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Investigation of the primary containment vessel inside the Fukushima Daiichi Nuclear Power Station showed that a significant amount of the molten corium reached the bottom of the pedestal region. The molten corium and concrete likely caused a complex interaction called Molten Corium Concrete Interaction. The solidification hysteresis of these ex-vessel debris significantly influences its properties. We performed a thermodynamic analysis using the TAF-ID database to infer the solidification path of U-Zr-Al-Ca-Si-O molten corium, which was chosen for a prototypic system of ex-vessel debris. The solidification path for the CaO-rich sim-corium showed that (i) melting as a single liquid phase above 2430 K, (ii) selective solidification of the oxide-rich corium mainly consisted of fuel materials, and (iii) solidification of the remaining materials as a silicate matrix. In contrast, the solidification path for the SiO$_{2}$-rich corium indicated that (i) formation of liquid miscibility gap above 2200 K between U-rich and Zr-rich oxidic melts, (ii) individual precipitation of solid phases in each liquid phase.

DOI： 10.1016/j.calphad.2022.102481

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Immediately after the 1F-accident, various attempts have been made to evaluate the fuel debris characteristics toward the decommissioning of 1F. The present review outlines those attempts. In the years immediately following the 1F-accident, the knowledge obtained from the 1F-site (especially from the damaged reactors of Units 1, 2 and 3) was extremely limited. The approximate location of fuel debris was investigated by muon tomography, and its characteristics were roughly estimated based on the past findings such as the results of the Three Mile Island-II accident investigation in the United States, which gave us information of prototypical accident scenarios and debris characteristics for pressurized water reactor accident. After that, various internal investigation robots were developed, and from 2017, investigation of the inside of the reactor containment vessel was started using these robots. Consequently, these three units were found to have core damage status and debris distribution that were rather different from what had been expected based on the typical accident scenario of a pressurized water reactor. In parallel, a small amount of U-bearing particle was recovered from the smear samples of these robots. The analysis of these particles is ongoing to get information relevant to fuel debrsi body. Furthermore, international collaboration is ongoing mainly under OECD/NEA, including accident analysis and debris characterization. From now on, one need to further understand 1F-accident scenario and progress debris characterization based on these 1F-site information.

DOI： 10.1080/00223131.2022.2040393

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DOI： 10.1080/00223131.2021.1985647

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The results of the several recent tests performed in JAEA/CLADS are outlined in this paper. However, particular point of this work is focused on the interesting effect that was found on the debris, that contained partially reacted B$_{4}$C (control blade debris). A Raman investigation of the control blade metallic debris helped to refine the governing mechanism of the B-compounds formation and transport, which is probably specific mostly for BWRs due to unique bundle configuration and materials morphology. All these factors may directly influence the accident progression in BWR and influence the final debris properties.

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DOI： 10.1080/00223131.2021.1906777

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Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000$^{\circ}$C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600$^{\circ}$C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

DOI： 10.3390/thermo1020011

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Language：English   Publishing type：Research paper (international conference proceedings)  

DOI： 10.1115/ICONE28-65129

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DOI： 10.1080/00223131.2020.1856734

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Current status and examples of the application for TAF-ID (under developing in OECD/NEA TAF-ID project) database are outlined.

DOI： 10.1016/j.calphad.2020.102212

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DOI： 10.1299/mej.19-00503

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For a criticality assessment of the fuel debris generated by the Fukushima Daiichi Nuclear Power Plant accident, knowing the segregation of neutron absorber materials, {\it i.e.}, Gd, B, and Fe, in the fuel debris is necessary. Although B may mostly evaporate during melting, Fe and Gd are expected to remain in the molten corium. To understand the redistribution behavior of Gd and Fe during the solidification of the molten corium, solidification experiments with simulated corium (containing UO$_{2}$, ZrO$_{2}$, FeO, and Gd$_{2}$O$_{3}$ with a small amount of simulated fission products such as MoO$_{3}$, Nd$_{2}$O$_{3}$, SrO, and RuO$_{2}$) were performed using a cold crucible induction heating method. The simulated corium was slowly cooled from 2,500$^{\circ}$C and solidified from the bottom to the top of the melt. An elemental analysis analysis of the solidified material showed that the Fe concentration in the inner region increased up to approximately 3.4 times that in the bottom region. This suggested that FeO might be concentrated in the residual melt and that, consequently, the concentration of Fe increased in the later solidification region. On the contrary, the Gd concentration in the periphery region was found to be approximately 2.0 times higher than that in the inner region, suggesting the segregation of Gd in the early solidified phase. No significant segregation was observed for the simulated fission products.

