Search

The vapor formation within the reactor pressure vessel (RPV) is regarded to represent heat removal from core materials to the coolant, while the hydrogen generation within the RPV is regarded to re...

Estimation and understanding of the state of the fuel debris and fission products inside the plant comprise an essential step in the decommissioning of Tokyo Electric Power Company Holdings’ Fukush...

The new R&D programme of JAEA/CLADS tests complements the previous investigations related to BWR severe accidents. A series of tests aim at closing the gaps in understanding of the Fukushima Dai-Ic...

Understanding the final distribution of core materials and their characteristics is important for decommissioning the Fukushima Daiichi Nuclear Power Station (1F). Such characteristics depend on the accident progression in each unit. However, boiling water reactor (BWR) accident progression involves great uncertainty. This uncertainty, which was clarified by MAAP-MELCOR crosswalk, cannot be resolved with existing knowledge and was thus addressed in this work through core material melting and relocation (CMMR) tests. For the test bundle, ZrO2 pellets were installed instead of UO2 pellets. A plasma heating system was used for the tests. In the CMMR-4 test, useful information was obtained on the core state just before slumping. The presence of macroscopic gas permeability of the core approaching ceramic fuel melting was confirmed, and the fuel columns remained standing, suggesting that the collapse of fuel columns, which is likely in the reactor condition, would not allow effective relocation of the hottest fuel away from the bottom of the core. This information will help us comprehend core degradation in boiling water reactors, similar to those in 1F. In addition, useful information on abrasive water suspension jet (AWSJ) cutting for debris-containing boride was obtained in the process of dismantling the test bundle. When the mixing debris that contains oxide, metal, and boride material is cut, AWSJ may be repelled by the boride in the debris, which may cut unexpected parts, thus generating a large amount of waste in cutting the boride part in the targeted debris. This information will help the decommissioning of 1F.

To deepen our understanding for the current debris status and investigate the debris-structure interactions in the pedestal region of Fukushima Unit-3, the Moving Particle Semi-implicit (MPS) method is further developed for simulation of multi-component liquid/solid relocation with solid-liquid phase changes aiming for plant-scale practices. The improvement of the existing MPS method mainly consists of two parts, namely 1) the improvement of numerical stability and accuracy, including a) applying second-order corrective matrix to the particle interaction models and b) particle shifting that can optimize particle configurations; and 2) improvement of calculation efficiency, including a) hybrid OpenMP and MPI parallelization and b) particle type-dependent speed-up algorithms to reduce calculation costs for particles with extremely high viscosity and low velocity. In the current study, the improved MPS method is validated against the experiments carried out at Waseda University, in which molten salt droplets were released to interact with aluminum pillars and solidify on them. Good agreement of the total height of the solidified salt and its distribution on the pillars has been achieved. The successful validation has shown the capability of the current MPS method for simulations of Unit-3 pedestal region.

Interpretation for the two-week long Unit 3 ex-vessel debris cooling behavior was conducted based on 1F plant and site data such as pressure, temperature, gamma ray level and live camera pictures. It was estimated that the debris relocated to the pedestal was in partial contact with liquid water for about initial two days. With the reduction of the sea water injection flowrate, the debris, both in-vessel and ex-vessel ones, became “dry”, in which the debris was only weakly cooled by vapor and this condition lasted for about four days until the increase of the sea water injection. During this dry period, in-vessel and ex-vessel debris were heat up and it took another four days to re-flood the heated up debris.

