuration around 41%. The specific surface of the bed is also evaluated. In the last part, a simple model is used to estimate the impact of the variation in geometry of the numeric debris beds on their flow properties. We show that the permeability and passability can vary respectively with a range of 30% and 15% depending on the number of fragment per pellet. The other benefits of the approach are finally discussed. Among them, the possibility to print 3D samples from the calculated images of debris beds appears as a promising perspective to perform experiments with realistic debris beds.

Overview of the VERDON-ISTP Program and main insights from the VERDON-2 air ingress test

In order to reduce the uncertainties concerning source term assessment under LWR severe accident conditions, the International Source Term Program has been launched. Four main R&D research axes have been addressed in this program: iodine behaviour in the RCS and containment, study of the boron carbide effect on fuel degradation and FP release, study of the air effect on fuel and FP behaviour, FP release from high burn-up UO2 and MOX fuels.As far as the source term quantification is concerned, four VERDON tests have been defined and were performed in the VERDON laboratory at the CEA Cadarache Centre. We present an overview of these four tests, devoted to the study of FP release from high burn-up UO2 and MOX fuels. We focus on the VERDON-2 air ingress test performed on MOX fuel with a specific loop dedicated to FP release and transport. It addressed two issues: Ru release under mixed steam-air conditions following lower head failure and transport/re-volatilization of Ru and volatile FPs in the primary circuit. The latter was performed thanks to the implementation of thermal gradient tubes exposed at different phases of the sequence. In this configuration, the VERDON facility combines the ability of the analytical approach to identify and understand the physical phenomena involved, with a semi-integral approach, which allows the results to be applied to source term assessments for reactor application. This VERDON-2 test has shown a rather unexpected low release of ruthenium from the MOX sample and no re-volatilization from the thermal gradient tubes. On the contrary, a significant re-volatilization of iodine has been observed when switching to mixed steam-air conditions, as well as a low, but measurable amount of iodine transported under gaseous form downstream of the thermal gradient tubes.

Phase-Field Modeling of Binary Eutectic Alloy Solidification with Convection

In the framework of a description of melt pool heat transfer under severe accident conditions, we introduce a computational fluid dynamics approach for the phase change based on the phase-field method. The approach is derived using the formalism of irreversible thermodynamics and depends on a phenomenological expression for the free energy of binary eutectic alloys. The free energy is constructed to describe sharp interfaces on sufficiently small length scales and is capable of representing the appearance of mushy layers in a volume-averaged large-scale perspective. In particular, a dynamic calculation procedure for the diffuse interface width is introduced based on free energy minimization. Numerical simulations using this approach are performed and compared with experimental and numerical results from the literature. These comparisons demonstrate that the new model improves numerical simulation results and is able to describe the dynamics of sharp and diffuse interfaces.

Lessons from Chernobyl and Fukushima and Advanced Control Concepts for Severe Accidents

The year 2016 marks the 30th anniversary of the accident in the No. 4 unit of the Chernobyl NPP and the 5th anniversary of the accident at the Fukushima NPP in Japan. The Chernobyl accident substantially changed the public’s attitude toward severe accidents at NPPs, promoting on the one hand modernization and improvement of safety systems and on the other hand a higher safety culture by harmonization of standards and regulations, and adoption of deeply echeloned protection and a scientifically validated approach to the analysis of accidents. The Fukushima accident revealed a new class of accidents – extreme actions caused by naturally occurring cataclysms and/or different external actions, which actually are identical to loss of a large part or all of the means for controlling an accident. The present article examines the main lessons on accidents in terms of understanding the phenomenology of the events occurring during severe accidents and confirms the importance of the deterministic component of the safety concept. At the same time, it emphasizes that special attention must be given to hydrogen safety, systems for prolonged removal of residual heat, emergency readiness, and emergency response under the extreme conditions of severe accidents. The computational means used to develop accident counter-measures must be properly set to take account of the design features of the reactor facility, the containment of the NPP, and the adjoining territories in order to determine the time required to implement the procedures for controlling accidents and mitigating the consequences.

The study of cesium ionization at high gas temperature in a primary circuit

A kinetic model has been presented in developing a better understanding of the mechanism behind the ions formation at high gas temperature in a primary circuit in conditions of severe accidents. The formation of ions is triggered by the thermal ionization of cesium atoms and irradiation of a hot steam. The kinetics of ions is considered in combination with the neutral chemistry of cesium and iodine species. Modeling results are presented for temperatures 2000 and 1200K and a dose rate of 1000kGy/s. The evolution of the concentrations of the most abundant ions and neutral reaction products are calculated.