DOI： 10.1016/j.jnucmat.2020.152093

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DOI： 10.1080/00223131.2019.1691070

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DOI： 10.1299/mej.19-00477

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As expected for accident tolerant fuels, investigation of steam oxidation for silicon carbide under the conditions beyond design basis accident scenarios is needed. Many studies focused on steam oxidation of SiC at temperatures up 1600$^{\circ}$C have been conducted and reported in the literature. However, behavior of SiC in steam at temperatures above 1600$^{\circ}$C still remains unclear. To complete this task, we have designed and manufactured a laser heating facility for steam oxidation at extreme temperatures. With the facility, we report the first results on the steam oxidation behavior of SiC at temperatures range of 1400-1800$^{\circ}$C for short term exposure of 1-7 h under atmospheric pressure. Based on the mass change of SiC samples, parabolic oxidation rate and linear volatilization rate were calculated. The oxidation layer appears to be maintained at 1800$^{\circ}$C in steam, but the bubble formation phenomenon suggests other volatilization reactions may limit its life.

DOI： 10.1016/j.jnucmat.2019.151939

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The importance of fuel chemistry has been revivaled since Fukushima-Daiichi Nuclear Power Station (FDNPS) accident. The inspection and analysis of damaged three units, which had been operated in March 11, 2011, showed large differences in the accident progression sequence for these units, because operators attempted to prevent or mitigate the accident progression of each unit by all means possible. Characteristics of fuel debris are considered to be largely influenced by the difference in the sequence and, hence, deviated from those predicted from prototypic accident scenarios, which had been mainly identified from the analysis of Three Mile Island-2 (TMI-2) accident and the following sim-tests. For the proper improvement of our knowledge on severe accident (SA), including non-prototypic conditions, one has to start improving the phenomenology of fuel/core degradation and fission product (FP) behavior. Advances in the chemistry is the most essential approach. The present review attempts to focus on the recent updates and remaining concerns after the FDNPS accident.

DOI： 10.1016/B978-0-08-102571-0.00015-X

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Silicon carbide (SiC) has recently attracted much attention as a potential material for accident tolerant fuel cladding. To investigate the performance of SiC in severe accident conditions, study of steam oxidation at high temperatures is necessary. However, the study focusing on steam oxidation of SiC at temperatures above 1600$^{\circ}$C is still certainly limited due to lack of test facilities. With the extreme oxidation/corrosion environment in steam at high temperatures, current refractory materials such as alumina and zirconia would not survive during the tests. Application of laser heating technique could be a great solution for this problem. Using laser heating technique, we can localize the heat and focus them on the test sample only. In this study, we developed a laser heating facility to investigate high-temperature oxidation of SiC in steam at temperature range of 1400-1800$^{\circ}$C for 1-7 h. The oxidation kinetics is then being studied based on the weight gain and observation on cross-sectioned surface of tested sample using field emission scanning electron microscope. Off-gas measurement of hydrogen (H$_{2}$) and carbon monoxide (CO) generated during the test is also being conducted via a sensor gas chromatography. Current results showed that the SiC sample experienced a mass loss process which obeyed paralinear laws. Parabolic oxidation rate constant and linear volatilization rate constant of the process were calculated from the mass change of the samples. The apparent activation energy of the parabolic oxidation process was calculated to be 85 kJ.mol$^{-1}$. The data of the study also indicated that the mass change of SiC under the investigated conditions reached to its steady stage where hydrogen generation became stable. Above 1800$^{\circ}$C, a unique bubble formation on sample surface was recorded.

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After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R\&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R\&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R\&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.

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Cooling rate of molten core materials during solidification significantly affects the segregation of major constituents of fuel debris. To understand general tendency of the segregation, liquefaction/solidification tests of simulated corium (UO$_{2}$, ZrO$_{2}$, FeO, B$_{4}$C and sim-FP oxides) were performed. Simulated corium was heated up to 2600$^{\circ}$C under Ar atmosphere and then cooled down with two different cooling processes; furnace cooling (average cooling rate is approximately 744$^{\circ}$C/min) and slow cooling (cooling rate in 2600$^{\circ}$C$\sim$2300$^{\circ}$C is 5$^{\circ}$C/min and in 2300$^{\circ}$C$\sim$1120$^{\circ}$C is approximately 788$^{\circ}$C/min). Element analysis detected three oxide phases with different composition and one metal phase in both solidified samples. Solubility of FeO in these oxide phases was mostly fixed to be 12$\pm$5at\% in both samples, which is in reasonable accordance with the value estimated from UO$_{2}$-ZrO$_{2}$-FeO phase diagrams. However, a significant grain-growth of one oxide phase, rich in Zr-oxide, was detected only in the slowly cooled sample. The composition of this particular oxide phase is comparable to the initial average composition. The condensation is considered to be caused by the connection of remaining liquid agglomerates during slow solidification.