Boron carbide is used as a neutron-absorbing material in Fukushima-Daiichi Nuclear Power Station (1F), producing borides that are twice as hard as oxides (such as UO2 and ZrO2). The high neutron absorption of boron affects the evaluation of re-criticality during the process of debris retrieval. Therefore, it is important not only to determine the presence of boron but also to investigate the distribution of boron inside the material in a non-destructive manner during decommissioning. To address the uncertainties in the core material relocation behavior of boiling water reactor (BWR) during a severe accident (SA), solidified melt specimens of a simulated fuel assembly were prepared by plasma heating. If core material melting and relocation (CMMR) specimens can be used to estimate the B distribution in 1F Unit-3, that will provide valuable information in the decommissioning of 1F. To address this, the authors focused on the energy-resolved neutron-imaging system, RADEN, which utilizes a wide energy range, from meV to keV. This is an innovative three-dimensional analysis technology for boride distribution that affects the evaluation of hardness and re-criticality. In the calibration standard samples (ZrxB1-x and FexB1-x), there was a good correlation between boron concentration and the energy-dependence of the cross sections of cold and epithermal neutrons. In the CMMR specimens, boron distribution was confirmed from the contrast difference between cold and epithermal neutrons. In the future, the results of calibration standard samples will be applied to the results of CMMR specimens. With this method, three-dimensional boron distribution will be measured, and the understanding of boride distribution 1F Unit-3 will be improved, which may be reflected in an improved SA code.

Water columns were adopted in the pressure measurement system of Fukushima-Daiichi Unit 3 to compensate for evaporation/condensation during normal operation. Some of these water columns evaporated partially during the accident condition jeopardizing correct understanding on actual pressure. Through inter-comparison of reactor pressure vessel (RPV) and suppression chamber (S/C) pressures with drywell (D/W) pressure, such water-column-change effect was evaluated. From this evaluation, correction for the specific effect was developed for RPV and S/C pressure data. With this corrected pressure, slight pressure difference among RPV, S/C, and D/W during the accident transient was evaluated. This information of pressure difference was then integrated with other available data, such as water level, containment atmosphere monitoring system, and environmental dose rate in the Fukushima-Daiichi site, into an interpretation of accident progression behavior focusing on RPV and primary containment vessel pressu...

The authors are developing an experimental technology for simulating severe accident (SA) conditions using simulant fuel material (ZrO2) that would contribute, not only to Fukushima Daiichi (1 F) decommissioning, but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of accident progression behavior. Nontransfer (NTR) type plasma, which has been in practical use with a large torch capacity as high as 2 MW, has the potential to heat subject materials to very high-temperatures without selecting the target to be heated. When simulating 1 F with SA code (Severe Core Damage Analysis Package (SCDAP), Methods for Estimation of Leakages and Consequences of Releases (MELCOR) and Modular Accident Analysis Program (MAAP)), the target of this core-material-melting and relocation (CMMR) experiment was to confirm that NTR plasma has a sufficient heating performance realizing large temperature gradients (>2000 K/m) expected under 1 F conditions. The authors selected NTR-type plasma-heating technology that has the advantage of continuous heating in addition to its high-temperature level. A prototype large-scale experiment (1 m × 0.3 m dia.), called CMMR-0, was conducted in 2016, in which a large temperature gradient was realized and basic characteristics of a heated test assembly were studied. However, the maximum temperature was limited in this test by the instability of the plasma torch under low-oxygen concentrations. It was clarified through this test that an improvement in plasma-heating technology was necessary to heat the large-scale test assembly. The CMMR-1/-2 experiments were carried out in 2017 with a test assembly similar to CMMR-0, applying the improved technology (higher heating power and controlled oxygen concentration). In these two tests, heating history was different, resulting in similar physical responses with more pronounced material melting and relocation in the CMMR-2 experiment. The CMMR-2 experiment was selected from the perspective of establishing an experimental technology. The CMMR-2 experiment adopted a 30-min heating period, wherein the power was increased to a level where a large temperature gradient was expected at the lower part of the core under actual 1 F accident conditions. Most of the control blade and channel box migrated from the original position. After heating, the simulated fuel assembly was measured by X-ray-computed tomography (CT) technology and by electron probe micro-analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective, in terms of applicability of the NTR-type plasma-heating technology to the SA experimental study, was obtained.

The D/W (Drywell) and S/C (Suppression Chamber) pressure data of Fukushima-Daiichi Nuclear Power Plant Unit 3 was analyzed in depth. This analysis provided valuable information related to the accident progression behavior on one hand, and gave a hint for understanding of the debris-to-coolant heat transfer when fuel debris relocated to the pedestal on the other hand. In this unit, the D/W and S/C pressure increased and decreased cyclically with a relationship, which seems to have been dependent on the composition of vapor and non-condensable gases in the S/C cover gas region. Based on this characteristic, the vapor pressure in the S/C cover gas region was evaluated for two pressure decrease cycles during and after the expected debris relocation to the pedestal respectively. This evaluation allowed an understanding that the S/C vapor pressure increased due to the heat transfer from the debris relocated to the pedestal. With this understanding, several possible S/C vapor pressure histories were assumed. Based on this parametric study it was suggested that some part of the S/C water was heated up within about 75 min when the main part of the fuel debris would have been relocating to the pedestal.

