Molten Core Research Articles

A simple model to understand physics of melt coolability under bottom flooding

Following a severe accident in a nuclear reactor, stabilization and termination of molten pool/debris is necessary to ensure safety of the reactor. In this regard, ex-vessel corium coolability still remains an unresolved issue in spite of several efforts being taken towards its understanding. Extensive research has been carried out in the field of corium coolability phenomenon, but the outcome is far from establishing a certainty of complete coolability of melt under all possible scenarios. Major thrust so far was on top flooding condition, which often results in uncoolable condition due to inability of water ingression deep into the solid crust. The research on coolability under bottom flooding condition started long back, but has gained lot of momentum recently. However, mainly the focus is on development of engineering system to enhance the coolability using bottom flooding. The understanding of the phenomenon is not yet totally brought out. In this paper, we have postulated the mechanism of melt coolability under bottom flooding. We have developed a simplified mechanistic model for prediction of coolability of molten corium under bottom flooding condition based on the postulations. The model has been validated with the experimental data from literature. Comparison of top flooding vis-à-vis bottom flooding has also been presented.

Distribution of fission product residual decay heat in stratified core melt of LWR and its influence on sidewall heat flux

The model of spatial distribution of fission product (FP) activity and decay heat in core melt of light water reactor under severe accident conditions is proposed. It is based on thermodynamical consideration of “U–Zr–O–Fe–FP” system as a two-phase (oxide and metallic) stratified melt. The thermochemistry of FP phase distribution, heat-mass transfer and radioactive transformations of ∼100 isotopes of Cs, Mo, Ru, Ba, Sr, Zr, La, Ce, Eu, Nd, Nb and Sb are considered self-consistently. The calculations carried out with specially designed code FPMC (fission products–molten core) showed that up to 15–25% of decay heat may be released in the metallic phase of the melt. Estimation of this value is fundamentally important in simulations of heat and mass transfer in stratified melt, such as in-vessel melt retention and reactor lower head failure. For numerical assessment of the influence of decay heat spatial distribution in stratified molten pool on the lower head failure time the multivariate simulation of the later stage of large break loss-of-coolant accident in VVER-1000 reactor was carried out by SOCRAT and HEFEST codes coupled with FPMC.

Design of integrated passive safety system (IPSS) for ultimate passive safety of nuclear power plants

The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the containment outside into integrated passive safety tanks (IPSTs). The coolant in the IPSTs is used for various functions in accident scenarios. Also, potential problems for the realistic installation of the IPSS are proposed and the solutions to these problems are schematically described. IPSS is the design for the passive safety enhancement in preparation for a loss of AC power. Consequently, it is designed for the supplementation and enhancement of current nuclear power plants, not as a replacement. The specific optimization design for each current or future reactor will be studied as further works.

Numerical Evaluation of the Cooling Performance of a Core Catcher Test Facility

A core catcher is considered as a promising engineered system to stabilize the molten corium in the containment during a postulated severe accident in a nuclear power plant. Conceptually, the core catcher consists of a carbon steel body, sacrificial material, protection material, and engineered cooling channel. The cooling capacity of the engineered cooling channel should be guaranteed to remove the decay heat of the molten corium. The flow in ex-vessel core catcher is a combined problem of a two-phase flow in the engineered cooling channel and a single-phase natural circulation in the whole core catcher system. In this study, the analysis of the test facility for the core catcher using the CUPID code, which is a three-dimensional thermal-hydraulic code for the simulation of two-phase flows, was carried out to evaluate its cooling capacity.

Spinel-derived single mode optical fiber

A silica-based highly acoustically-anti-guiding optical fiber, fabricated using a molten core approach employing spinel (MgAl2O4) for the first time, is presented. To our knowledge, this is the first truly single mode optical fiber fabricated from a precursor crystal. It is shown that MgO increases the acoustic velocity when added to silica (some physical parameters of MgO are identified) and that the Brillouin gain in the core is less than one third that in the cladding in one of the fibers. This results from a massive acoustic waveguide attenuation term that broadens the spectrum to well over 200 MHz. For the first time, to the best of our knowledge, this also enabled the validation of the intrinsic Brillouin line-width (~20 MHz) of pure silica in fiber form via a direct measurement.

Simulation of external reactor vessel cooling in a lumped-parameter code

Abstract In the course of a severe accident in a nuclear power plant the integrity of the reactor vessel may be preserved by External Reactor Vessel Cooling (ERVC). By flooding the cavity surrounding the reactor vessel the decay heat is removed from molten core material through the vessel wall to water at the submerged vessel surface. In order to simulate the boiling process at the external vessel wall during ERVC in the framework of lumped parameter codes a model based on a boiling model used in the code ANSYS CFX 13.0 is developed.

