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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...

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.

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.

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.

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 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.

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.

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.