To provide supportive information to understand the current debris status in Fukushima Daiichi Nuclear Power Plant Unit-3, sensitivity analyses have been carried out with MELCOR-2.2 for two scenarios, with/without direct leakage from the Reactor Pressure Vessel (RPV) to the Drywell (D/W). The particular focus of the analyses is the estimated debris thermal energy up to and at the time of the major core slumping event, which may be the key to determine the following debris cooling and failure mechanism of the lower head of the RPV. In the analysis, the debris relocation velocity from the core region to the RPV lower head (VFALL), the number of Safety Relief Valves (SRVs) opening, and the amount of core slumping at the major RPV pressure peak were tuned so that the plant data of RPV and PCV pressure from ca. 9:00 to 12:00 March 13th 2011could be reproduced in each scenario. Best-estimate case conditions are summarized for the with/without direct leakage from RPV to D/W scenarios. As a result, as a consideration common to both scenarios, MELCOR analysis tends to estimate significant core oxidation before Automatic Depressurization System (ADS) actuation, and concomitantly estimates little core oxidation from the time after ADS to the time of the major RPV pressure peak. In addition, although a remarkable difference was found in the amount of hydrogen generated in RPV during the major core slumping between the best-estimate with/without leakage cases, limited difference can be observed for the core oxidation heat generation during the major core slumping. This is because the early Zircaloy oxidation prior to ADS actuation in the current MELCOR modelling resulted that the hydrogen generation source during the major core slumping at 12:00 March 13th was primarily from oxidation of stainless steel, which was not as exothermic as that of Zr oxidation.

The onsite investigations of Fukushima Daiichi Nuclear Power Plant (1F) Units-2 and 3 indicate possibilities of multiple breaches of the Reactor Pressure Vessels (RPVs). In the meantime, some analytical works indicate possibilities that the core materials of 1F Units-2 and 3 were once relocated to the lower plenum of the RPVs and cooled (quenched) before the water inventory boiled-off (dry out) and the once-cooled debris re-melted. This study utilizes Lagrangian-based Moving Particle Semi-implicit (MPS) method to investigate such complex solid-liquid multiphase re-melting and the debris-vessel wall interactions to obtain new insight on 1F Units-2 and 3 RPV failure modes to help the future debris retrieval from these reactors. Two major modifications/developments have been carried out based on the previously developed MPS method. Namely, stabilization of rigid-body contact model and the further improvement of the speedup algorithm to enable large and long scale debris-bed re-melting analyses of the real plant scale. The numerical accuracy of the pressure boundary condition at fluid-wall boundary has also been improved. Sensitivity analyses have been carried out with the developed new MPS method. The results indicate the possibility that the lateral part of the RPV may have been subject to not only convective heat transfer of metallic melt pool, but also by conductive heat transfer by oxidic debris conglomerate. The results also indicate that following the initial vessel breach, discharge of metallic melt induces relocation of oxidic debris conglomerates, leading to concentration of the heat source around the central (bottom) part of the lower head. However, large uncertainties associated with 1F are acknowledged and further model validations are necessary before drawing further insights.