The Upside of Down: Disaster and the Imagination 50 Years On

The Upside of Down: Disaster and the Imagination 50 Years On

Feasibility of flooding the reactor cavity with liquid gallium coolant for IVR-ERVC strategy

In this paper, a new approach replacing the ERVC coolant by a liquid metal instead of water is studied to avoid the heat removal limit of CHF during boiling of water. As the flooding material, gallium is used in terms of the melting and boiling points. Gallium has the enough low melting point of ∼29.7°C to ensure to maintain liquid state within the containment building. A gallium storage tank for the new flooding system of the ERVC is located in higher position than one of the reactor cavity to make a passive system using the gravity for the event of a station blackout (SBO). While the decay heat from the reactor vessel is removed by gallium, the borated water which is coming out from the reactor system plays a role as the ultimate heat sink in this ERVC system. In the system, two configurations of gallium and borated water are devised depending on whether the direct contact between them occurs. In the first configuration, two fluids are separated by the block structure. The decay heat is transported from molten corium to gallium through the vessel wall. Then the heat is ultimately dissipated by boiling of water in the block structure surface facing the borated water. In the second configuration, the cavity is flooded with both borated water and gallium in the same reactor cavity space. As the result, two layers of the fluids are naturally formed by the density difference. Like the first configuration, finally the heat removal is achieved by boiling of water via gallium. The CFD analysis shows that the maximum temperature of gallium is much lower than its boiling point while the natural circulation is stably formed in two types of the configurations without any serious risk of thermal limit.

Analysis on containment depressurization under severe accidents for a Chinese 1000 MWe NPP

Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment.

Effect of oxygen-to-metal ratio on properties of corium prepared from UO2 and zircaloy-2

The UO2 and zircaloy system has been studied in severe accident research because these materials form most of a molten corium. In addition, during core meltdown, it is expected that water vapor and hydrogen which is produced in the H2O–zircaloy reaction fill the reactor core space. In this study, simulated corium specimens were prepared considering the ambient atmosphere in an accident, and adjustment of the oxygen-to-metal (O/M) ratio. Phase observations by XRD and EPMA and measurement of thermal conductivity were then done on the specimens. It was confirmed that the prepared O/M ratio had hardly any effect on melting temperature although UO2+x has obvious dependency on the O/M ratio. Thermal conductivity of the molten specimens showed little dependency on the O/M ratio and temperature. It appeared that the thermal conductivity had already been significantly decreased by the solution of U and Zr. Microstructure information as lattice constant and phase segregation was obtained by XRD and EPMA observation.These properties will be basic data for the development of conditioning techniques for an actual corium.

20813 BWR複雑流路内における溶融燃料落下挙動(086-4【熱流体工学のフロンティア(4)】)

It is important to estimate the behavior of molten core jet in RPV (Reactor Pressure Vessel) of BWR when severe accident occurred as can be seen from Fukushima Danchi nuclear power plant accident. Thus we are developing a simulation code to estimate the behavior of molten core falling in a lower plenum of BWR. Since there are many complicated structures in the plenum, it is expected that molten core would be injected in the complicated flow channel. The objective of the present study is to investigate the influence of complicated structures m a lower plenum on jet by experiment. To investigate it, we observe jet injection which simulated the molten core falling accident experimentally. We constructed a model of a lower plenum and a steady jet injection equipment. Results of preliminary experiment showed that jet behavior is affected by complicated flow structure

S083021 MAAP5コードによるBWRプラントのシビアアクシデント解析([S083-02]原子力システムおよび要素技術)

Severe accident analysis was performed for a station blackout (TBU) sequence in a typical BWR-5 plant with modified Mark-II type containment by MAAP5.01 code. Core heatup, degradation and relocation processes, reactor vessel failure and resultant impact on the containment including hydrogen production were investigated. Sensitivity analysis focusing on direct contact heating (DCH), which affect on the consequence of TBU sequence, was also performed. If extensive DCH does not occur in the TBU sequence, failure of the containment vessel can be postponed. On the other hand, concrete ablation at a floor and a side wall inside the pedestal due to molten core - concrete interaction (MCCI) significantly increases, because a large amount of debris with high temperature stays inside the pedestal.

C114 格子ボルツマン法HCZモデルによるジェットブレイクアップ挙動評価 : (2)ブレイクアップに及ぼす周囲流動場の影響(OS9 熱・流動(2))

In order to understand the jet breakup behavior of the molten core material into the coolant during a core disruptive accident (CDA) for a sodium-cooled fast reactor (SFR), we simulated the breakup using the lattice Boltzmann (LB) HCZ model The surface and fragmentation behavior was investigated in detail in the simulation results for examining the effect of ambient fluid around the jet As a result, we observed that the jet breaks up by fragmentation which occurs at side of the jet Thus, it was confirmed that the mechanism of the jet breakup behavior follows that of the Epstein's correlation when the dominant phenomenon for the jet break up is hydrodynamic fragmentation

20812 溶融物のジェットブレークアップ挙動に及ぼす微粒化の影響(086-4【熱流体工学のフロンティア(4)】)

For the safety design of a Fast Breeder Reactor (FBR), Post Accident Heat Removal (PAHR) is required when a hypothetical Core Disruptive Accident (CDA) occurs. In PAHR, it is necessary to understand the interaction between molten core material and coolant m order to estimate whether the molten material jet is completely solidified by the coolant. The objective of the present study is to clarify the dominant factors of the fragmentation behavior on the jet breakup. In this study, we inject molten material and transparent fluid into coolant at free fall speed. The jet breakup and fragmentation behavior of the molten material and the transparent liquid are observed with a high speed video camera. In addition, we simulate liquid jet using the lattice Boltzmann method m order to investigate the jet breakup and the fragmentation behavior, which are difficult to be visualized or measured in the experiment.