DOI： 10.3327/taesj.J18.029

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Language：English   Publishing type：Research paper (scientific journal)  

DOI： 10.1080/00223131.2019.1592724

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This paper presents recent results on high-temperature control blade degradation at the very beginning phase of a severe accident in BWRs. The large-scale experiment has been performed in JAEA-CLADS laboratory using Large-scale Equipment for Investigation of Severe Accidents in Nuclear reactors (LEISAN). A prototypic sequence of Fukushima Daiichi (1F) Unit 2 has been taken. It has been shown that due to specific conditions happened at Unit 2 the lack of available oxygen allowed metallic melt more flexibility for relocation and inhomogeneous redistribution of melt components due to specially recreated temperature gradient. Phase composition of remained B$_{4}$C control blade claddings at different elevations, and phase composition of melt has been investigated by complementary methods and have shown significant difference in elevations together with stratification of metallic components with origination of high melting point layers due to redistribution of steel components and involvement of B and C. It allowed absorber blade residuals with B$_{4}$C inside being severely damaged by melting to survive at 1475$^{\circ}$C and protect B$_{4}$C from direct contact with steam.

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In order to obtain a precise understanding of molten material behavior inside RPV and to improve the accuracy of the SA code, a new computational fluid dynamics (CFD) code with multi-phase, multi-physics models, which is called JUPITER, was developed. It optimized the algorithms of the multi-phase calculation. Besides, the chemical reactions are also modeled carefully in the code so that the melting process could be treated precisely. A series of verification and validation studies are conducted, which show good agreement with analytical solutions and previous experiments. The capabilities of the multi-physics models in JUPITER code provide us another useful tool to investigate the molten material behaviors in the relevant severe accident scenario.

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It is predictively evaluated that degradation of fuel assembly proceeded in a certain steam-starved condition at the early stage of a SA at 1F unit 2 (BWR). As for PWR fuel assembly, effective steam flow rate was properly indicated by normalizing to a unit of g-H$_{2}$O/sec/rod which is used as an important parameter for evaluating fuel degradation progression. Due to the inhomogeneous configuration of BWR fuel assembly, the difference of Zry oxidation and hydrogen uptake between the inside and outside of the channel box cannot be properly evaluated by this normalization. Instead of g-H$_{2}$O/sec/rod, proper evaluation unit for BWR configuration is necessary. To accumulate Zry oxidation and hydrogen uptake data for steam-starved conditions, high temperature oxidation tests were performed using a simulated BWR fuel bundle sample. The use of equivalent diameter of the cross section of BWR fuel assembly was proposed for normalization of effective steam flow rate.

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High temperature interaction of chemical vapor deposition SiC with steam was investigated at 1700-1800$^{\circ}$C for 0.1-3 h in a mixture of steam and argon gas containing 98\% of steam at 1 atm. At the investigated conditions, although a dense oxide layer was observed on sample surface, significant mass loss of SiC occurred. Below 1700$^{\circ}$C, the oxidation kinetics seems to follow the para-linear laws. The apparent activation calculated based on the data of this study is to be 370 kJ/mol. Rapid degradation and bubbling of SiC at 1800$^{\circ}$C were observed after 1 h oxidation. This suggested that chemical interaction behaviours above 1700$^{\circ}$C might be changed due to the liquefaction of silica.

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Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B$_{4}$C) control rods potentially generate far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B$_{4}$C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B$_{4}$C starting at temperatures around 1250$^{\circ}$C, well below the significant interaction temperatures of UO$_{2}$-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

DOI： 10.1016/j.jnucmat.2017.12.004

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In Japan, the research and development (R\&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R\&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R\&D project are reported.

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Language：English   Publishing type：Research paper (scientific journal)  

DOI： 10.1299/mej.16-00567

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Scopus

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At the fuel assembly degradation stage in severe accidents, chemical features of the intermediate products are expected to be changed depending upon the accident progressions. These differences are originated from the differences in oxygen potential and temperature, and are highly important for evaluating the relocation and stratification progress of the fuel debris. Two types of sim-test with the different oxygen potentials were performed to investigate these tendencies. The chemical features of the intermediate materials used in the tests were determined from the observations for the control blade and channel box degradation in our previous study. The present results indicate that the U concentration in the metallic layer is largely varied depending upon the oxygen potential of the atmosphere. Also, when the B$_{4}$C-Fe alloy, as of a typical intermediate product, coexists with UO$_{2}$ and Zr, the apparent red-ox reaction rate between UO$_{2}$ and Zr are mitigated.