Estimation of the final debris distribution at the Fukushima Daiichi Nuclear Power Station (1F) is inevitable for a safe and effective decommissioning. It is necessary to clarify possible failure modes of the reactor pressure vessel (RPV), which is influenced by the thermal status of slumped debris that highly depends on the in-vessel accident progression. In this study, the accident analysis of 1F Unit 2 (1F2) was conducted using the RELAP/SCDAPSIM code, in order to understand better the in-vessel accident progression. One of the unsolved issues of 1F2 is the mechanism of three pressure peaks measured through late Mar. 14 to early Mar. 15, 2011. Despite various analyses, its mechanism is not clearly understood. Comparing the results of previous boiling water reactor (BWR) core degradation experiments and that of 1F2 numerical analysis, it can be estimated that most relocated metallic materials had solidified at the core bottom at the onset of first pressure peak. It is likely that the pressure increase occurred due to the evaporation of injected water reaching the heated core plate structures. Between the first and second pressure peaks, the water is assumed to have been injected continuously and the water level was likely to have recovered to BAF at the initiation of the second pressure peak. Probable slumping of a certain amount of molten materials initiated the second pressure peak and the subsequent gradual pressure increase continued possibly due to massive reaction between coolant and remaining Zircaloy in the core. Assuming the closure of the safety relief valve (SRV) at 0:00 on Mar. 15, the third pressure peak was well reproduced in the analysis. Although the total amount of the slumped material was evaluated, large deviation exists among the cases and uncertainty is still large.

Peak shape analysis was performed for the energy spectra of Doppler-broadened prompt γ -rays generated by neutron capture reactions with various boride or boron samples. Significant differences were observed between nonmetallic and metallic borides. Minor differences between the peak shapes of prompt γ -rays from zirconium- and ferro-borons were evaluated by a peak fitting method. The identification performance for zirconium- and ferro-borons and other types of borides was measured.

The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th, 2011 put the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions and core meltdown due to station blackout. Although research efforts have been made by various parties to study the accident scenarios since the Fukushima accident, there remain unresolved issues regarding the core degradation behavior suggested by measurement data such as water level, Reactor Pressure Vessel (RPV) pressure and Primary Containment Vessel (PCV) pressure. To analyze and resolve such issues would be helpful to promote further understanding of the severe accident scenario at Fukushima units as well as the decommissioning work undergoing. The current study focuses on a detailed analysis of the RPV pressure peak event that occurred in Unit-3 at 12:00 on March 13th 2011. Sensitivity analysis cases were carried out with MELCOR 2.2 code i with sensitivity parameters that can influence the RPV pressure behavior, such as the debris quenching heat transfer coefficient, the number of opening SRVs during the RPV pressure peak event, amount of core slumping and particulate debris diameter. The cases that could reproduce the RPV pressure peak were further discussed to show likely debris bed energy history and the water mass history in the lower plenum during the RPV pressure peak event. The current study suggests that 1) Opening of SRVs equivalent to the total area of 4–6 fully-open SRVs (or equivalent leak area) could have occurred during the pressurization phase of the RPV accompanied by heavy debris quenching effect, while the opening of SRVs equivalent to a total area of at least 2 fully-open SRVs (or equivalent leak area) could have occurred during the depressurization phase of the RPV accompanied by moderate debris quenching effect. 2) The particulate debris diameter is not a very sensitive parameter when evaluating the debris quenching effect of Unit-3 in the current MELCOR modeling. 3) The current modeling suggests that around 70–110 GJ of energy can be removed by coolant during the debris quenching period with 30 tons of water reduction from the lower plenum.

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. When the molten corium, which is generally comprised of UO2, ZrO2 and metals such as zircalloy and stainless steel, is discharged into a dry reactor cavity, a stratified molten pool configuration with two immiscible oxidic and metallic phases can be expected to form and lead to MCCI. Compared to a homogenous oxidic molten pool configuration, the metallic phase in the stratified molten pool might influence the crust formation on the corium-concrete interface and consequently cause different concrete ablation behavior to evaluate MCCI progression concerning containment failure. In terms of this issue, past experimental studies, such as COMET-L, VULCANO VBS and MOCKA test series, have been carried out to investigate the influence of such oxidic and metallic stratified pool configuration on MCCI. The experimental results have shown that the metallic phase can have a significant impact on the axial and radial ablation kinetics that could influence the ablation patterns of reactor pit. As regards numerical studies, past numerical modeling of MCCI was generally based on Eulerian methods and simplified empirical approach to simulate solid/liquid phase change and evolving of corium/crust/concrete interface. Such modeling might be efficient but have shown deficiencies and inadequacies due to its Eulerian and empirical nature, which has suggested a necessity to seek for a more mechanistic approach for modeling of MCCI. In this sense, Moving Particle Semi-implicit (MPS) method is considered suitable for MCCI analysis for its advantages of tracking interfaces and modeling phase change accurately as a Lagrangian particle method. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. Solid/liquid phase change was simulated with types of solid and liquid particles with thermal and physical properties including temperature and solid fraction, which enabled tracking of the solid/liquid status of each particle to achieve accurate free surface and corium/crust/concrete interface capturing. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