A Study on Mechanism of Early Failure of Inner Duct Wall within Fuel Subassembly with High Heat Flux from Molten Core Materials based on Analysis of an EAGLE Experiment Simulating Core Disruptive Accidents in an LMFBR

A Study on Mechanism of Early Failure of Inner Duct Wall within Fuel Subassembly with High Heat Flux from Molten Core Materials based on Analysis of an EAGLE Experiment Simulating Core Disruptive Accidents in an LMFBR

Modelling dynamics in protein crystal structures by ensemble refinement

Single-structure models derived from X-ray data do not adequately account for the inherent, functionally important dynamics of protein molecules. We generated ensembles of structures by time-averaged refinement, where local molecular vibrations were sampled by molecular-dynamics (MD) simulation whilst global disorder was partitioned into an underlying overall translation-libration-screw (TLS) model. Modeling of 20 protein datasets at 1.1-3.1 Å resolution reduced cross-validated R(free) values by 0.3-4.9%, indicating that ensemble models fit the X-ray data better than single structures. The ensembles revealed that, while most proteins display a well-ordered core, some proteins exhibit a 'molten core' likely supporting functionally important dynamics in ligand binding, enzyme activity and protomer assembly. Order-disorder changes in HIV protease indicate a mechanism of entropy compensation for ordering the catalytic residues upon ligand binding by disordering specific core residues. Thus, ensemble refinement extracts dynamical details from the X-ray data that allow a more comprehensive understanding of structure-dynamics-function relationships.DOI:http://dx.doi.org/10.7554/eLife.00311.001.

Ex-vessel molten core solidification behavior by moving particle semi-implicit method

Development of analysis code for predicting the molten core solidification behavior during the severe accident in nuclear power plants is necessary. Although agreeable results have been achieved using existing codes such as CORFLOW and LAVA to analyze some melt spreading experiments, these codes have a common problem of versatility because they all use empirical equations such as mass flux depending on distance from wall to calculate free surface of fluid. Moving particle semi-implicit (MPS) method which calculates free surface without empirical equations is suitable for analyzing the solidification behavior of fluid with large deformation. In this work, the MPS code for incompressible fluid was developed for calculating thermal field, solid-liquid phase transition, and temperature dependence of viscosity. Based on the MPS model, the FARO-L26S core melt solidification experiment was analyzed. For the leading edge position of the melt in this experiment, three-dimensional MPS simulations predicted agreeable multistage curve results with the experiment.

Radiation heat transfer in a pressurized water reactor lower head filled with molten corium

Following a core melt, molten material may slump to the lower head of a reactor pressure vessel (RPV). In that case, some structures like lower parts of fuel elements and a core support plate would remain intact. Since the melt is at high temperature and there are no obstacles between the melt and the supporting plate, the plate is exposed to an intense radiation heating. The radiation heat exchange model of the lower head was developed and applied to a finite element code COUPLE which is a part of the detailed mechanistic code RELAP5/SCDAPSIM. The radiation enclosure consisted of three surfaces: the upper surface of the relocated material, the inner surface of the RPV wall above the relocated material and the lower surface of the core support plate. View factors were calculated for the enclosure geometry that is changing in time because of intermittent accumulation of molten material. The enclosure surfaces were covered by mesh of polygonal areas and view factors were calculated, for each pair of the element areas, by solving the definite integrals using the algorithms for adaptive integrations by means of Gaussian quadrature. Algebraic equations for radiosity and irradiation vectors were solved by LU decomposition and the radiation model was explicitly coupled with the heat conduction model. The results show that there is a possibility of the core support plate failure after being heated up due to radiation heat exchange with the melt.

Infinite multiplication factor of low-enriched UO2–concrete system

The possibility of criticality of fuel debris in a form of uranium dioxide (UO2)–concrete mixture is evaluated by calculating the infinite multiplication factor (k ∞) for a study of criticality control on the fuel debris generated through the molten core concrete interaction in a severe accident of a light water reactor. The infinite multiplication factor can be greater than unity, which means that handling of the mixture is subject to criticality control. This paper shows that concrete provides efficient neutron moderation and points out the necessity of further investigations on the criticality of UO2–concrete system for actual handling of fuel debris.

직접냉각방식 및 간접냉각방식 Core Catcher의 성능비교

The European nuclear design requirements, which should be satisfied by nuclear reactors in Europe, usually recommend a so-called core catcher, which is a molten core ex-vessel cooling facility, to manage a severe accident at a nuclear reactor. Two different types of core catcher concepts are compared to determine their abilities to manage severe accidents and cool core melts. The study reveals that direct cooling is better for cooling capacity and is convenient to construct, while indirect cooing is better for the management of a severe accident.