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To understand the degradation process of control blade channel, control blade degradation tests using sim-materials were performed under various atmospheres with the temperature increase and temperature gradient conditions. In the steam atmosphere with a flow rate of 0.0125 g/s/rod, control blade, channel box, and fuel rods were degraded, especially at the upper part of the test piece, which was similar to that observed in argon atmosphere test. However, the observed degradation was rather different at a flow rate of 0.0417 g/s/rod. At the upper part of the test piece, only the control blade was degraded preferentially and did not react with channel box wall. In contrast, the eutectic reaction of S.S./B$_{4}$C-melt and Zry occurred at the lower part. These observations suggested the existence of a threshold condition for the control rod degradation between 0.0125 and 0.0417 g/s/rod which is significantly affected by the thickness of the oxide layer on Zry.

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In order to precisely investigate molten core relocation behavior in the Fukushima Daiichi Nuclear Power Station, we have developed the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior including solidification and relocation based on the three-dimensional multiphase thermal-hydraulic simulation models. At the moment, multicomponent analysis method which can be treated any number of component as a fluid or solid body, Zr-water reaction model and simple radiation heat transfer model were implemented and showed that multicomponent melt flow and its solidification were confirmed in the simplified core structure system. However, the validation of the JUPITER using high temperature molten material has not been performed yet. In this paper, in order to evaluate the validity of the JUPITER, especially, for high temperature melt relocation experiment, we compared between numerical and experimental results for that system. As a result, qualitatively reasonable result was obtained. And also we performed melt relocation simulation on actual core structures designed by three dimensional CAD (Computer-Aided Design) and then we estimated phenomena which might be actually occurred in SAs.

DOI： 10.1115/ICONE24-60453

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For understanding the control blade degradation mechanism of BWR, the thermodynamic database for the fuel assembly materials is a useful tool. Although iron, boron, and carbon ternary system is a dominant phase diagram, phase relation data is not sufficient for the region in which the boron and carbon compositions are richer than the eutectic composition. The phase relations of three samples were analyzed by X-ray diffraction, scanning electron microscope and energy dispersed X-ray spectrometry. The results indicate that Fe$_{3}$(B,C) phase only exists in the intermediate region at 1273 K and that the solidus temperature widely maintains at about 1400 K for all three samples, which are different from the calculated data using previous thermodynamic database. The difference might be originated from the over-estimations of the interaction parameter between boron and carbon in Fe$_{3}$(B,C).

DOI： 10.1080/00223131.2015.1016465

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In order to explore the applicability of the chlorination by MoCl$_{5}$ as a potential pretreatment technique for waste treatment of fuel debris by pyrochemical methods, chlorination experiments of UO$_{2}$ and (U$_{0.5}$Zr$_{0.5}$)O$_{2}$ simulated fuel debris were carried out in two steps: the first one is a chlorination reaction by homogeneous heating, the second one is a volatilization of molybdenum by-product by heating under temperature gradient condition. Most of UO$_{2}$ and (U$_{0.5}$Zr$_{0.5}$)O$_{2}$ powder were converted to UCl$_{4}$ or UCl$_{4}$ and ZrCl$_{4}$ mixture at 573 K, respectively. In the case of (U$_{0.5}$Zr$_{0.5}$)O$_{2}$sintered particle, most of sample was converted to the chlorides because the products evaporated and be separated from sample surface at 773 K, while only the surface of the sample disk was converted to the chlorides at 573 and 673 K. Most of molybdenum by-product and ZrCl$_{4}$ were separated from UCl$_{4}$ by volatilization at 573 K.

DOI： 10.1080/00223131.2015.1070110

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Chemical interactions between S.S. and B$_{4}$C may lead to the melting of control rods and subsequent relocation of control rod materials in the early stage of severe accident. The liquid products interact with the surroundings such as control rod blades, Zircaloy channel boxes and fuel rods, which may accelerate the fuel bundle degradation. Previous studies on the chemical interactions between S.S. and B$_{4}$C have been performed using the B$_{4}$C powders or pellets. However, almost all of control rods of BWR in Japan consist of granular B$_{4}$C filled in S.S. tubes. In the present study, the chemical reaction tests were carried out using the materials adopted in BWR in Japan, namely granular B$_{4}$C and 304L type S.S. and it was concluded that the kinetics of chemical interaction between S.S. and granular B$_{4}$C is clearly different from that between S.S. and pellet or powder B$_{4}$C.