Authors are developing an experimental technology to realize experiments simulating Severe Accident (SA) conditions using simulant fuel material (ZrO2 with slight addition of MgO for stabilization) that would contribute not only to Fukushima Daiichi (1F) decommissioning but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of the accident progression behavior. Based on the results of the prototype test, improvement of plasma heating technology was conducted. The Core Material Melting and Relocation (CMMR)-1/-2 experiments were carried out in 2017 with the large-scale simulated fuel assembly (1 m × 0.3 mϕ) applying the improved technology (higher heating power and controlled oxygen concentration). In these two tests, heating history was different resulting basically in similar physical responses with more pronounced material melting and relocation in the CMMR-2 experiment. The CMMR-2 experiment is selected here from the viewpoint of establishing an experimental technology. The CMMR-2 experiment adopted 30-min heating period, the power was increased up to a level so that a large temperature gradient (> 2,000 K/m) expected at the lower part of the core in the actual 1F accident conditions. Most of the control blade and the channel box migrated from the original position. After the heating, the simulated fuel assembly was measured by the X-ray Computed Tomography (CT) technology and by Electron Probe Micro Analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective in terms of applicability of the non-transfer type plasma heating technology to the SA experimental study was obtained.

A new experimental program using nontransfer (NTR) type plasma heating is under consideration in Japan Atomic Energy Agency (JAEA) to clarify the uncertainty on core-material relocation (CMR) behavior of boiling water reactor (BWR). In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm × 107 mm × 222 mm (height)). An excellent perspective in terms of applicability of the NTR plasma heating to melting high melting-temperature materials such as ZrO2 has been obtained. In addition, molten pool was formed at the middle height of the test piece indicating its capability to simulate the initial phase of core degradation behavior consistent with the real UO2 fuel PHEBUS fission products (FP) tests. Furthermore, application of electron probe micro-analyzer (EPMA), scanning electron microscope (SEM)/energy dispersive X-ray spectrometry (EDX), and X-ray computed tomography (CT) led to a conclusion that the pool formed consisted mainly of Zr with some concentration of oxygen which tended to be enhanced at the upper surface region of the pool. Based on these results, an excellent perspective in terms of applicability of the NTR plasma heating technology to the severe accident (SA) experimental study was obtained.

In order to eliminate the energetic potential in the case of postulated core-disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner-duct structure (FAIDUS) has been considered. Recently, a design option of FAIDUS which leads molten fuel to upward discharge has been considered as the reference core design of the Japan Sodium-Cooled Fast Reactor (JSFR). In this study, a series of experiments which consisted of three out-of-pile tests and one in-pile test were conducted to obtain experimental knowledge of the upward discharge of molten fuel. Experimental data which showed a sequence of upward fuel discharge and effects of initial pressure conditions on upward discharge were obtained through the out-of-pile and in-pile test. Preliminary extrapolation of the present results to the supposed condition in the early phase of the CDA in the JSFR design suggests that the sufficient upward flow rate of molten fuel is expected to prevent the core melting from progr...

In core-disruptive accidents of sodium-cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels with large hydraulic diameters, such as the control-rod guide tube and a concept of the Fuel Assembly with Inner Duct Structure have a potential to provide effective fuel-discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments were conducted to investigate fuel-discharge behaviour through the sodium-filled channels. In the first series of experiments, an alloy with low melting temperature was ejected into a water channel to clarify dominant phenomena for melt discharge through the coolant-filled channel and to develop methodologies for evaluating the effects of coolant on melt discharge. In the second series of experiments, a molten alumina was discharged through the sodium-filled channel in order to verify the applicability of the knowledge and evaluation methodologies ...