Effects of subcooling on downward facing boiling heat transfer with micro-porous coating formed by Cold Spray technique

External reactor vessel cooling (ERVC) is an effective strategy to achieve in-vessel retention (IVR) of core melt in the reactor pressure vessel (RPV) under severe accident conditions. Among the available strategies, micro-porous coating technique has been known to enhance the thermal margin of the RPV. In this study, a new and versatile micro-porous coating technique applicable to commercial size reactors known as “Cold Spray” has been developed to coat a hemispherical test vessel. Quenching boiling experiments at different degrees of subcooling (10°C, 5°C, 3°C, 1°C, and 0°C) were performed using bare and micro-porous coated vessels. Visual observations of the quenching process along with quantitative analyses of the boiling data were performed. It was found that the critical heat flux (CHF) limit varies significantly with the angular location at all subcooled conditions. Higher cooling rates and CHF limits were obtained with higher degrees of subcooling. A micro-porous coating formed by Cold Spray significantly improved the CHF limit compared to the bare vessel. In fact, nearly 90% enhancement was achieved using the Cold Spray coated vessel. CHF correlations for both bare and micro-porous coated vessel have been proposed capturing the effects of subcooling and angular variation along the outer surface of the hemispherical test vessels.

Radionuclide Disperse To Environtment From Pwr Reactor On Severe Accident Condition Using Maccs Program

<p>The atmosphere is an important pathway in the transfer of radionuclides from nuclear power plants into the environment and population. Acceptance of radiation dose to the environment and population affected by the radionuclides release and site conditions surrounding of the nuclear power plant. The radionuclides release in the atmosphere is determined by the dispersion coefficient parameter. The aim of this paper is to obtain dispersion coefficient and radionuclide released in Sebagin (West Bangka district) caused by severe accident condition from the PWR Nuclear Power Plant. Dispersion analysis of radionuclides into the environment from nuclear power PWR on severe accident conditions have been done using MACCS program. Reference for the calculation of source term fraction is selected from calculation results of the MELCOR computer code and it is implemented to PWR reactors Westinghouse 3411 MWth subject. The calculation of radionuclides release performed using MACCS program for aspiring nuclear power plant site in West Bangka. Simulation calculations for the area radius from 0.80 kmup to 20 km from the nuclear power plant site are performed. Meteorological datas used in calculation are the meteorology data from Sebagin meteorological stations for the years of 2012 period. The result is the dispersion coefficient decreases as a function of time and distance. The concentration of radionuclides through soil pathway decreases as a function of the distance, and the dominant contributor of radionuclide radiation Xe-133 and I-131. Radionuclide concentrations obtained through the air pathway decreases as a function of distance, and dominant contributors of radionuclide radiation is contributed also from I-131 and Xe-133. The presence of I-131 radionuclides are giving dangerous to humans, it is necessary to further treatment for prevent its impacts. </p>

Enhancement of Downward-Facing Saturated Boiling Heat Transfer by the Cold Spray Technique

Abstract In-vessel retention by passive external reactor vessel cooling under severe accident conditions is a viable approach for retention of radioactive core melt within the reactor vessel. In this study, a new and versatile coating technique known as “cold spray” that can readily be applied to operating and advanced reactors was developed to form a microporous coating on the outer surface of a simulated reactor lower head. Quenching experiments were performed under simulated in-vessel retention by passive external reactor vessel cooling conditions using test vessels with and without cold spray coatings. Quantitative measurements show that for all angular locations on the vessel outer surface, the local critical heat flux (CHF) values for the coated vessel were consistently higher than the corresponding CHF values for the bare vessel. However, it was also observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit vary appreciably along the outer surface of the test vessel. Nonetheless, results of this intriguing study clearly show that the use of cold spray coatings could enhance the local CHF limit for downward-facing boiling by > 88%.

Analysis of flammability in the attached buildings to containment under severe accident conditions

Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment.The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied.The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07.The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H2 and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas.The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the analyzed exercise provided three conditions are met: H2 and CO recombiner devices are found inside the containment; corium is submerged and cooled down to quenching by flooding the reactor cavity; and the containment remains isolated along the accident evolution so that gases flowing into attached buildings to containment are limited to the so-called allowable leakage.