DOI： 10.1080/00223131.2015.1011721

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Transmutation of long-lived radioactive nuclides including minor actinides (MA: Np, Am, Cm) has been studied in Japan Atomic Energy Agency (JAEA). Accelerator-driven system (ADS) is regarded as one of the powerful tools for transmutation of MA under the double strata fuel cycle concept. Uranium-free nitride fuel was chosen as the first candidate fuel for MA transmutation using ADS. To improve the transmutation ratio of MA, reprocessing of spent fuel and reusing MA recovered from the spent fuels is necessary. Our target is to transmute 99\% of MA arisen from commercial power reactor fuel cycle, with which the period until the radiotoxicity drops below that of natural uranium can be shorten from about 5000 years to about 300 years. A pyrochemical process has been proposed as the first candidate for reprocessing of the spent nitride fuel. This paper overviews the current status of the nitride fuel cycle technology. Our recent study on fuel fabrication, fuel property measurements, reprocessing of spent fuel, development of the property database of MA nitride fuel, and fuel behavior simulation code are introduced. Our research and development (R\&D) plan based on the roadmap of the development is also introduced.

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DOI： 10.1016/j.jnucmat.2015.04.043

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Thermodynamic database is a useful tool for evaluating preliminary chemical reactions during molten core concrete interaction and phase relations in fuel debris. Japan Atomic Energy Agency and Nippon Steel Sumitomo Metal Corporation are collaborating to develop an open thermodynamic database by combining their own nuclear- and steel/slag-databases. The phase diagrams of Al$_{2}$O$_{3}$-UO$_{2}$, Al$_{2}$O$_{3}$-ZrO$_{2}$, SiO$_{2}$-UO$_{2}$, and UO$_{2}$-ZrO$_{2}$ sub-systems and those of Al$_{2}$O$_{3}$-UO$_{2}$, SiO$_{2}$-UO$_{2}$, and Al$_{2}$O$_{3}$-ZrO$_{2}$ sub-systems were able to be drawn properly using the cell model parameters and the sub-lattice model parameters assessed in the present study, respectively. The concerns for improving accuracy on the assessment were extracted in several sub-systems.

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To continue the utilization of the nuclear fission energy, the management of the high-level radioactive waste is one of the most important issues to be solved. Partitioning and Transmutation technology of HLW is expected to be effective to mitigate the burden of the HLW disposal by reducing the radiological toxicity and heat generation. The Japan Atomic Energy Agency (JAEA) has been conducting the research and development on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. This paper overviews the recent progress and future R\&D plan of the study on the ADS and related fuel cycle technology in JAEA.

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For improvement of Severe Accident (SA) codes, a detailed core relocation process has to be clarified because the numerical methods in SA codes for analyzing the core relocation process are simplified. Especially, it is very important to understand precisely the melting behavior of fuel elements including fuel rods, fuel channel boxes and control rods. Then we have been developing the numerical method for analyzing fundamental melting behavior of the fuel elements based on the original POPCORN code. The POPCORN code was developed in Japan Atomic Energy Agency (JAEA) based on the Moving Particle Semi-implicit (MPS) method. In the present study, we introduced the numerical analysis models for phase change, surface tension and multi components to the original POPCORN code. This paper shows the outline of the simulation methods and results of numerical simulations, and it was confirmed from the present results that melting behavior called candling can be simulated numerically.

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Two type tests were performed to examine the prevention effect of the oxide layers to liquefaction interaction between Zircaloy and core component materials. The oxide layer of Zircaloy was clearly confirmed to prevent the liquefaction interaction under the temperature of melting point of stainless steel even if the oxide layer on Zircaloy of approximately 30 micro meters, which is thinner than it is predicted to be formed under typical accident condition. The oxide layers were able to be formed even in the region where the gas flow is significantly limited by narrow arrangement. Although the oxide layers at the inner position of upper end plug was hard to form, the prevention effect of the oxide layers was sufficiently observed. The axial variation of the thickness of the oxide layers was observed. It suggested that variation of partial pressure of H$_{2}$O should be considered to evaluate the growth rate of the oxide layers for detail.