A probabilistic safety assessment (Level-2 PSA) methodology was developed for comprehensive risk evaluation of sodium-cooled fast reactors. As part of this development, in this paper, phenomenological event trees were developed as well as technical database to quantify the probability of event sequences in the Level-2 PSA, focusing on the initiating and transition phases of unprotected events. Typical and important accident categories were selected: unprotected loss of flow (ULOF), unprotected transient overpower (UTOP) and unprotected loss of heat sink (ULOHS). Based on the state-of-the-art knowledge, the headings of these event trees were selected so that dominant factors in accident consequences can be represented appropriately. For each of the headings, available information for the probability quantification were reviewed and integrated as the technical database for the Level-2 PSA. It was clarified that the headings of the ULOF category, for which experimental database and evaluation models have been reasonably established, can be commonly applied to certain part of the different accident categories except for some specific points, which were identified in this study. For the ULOHS category, an additional event tree is necessary before the core disruption providing various boundary conditions for the initiating phase. For the transition phase, dominant factors were also identified through parametric analyses. In the Japan sodium-cooled fast reactor, an inner duct is introduced into a fuel subassembly for enhancing molten fuel discharge from disrupted core in the transition phase. The parametric study showed that the analytical case without the fuel discharge through the inner duct resulted in an occurrence of recriticality regardless of the fuel discharge through control-rod guide tubes. This suggests that the fuel discharge through the inner duct is essential to avoid severe recriticality in the transition phase.

In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropri...

In the CABRI-FAST and CABRI-RAFT programs within a collaboration with the Institut de Radioprotection et de Surete Nucleaire (IRSN) and Forschungszentrum Karlsruhe (FZK), five pulse-type transient overpower tests were performed in order to study fuel pin behavior and failure condition in the Unprotected Loss-of-Flow (ULOF) accident. In these tests, two types of low-smear-density fuels irradiated in the French Phenix reactor at different burn-up levels were used so that an experimental database extension from the former CABRI-1 and CABRI-2 programs can be obtained. Pin failure took place in three of these tests giving information on the failure threshold. In two tests, no pin failure took place and useful information related to the transient fuel behavior up to failure and failure mechanism was obtained. These test results were interpreted through detailed analysis of experimental data and PAPAS-2S code calculations. In these calculations, pretransient fuel characteristics obtained from the sibling fuels w...

In the TPA2 test of the CABRI-RAFT program, which is part of a fast reactor safety study, fuel-to-steel heat transfer characteristics within a molten fuel/steel mixture system have been investigated. This test was performed in the French CABRI reactor and used a test capsule that contained fresh 12.3%-enriched UO2 pellets with embedded stainless steel balls. Following a preheating phase, the capsule was subjected to a transient overpower that resulted in fuel melting and steel vaporization. The observed steel vapor pressure buildup was quite low, which suggested the presence of a mechanism that significantly reduced the fuel-to-steel heat transfer. A detailed experimental data evaluation by SIMMER-III led to one possible interpretation that the steel vaporization at the surface of the steel ball blanketed the steel from the molten fuel.

For the transition phase analysis of core disruptive accidents, the development of a three-dimensional reactor safety analysis code, SIMMER-IV, has been carried out based on the technology of the two-dimensional SIMMER-III code. The world first application of SIMMER-IV to a small-sized sodium-cooled fast reactor has also been attempted to clarify event progression in the early stage of the transition phase. This SIMMER-IV calculation is compared to the two-dimensional case calculated by SIMMER-III, neglecting the presence of control rod guide tubes. The present analysis with the three-dimensional representation suggests that the conventional scenario leading to rather early high-mobility fuel pool formation is unrealistic and the degraded core tends to keep low mobility in the early stage of transition phase.