Two-Dimensional Axisymmetric Thermostructural Analysis of APR1400 Reactor Vessel Lower Head for Full-Core Meltdown Accident

In severe accident conditions, the molten core material forms an internally heated debris bed and eventually becomes a molten pool of corium, which will cause or induce thermal and mechanical loads to the reactor vessel lower head (RVLH) resulting in penetrations leading to failure. A good understanding of the mechanical behavior of the RVLH is essential for estimating structural integrity and improving accident mitigation strategies.Coupled thermomechanical analysis using ANSYS, a general-purpose finite element method analysis code, was used to evaluate the possibility and timescale of failure. A two-dimensional axisymmetric finite element model was adopted based on APR1400 design data with relevant material properties including creep data.From the study, it was found that the possibility of plastic and creep failure of the RVLH for the APR1400 was considerably low for a full-core meltdown of the reactor core under ex-vessel cooling conditions with an ambient temperature of 130°C and constant decay heat from the corium, but the lower head may fail unless the increased internal pressure can be reduced on time. Plastic failure can be a major cause of lower head failure of a reactor vessel in high internal pressure conditions and creep failure is not negligible, since failure mechanisms under long-lasting periods are considered. This study found that the APR1400 RVLH failure time is around 220 h using 15% creep strain failure criteria from the postulated accident condition.

The effect of self-leveling on debris bed coolability under severe accident conditions

Nordic-type boiling water reactors employ melt fragmentation, quenching, and long term cooling of the debris bed in a deep pool of water under the reactor vessel as a severe accident (SA) mitigation strategy. The height and shape of the bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry depends on its formation process (melt release, fragmentation, sedimentation and settlement on the containment basemat), but it also changes with time afterwards, due to particle redistribution promoted by coolant flow (self-leveling). The ultimate goal of this work is to develop an approach to the assessment of the probability that debris in such a variable-shape bed can reach re-melting (which means failure of SA mitigation strategy), i.e. the time necessary for the slumping debris bed to reach a coolable configuration is larger than the time necessary for the debris to reach the re-melting temperature. For this purpose, previously developed models for particulate debris spreading by self-leveling and debris bed dryout are combined to assess the time necessary to reach a coolable state and evaluate its uncertainty. Sensitivity analysis was performed to screen out less important input parameters, after which Monte Carlo simulation was carried out in order to collect statistical characteristics of the coolability time. The obtained results suggest that, given the parameters ranges typical of Nordic BWRs, only a small fraction of debris beds configurations exhibits the occurrence of dryout. Of the initially non-coolable configurations, a significant portion becomes coolable due to debris bed self-leveling.

Effectiveness of the debris bed self-leveling under severe accident conditions

Melt fragmentation, quenching and long term coolability in a deep pool of water under the reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. The success of such strategy is contingent upon the natural circulation effectiveness in removing the decay heat generated in the porous debris bed. The maximum height of the bed is one of the important factors which affect the debris coolability. The two-phase flow within the bed generates mechanical energy which can change the geometry of the debris bed by the “self-leveling” phenomenon. In this work we developed an approach to modeling of the self-leveling phenomenon. Sensitivity analysis was carried out to rank the importance of the model uncertainties and uncertain input parameters i.e. the conditions of the accident scenario and the debris bed properties. The results provided some useful insights for further improvement of the model and reduction of the output uncertainties through separate-effect experimental studies. Finally, we assessed the self-leveling effectiveness, quantified its uncertainties in prototypic severe accident conditions and demonstrated that the effect of self-leveling phenomenon is robust with respect to the considered input uncertainties.

Chemical State Mapping of Degraded B4C Control Rod Investigated with Soft X-ray Emission Spectrometer in Electron Probe Micro-analysis

B4C is widely used as control rods in light water reactors, such as the Fukushima Daiichi nuclear power plant, because it shows excellent neutron absorption and has a high melting point. However, B4C can melt at lower temperatures owing to eutectic interactions with stainless steel and can even evaporate by reacting with high-temperature steam under severe accident conditions. To reduce the risk of recriticality, a precise understanding of the location and chemical state of B in the melt core is necessary. Here we show that a novel soft X-ray emission spectrometer in electron probe microanalysis can help to obtain a chemical state map of B in a modeled control rod after a high-temperature steam oxidation test.