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Nitride fuels have several advantages, such as high thermal conductivity and high metal density like metallic fuels, and high melting point and isotropic crystal structure like oxide fuels. Since the late 1990s, the partitioning and transmutation of minor actinides (MA) has been studied to decrease the long term radio-toxicity of high level waste and mitigate the burden on the final disposal. Japan Atomic Energy Agency (JAEA) has been proposing dedicated transmutation cycle using the Accelerator-Driven System (ADS) with the nitride fuels containing MA. We have been developing the nitride fuel cycle including pyrochemical process. Our focus is on electrolysis of nitride fuels and refabrication of nitride fuel from the recovered actinides because other processes are similar to the technology for the metal fuel treatment and have been studied elsewhere. In this paper, we summarized our activity on developments of the pyrochemical treatment of the spent nitride fuels.

DOI： 10.1007/s11426-014-5205-9

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As for improvement of Severe Accident (SA) codes, a detailed core relocation process has to be clarified because higher prediction results are obtained by numerical methods in the SA codes. Especially, it is very important to understand precisely the melting behavior of fuel elements. Then we have been developing the numerical method for analyzing fundamental melting behavior of the fuel elements based on the original POPCORN code. The POPCORN code was developed in Japan Atomic Energy Agency (JAEA) based on the Moving Particle Semi-implicit (MPS) method. In the present study, we introduced the numerical analysis methods for phase change, surface tension and multi components to the original POPCORN code. This paper shows outline of simulation methods and results of numerical simulations. From the results, melting behavior called candling can be simulated numerically.

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Control blade degradation is predicted to occur at the early stage of severe accident of boiling water reactor. Simulation tests and modeling are being performed by collaboration among JAEA, Hitachi-GE Nuclear Energy (Hitachi-GE), Nippon Nuclear Fuel Development (NFD), and Nippon Steel \& Sumitomo Metal Corporation (NSSMC). Preliminary simulation tests under inert atmosphere indicated that complicated chemical reactions progressed and that finally the mixture of degraded control blade and the channel box separated into two parts. The first part rich in Zr had low melting temperature and flowed down from the high temperature region like candling. The second part rich in steel, B, and C mechanically collapsed down at temperatures higher than approximately 1673K. The interaction between the control blade and the channel box was highly exothermic. Preliminary model for control blade degradation is being developed using VOFFLUENT, which successfully simulates the early stage ofthe control blade degradation.

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In the accident of Fukushima-Daiichi Nuclear Power Plant, degraded fuels containing Zircaloy probably reacted with B$_{4}$C control blades containing stainless steel cladding or blade sheath. Since light elements like B and C are able to react easily with various elements and form various chemical species, several concerns are pointed out, such as variation in volatility and heat generation by oxidation of B and C. The chemical states of degraded fuel were evaluated on the assumption of thermodynamic equilibrium under various conditions of oxygen potential and temperature. The chemical behavior of B affects significantly the variation in oxygen potential with progressing severe accident, and many kinds of volatile compounds are formed by oxidation. The behavior of B causes the changes of volatility of FPs, such as Sr, Cs and Mo.

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The depth profiles of hydrogen were measured at outer surface of fuel claddings corroded in high temperature steams at 1073-1473 K to examine the barrierness of surface oxide layer against the hydrogen absorption/desorption. The results indicated that the oxide layer would be no longer the barrier against the hydrogen under some conditions although it remained as the barrier against the oxidation.

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Gibbs free energies of formation of six Ce-Cd intermetallic compounds, CeCd, CeCd$_{2}$, CeCd$_{3}$, CeCd$_{58/13}$, CeCd$_{6}$ and CeCd$_{11}$, were evaluated systematically using electrochemical techniques in the temperature range from 673 to 923 K in the LiCl-KCl-CeCl$_{3}$-CdCl$_{2}$ molten salt bath. The linear dependence of the Gibbs free energies of formation on temperature yields to the enthalpies and entropies of formation of these intermetallic compounds. By extrapolating the molar Gibbs free energy of Ce-Cd intermetallic compounds to the Cd distillation temperature, it was clear that the molar Gibbs free energy of Ce in Ce-Cd intermetallic compounds decreases gradually from CeCd$_{11}$ to CeCd$_{2}$ and attains to the minimum value at CeCd$_{2}$. This suggests on the Cd distillation from the U-Pu-Ce-Cd alloy that the dissolution of U or Pu into CeCd$_{2}$ should be mostly taken into consideration.

DOI： 10.1016/j.jpcs.2014.04.004

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Language：Japanese   Publisher：Atomic Energy Society of Japan  

In the severe accident of Fukushima-Daiichi Nuclear Power Plant, large amount of sea water was introduced into reactor pressure vessels. Not only sodium chloride but also several minor elements contained in sea water possibly reacted with degraded fuel debris or molten corium. Varous concerns are pointed out, such as volatilization of FP, characterization of fuel debris, formation of corrosive gases, and etc. Thermodynamic evaluation can give useful information on the general tendency of these sea-water effects. Volatility of Cs, Sr, and Te is potentially increased due to the change in the chemical species. Corrosive gases, such as HCl, H$_{2}$S, and etc. are possibly generated from sea-water heated at high temperature. These phenomena are predicted to be varied with change in oxygen potential.