Experimental verification of a reactor safety analysis code, SIMMER-III, was undertaken for transient behaviors of large-scale bubbles with condensation. The present study aimed to verify the code for numerical simulations of relatively short-time-scale multi-phase, multi-component hydraulic problems. Among these, vaporization and condensation, or simultaneous heat and mass transfer, play important roles. In this study, a series of transient bubble behavior experiments dedicated to condensation phenomena with noncondensable gases was carried out. In the experiments, a pressurized mixture of noncondensable gas and steam was discharged as a large-scale single bubble into a cylindrical pool filled with stagnant subcooled water. The concentration of noncondensable gas was taken as an experimental parameter as was the species of noncondensable gas. The characteristics of transient behavior of large-scale bubbles with condensation observed in the experiments were estimated through experimental analyses using SIMMER-III. In the experiments with steam condensation, dispersion of the gas mixture discharged into the liquid pool was accompanied by vapor condensation at the bubble surface. SIMMER-III simulations suggested that the noncondensable gas had a less inhibiting effect on the condensation of large-scale bubbles. This is a different characteristic to that of the quasi-steady condensation of small-scale bubbles observed in our previous experiments.

The possibility of severe recriticality could be excluded if the molten core materials are discharged from reactor core in the early stage of core disruptive accident (CDA). Based on this idea, several design measures for future commercial liquid metal-cooled fast breeder reactors (LMFBRs) have been proposed to enhance the molten fuel discharge from core in order to prevent formation of the core-wide molten pool with high mobility. One promising concept in these design candidates is modified-FAIDUS (Fuel subassembly with Inner DUct Structure). The event progression in unprotected loss of flow (ULOF) accident in a sodium-cooled large scale FBR with modified-FAIDUS was analyzed to assess the effectual performance of modified-FAIDUS in preventing severe recriticality using the SAS4A and SIMMER-III codes. Two parametric cases were performed covering the uncertainty of duct wall failure mechanism, one with stable fuel crust and another with unstable crust condition. The calculation showed that the final amount of discharged fuel from core in both cases was more than 20% of initial core inventory. The degraded core after fuel discharge is composed of the mixture of solidified fuel, swollen fuel chunks and molten steel, of which low mobility prevents massive fuel motion. The reactor power lowered to decay heat level and the reactivity lowered around −20 $, thus, the possibility of severe recriticality was eliminated.

The WF (wall failure) test of the EAGLE program, in which ∼2 kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR (Impulse Graphite Reactor) of NNC/Kazakhstan. In this test, a 3 mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10 mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 s after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events. A preliminary interpretation on the WF test results is presented in this paper.

In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear densitymore » and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long transient timescale, such low smear density fuel has a potential to allow gas escape to plenum leading to a very effective mitigation of swelling-induced PCMI.In case of very high cladding temperature near its melting point, plenum-gas blowout at cladding rupture takes place before fuel disintegration. Fuel-disintegration behavior under this condition is dominated by fuel enthalpy, and no special effect of the high burnup can be identified through comparison with the CABRI-1 test results.« less

SAS-SFR (derived from SAS4A) is presently the most advanced computer code for simulation of the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors (SFR). In the past two decades, intensive model improvement works have been conducted for SAS-SFR utilizing the experimental data from the CABRI programs. The main target of the present work is to confirm validity of these improved models through a systematic and comprehensive set of test analyses to demonstrate that the improved models has a sufficient quality assurance level for applications to reactor conditions. In order to reach these objectives, an approach of PIRT (Phenomena Identification and Ranking Table) on a set of accident scenarios has been applied. Based on the fact that there have been a significant amount of validation studies for decades, development of the code validation matrix concentrated on key issues. Different accident scenarios have been chosen for the PIRT considering typical SFR accident transients that address a large range of phenomena. As the most important and typical Core Disruptive Accident scenarios leading to generalized core melting and to be addressed with SAS-SFR in the present study, ULOF (Unprotected Loss Of Flow), UTOP (Unprotected Transient OverPower) and ULOHS (Unprotected Loss Of Heat Sink) are selected. The PIRT process applied to a given accident scenario consists in an identification of the phenomena involved during the accident, the evaluation of the importance of the phenomena regarding to the evolution and consequences, and the evaluation of the status of knowledge based on the review of available experimental results. The identified phenomena involved in ULOF are explained as follows for the primary phase. Starting from initiating events, a loss of grid power leading to flow coast down without scram is assumed. The scenario up to coolant boiling is the main point within the first part of the ULOF phenomenological chart. Those elements related to reactivity feedback, such as heat up of coolant, fuel and various structures and their deformation due to the thermal transient are picked up. Depending on the time scale before boiling starts, primary, secondary and tertiary loop heat transfer including the DHR (Decay Heat Removal) system response is concerned since it defines the core inlet coolant temperature. Core inlet coolant temperature gives direct impact on the thermal condition of the core. It also affects reactivity through thermal expansion of the grid plate. In the second part of the ULOF phenomenological chart, elements such as coolant boiling, mechanical response of the fuel pin leading to cladding failure, FCI (Fuel-Coolant Interaction) and post-failure material relocation are picked up. This part of the chart is basically common to the ULOHS. Respective identified phenomena are to be simulated in the SAS-SFR code. To validate the function of the models in the code, ten high priority CABRI experiments are selected. Validation studies on these tests are underway. With the present study, important phenomena involved in ULOF, UTOP and ULOHS were identified and an evaluation matrix for the selected CABRI experiments was developed.