Analysis of containment venting strategy under severe accident conditions

Under severe accidents without containment heat removal, the containment integrity can be challenged due to over-pressurization by the steam and non-condensable gas generation. Containment filtered venting has been considered as an effective measure to avoid or delay the containment failure by over-pressurization. In order to minimize the environmental effects and maximize the effectiveness of filtered venting, it is critical to initiate venting in timely manners with sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to minimize the space and the cost requirement. In this study, the various venting strategies are investigated with respect to possible release flow characteristics. First of all, the accident scenarios representing the core meltdown accident with containment pressurization are selected by reviewing relevant literature. With those scenarios, thermal hydraulic behaviors in the containment and the discharge flow depending on the different venting strategies (i.e. vent line size and vent initiation pressure) are analyzed. MAAP5 model for the OPR1000 Korean nuclear power plant has been used for simulation.

비 가시 환경에서의 LRF와 CCD 카메라의 성능비교

In this paper, range measurement performance of LRF (Laser Range Finder) module and image contrast of color CCD camera are evaluated under the aerosol (high temperature steam) environments, which are simulated severe accident conditions of the LWR (Light-Water-Reactor) nuclear power plant. Data of LRF and color CCD camera are key informations, which are needed in the implementation of SLAM (Simultaneous Localization and Mapping) function for emergency response robot system to cope with urgently accidents of the nuclear power plant.

Fission products behaviour in UO2 submitted to nuclear severe accident conditions

The objective of this work was to study the molybdenum chemistry in UO2 based materials, known as SIMFUELS. These materials could be used as an alternative to irradiated nuclear fuels in the study of fission products behaviour during a nuclear severe accident. UO2 samples doped with 12 stable isotopes of fission products were submitted to annealing tests in conditions representative to intermediate steps of severe accidents. Samples were characterized by SEM-EDS and XAS. It was found that Mo chemistry seems to be more complex than what is normally estimated by thermodynamic calculations: XAS spectra indicate the presence of Mo species such as metallic Mo, MoO2, MoO3 and Cs2MoO4.

THAI experimental programme for containment safety assessment under severe accident conditions

Abstract The THAI (THAI = Thermal hydraulics, Hydrogen, Aerosols, Iodine) experimental programme aims to address open questions concerning the behavior of hydrogen, iodine and aerosols in the containment of water cooled reactors. Since its construction in 2000, THAI programme is being performed in the frame of various national projects (sponsored by German Federal Ministry for Economic Affairs and Energy, BMWi) and two international joint projects (under auspices of OECD/NEA). THAI experimental data have been widely used for the validation and further development of Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes with 3D capabilities. Selected examples of code benchmark exercises performed based on the THAI data include; hydrogen distribution experiment (ISP-47 and OECD/NEA THAI code benchmark), hydrogen combustion behaviour (ISP-49), hydrogen mitigation by PARs (OECD/NEA THAI-2 code benchmark), iodine/surface interactions, iodine mass transfer, and iodine transport and multi-compartment behaviour (EU-SARNET and EU-SARNET2), thermal-hydraulic tests (German CFD-network). In the present paper, a brief overview on the THAI experiments and their role in the containment safety assessment is discussed.

Calculation of PDS-XADS core closed-loop transfer function by using feedback with the lumped-model

Abstract In this paper, the PDS-XADS LBE-cooled core open-loop transfer function was calculated by considering the source importance in point-kinetic equations. For this purpose, the overall-feedback transfer function was calculated considering the lumped-model for 14-steps of subcritical levels. Following effects were considered in three steps: 1. Doppler broadening, fuel expansion, coolant density and structure expansion, 2. Delayed-reactivity and void-worth inserted to prior step, 3. Severe-accident condition, inserted to prior steps. The linear stability analysis was modeled by using the Bode diagrams, Nyquist stability criterion and Nichols chart in MATLAB for each subcritical level and six groups of delayed neutrons. For optimized subcritical level determination, a conservative severe accident was considered. According to calculation results and analysis, the PDS-XADS core is stable and in optimized subcritical level, has the higher safety margin. The results are in good agreement with SIMMER-III code and main neutronic results. The optimized subcritical level by using the lumped-model is 0.97687.