DOI： 10.3327/taesj.J13.001

CiNii Books

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Language：English  

Electrorefining is one of the main steps of pyroreprocessing where spent nuclear fuels are recycled. Electrorefining is conducted in a molten salt of LiCl-KCl eutectic (59:41 mol\%) containing actinide chlorides (AnCl$_{3}$) at 773 K. In order to operate and maintain the electrorefiner, it is necessary to accumulate fundamental data on LiCl-KCl-AnCl$_{3}$ salt such as the melting point. In this study, based on X-ray diffraction and differential thermal analysis, a partial phase diagram of (LiCl-KCl)eut.-UCl$_{3}$ pseudo-binary system and partial phase diagram of LiCl-KCl-UCl$_{3}$ system were developed, which UCl$_{3}$ concentration was up to 20 mol\%.

DOI： 10.1016/j.jnucmat.2013.06.019

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Language：English  

Neptunium trichloride of high purity was synthesized by the solid-state reaction of neptunium nitride, which was prepared from the oxide by the carbothermic reduction method, and cadmium chloride in a similar manner as reported for synthesis of AmCl$_3$. Lattice parameters of hexagonal NpCl$_3$ were determined from the X-ray diffraction pattern to be a = 0.7421 $\pm$ 0.0006 nm and c = 0.4268 $\pm$ 0.0003 nm, which fairly agree with the reported values (a = 0.742 $\pm$ 0.001 nm and c = 0.4281 $\pm$ 0.0005 nm). Melting temperature of NpCl$_3$ was measured with about 1 mg of the sample which was hermetically encapsulated in a gold crucible using a differential thermal analyzer with heating and cooling rate of 10 K/min in an argon gas flow (50 mL/min). The melting temperature of NpCl$_3$ was determined to 1070 $\pm$ 3 K, which is close to the recommended value 1075$\pm$30 K, which was derived from the mean value of the melting temperature for UCl$_3$(1115K) and that for PuCl$_3$ (1041 K).

DOI： 10.1016/j.jnucmat.2013.01.316

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Language：English  

Estimation of the accident progress and status inside the reactor is required to properly and reliably conduct decommissioning of the Fukushima-Daiichi NPPs. For that, it is necessary to obtain additional experimental data and revised models for the estimation using computer codes with increased accuracies. JAEA has selected phenomena to be reviewed and developed in terms of thermo hydraulic behavior in the reactor, progression of fuel bundle degradation, failure of the lower head of the reactor pressure vessel, and analysis of the accident, considering previously obtained information, conditions specific to the Fukushima-Daiichi NPP accident, and recent progress of experimental and analytical technologies. This paper introduces the selected phenomena to be reviewed and developed and recent results from the JAEA's corresponding research programs.

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Language：English  

High temperature interaction tests between UO$_{2}$ and Zr were performed at around 2173 K, to make clear the UO$_{2}$/ $\alpha$-Zr(O) interaction and the mechanism of degradation, for developing the improved models for advanced severe accident analysis codes. A Zr plate was inserted in a UO$_{2}$ crucible, and heat treated at 2173 K in stream of Ar. After the heat-treatment, the samples were subjected to surface microanalysis. The middle region of Zr sample shows streak-like structures which are extended towered the top. It is confirmed that the streak-like structures were mainly consist of U from the EDX results, and the structures revealed that the U-rich phase was liquid during the heat-treatment. It seems that the U-rich liquid grew selectively toward the area where the oxygen concentration was low.

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Language：English  

The paper summarizes the first results of a thorough SEM investigation uncovering the process of channel box wall penetration by Fe-Cr-Ni-B containing melt. The preliminary oxidation of channel box is shown to play an important role on severe accident progression resulted in the suppression of channel box massive destruction. Only one small droplet came out to the other side of channel box. The mechanism of local beginning of oxide layer destruction with subsequent Zircaloy-4 channel box penetration is under discussion.