Slow ramp-type transient-overpower tests were performed within the framework of the international CABRI-2 experimental program. The implemented power transients of ~1% nominal power/s correspond to...

Two three-pin cluster tests simulating the Unprotected Loss-of-Flow (ULOF) accident of Sodiumcooled Fast Reactors (SFRs) were conducted focusing on postfailure fuel relocation and freezing behavior. These tests supplied complementary information to the existing CABRI tests with a single-pin geometry. Based on detailed data evaluation and theoretical interpretation for the three-pin cluster tests, it is concluded that axial fuel relocation and freezing are dominated by local fuel enthalpy, and the relation between penetration length and local fuel enthalpy observed in these CABRI tests is basically applicable to the large-bundle condition. It is also clarified that a fuel/steel mixture tends to create tight blockages near the axial ends of the relocating fuel. Part of the fission gas released from the heating-up and melting fuel is expected to be trapped within the bottled-up region between the upper and lower blockages and will keep this region pressurized for a relatively long period. © 2011 Atomic Energy Society of Japan. All Rights Reserved.

In this paper an improved method to evaluate key phenomena in the initiating-phase energetics of unprotected loss-of-flow (ULOF) whole-core accidents in liquid-metal fast breeder reactors is presented. Three phenomena, namely, axial fuel expansion, fuel failure, and postfailure fuel motion, have been examined through the CABRI-1 in-pile experiments and analyses with special emphasis on the self-limiting mechanisms of the energetics potential. For the experiment analyses, the SAD3D, PAPAS-2S, and SAD4A computer codes are employed selectively to obtain a detailed investigation of the phenomena and to validate physical models. The improved knowledge obtained through the research efforts in CABRI-1 and relevant safety experiments has been implemented in the revised SAS3D code. This evaluation method, which accounts for the self-limiting mechanisms, has been applied to a reactor analysis of an energetic ULOF sequence. The results of the application study confirm the importance and effectiveness of the method.

In the CABRI-FAST experimental program, four in-pile tests were performed with slow-power-ramptype transient-overpower conditions (called hereafter as “slow TOP”) to study transient fuel pin behavior under inadvertent control-rod-withdrawal-type events in liquid-metal-cooled fast breeder reactors. The slow TOP test with a preirradiated solid-pellet fuel pin under a power ramp rate of approximately 3%Po/s was realized as a comparatory test against an existing test in the CABRI-2 program where approximately 1%Po/s was adopted with the same type of fuel pin. In spite of the different power ramp rates, the evaluated fuel thermal conditions at the observed failure time are quite similar. Three slow TOP tests with the preirradiated annular fuel resulted in no pin failure showing a high failure threshold. Based on posttest examination data and a theoretical evaluation, it was concluded that intrapin free spaces, such as central hole, macroscopic cracks, and fuel-cladding gap, effectively mitigated the fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in the case of a very large amount of fuel melting. These CABRI-FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database under various fuel and transient conditions. © 2009 Taylor and Francis Group, LLC.

High heat flux tests of RF window disks have been demonstrated. The results show that the modified ceramic disk, which has been designed based on stress-thermal analysis, can withstand heat deposition much higher than that of the reference disk. The results confirmed the validity of design of the modified disk for use in high-power klystron windows.