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Language：Japanese  

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Language：English  

Uranium-free nitride fuel was chosen as the first candidate for transmutation of long-lived minor actinides (MA) using accelerator-driven system (ADS) under the double strata fuel cycle concept by Japan Atomic Energy Agency (JAEA). The advantages of nitride fuel are good thermal properties and large mutual solubility among actinide elements. A pyrochemical process is proposed as the first candidate for the reprocessing of the spent nitride fuel, because this technique has some advantages over aqueous process, such as the resistance to radiation damage, which is an important issue for the fuels containing large amounts of highly radioactive MA. This paper overviews the recent progress and future R\&D plan of the study on the nitride fuel cycle technology in JAEA.

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Language：English  

The chemical and transport behaviors of Pu and U in the corroded debris must be understood for the criticality safety control of Pu and U in the debris, especially for the removal operations and storage. Therefore, the chemical changes of UO$_{2}$ and PuO$_{2}$ powders, disks and U and Pu metal disks in H$_{2}$O$_{2}$ aqueous solution have been checked, where H$_{2}$O$_{2}$ is formed by the radiolysis of H$_{2}$O. As a result, UO$_{2}$ changed to hydrated uranium peroxide, whereas the PuO$_{2}$ remained unchanged. U metal was more reactive with H$_{2}$O$_{2}$ aqueous solution than Pu metal. The chemical changes of the mixed UO$_{2}$/PuO$_{2}$ powders in H$_{2}$O$_{2}$ aqueous solution have been investigated. The dried slurry of the middle zone of H$_{2}$O$_{2}$ aqueous solution was mainly composed of hydrated uranium peroxide, whereas the dried powder of the bottom zone was mainly composed of PuO$_{2}$.

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Language：English  

A new research program on severe accidents is lunched for the evaluation of FP release and transport behavior in BWR system. The purpose of the program is to improve the FP release and transport model using experimental database about FP chemistry focusing on Cs and I chemistry. In this program, effects of B including in control rod materials, B$_{4}$C for the Cs and I chemistry are paid attention. The experimental database used for the improvement will consist of results to obtain with newly-prepared test device under atmosphere with broad-ranging oxygen and/or steam partial pressure simulated those in BWR. The state of preparation for these experimental studies and analyses is introduced. In addition, the preliminary test was moved into action to show B chemical effect on Cs and I transport under one of the processes, which is deposited Cs compounds and B vapor and aerosol interaction. In this experiment, a "B stripping effect" to deposited CsI was observed.

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Language：English   Publishing type：Internal/External technical report, pre-print, etc.  

The purpose of this study is to prepare a property database of nitride fuel needed for the fuel design of accelerator-driven system (ADS) for transmutation of minor actinide (MA). Nitride fuel of ADS is characterized by high content of Pu and MA as principal components, and addition of a diluent material such as ZrN. Experimental data or evaluated values from the raw data on properties Pu and MA nitrides, and nitride solid solutions containing ZrN are collected and summarized, which cover the properties needed for the fuel design of ADS. They are expressed as an equation as much as possible for corresponding to a variety conditions. Error evaluation is also made as much as possible. Since property data on transuranium (TRU) nitrides are often lacking, those on UN and (U,Pu)N are substitutionally shown in such cases in order to facilitate the fuel design with a tolerable accuracy by complementing the database.

DOI： 10.11484/jaea-data-code-2014-001

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Language：Japanese  

By installing the laser flash apparatus and the drop calorimeter in the glove box, the thermal diffusivity and the heat capacity measurements of nitride containing MA elements of long-lived radioactive nuclides were enabled. The sample holder and the platinum container were designed to measure the thermal diffusivity and the heat capacity of very small quantity of MA nitride samples. The thermal conductivities of MA nitride increased with temperature, unlike that of conventional oxide-type nuclear fuels. In addition, the thermal conductivities of MA nitride decreased with increasing Am contents. The thermal conductivity of ZrN-based MA nitride, which is proposed as a candidate material for the ADS fuel, was fitted to equations as functions of the temperature and ZrN concentration. The predicted values agreed well with the experimental ones, indicating that the thermal conductivity of nitride fuel for ADS can be predicted for a practical design.

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Language：English  

JAEA has proposed a double strata concept for partitioning and transmutation of transuranium elements, in which the second P-T cycle is consisted of an ADS for transmutation and of a pyro-reprocessing for partitioning. Since the nitride fuels for the ADS contain large amounts of TRU, comprehensive knowledge on electrochemical properties for TRU, such as Np, Pu, Am, and Cm, in LiCl-KCl eutectic molten salt is necessary for developing the pyro-reprocessing. It is also required to study anodic dissolution behavior of the spent nitride fuels and re-nitridation of recovered TRU. Based on knowledge, a process flow diagram is being constructed. This paper highlights the fundamental dataset on the electrochemical properties of TRU obtained in JAEA, as well as preparation of facilities and TRU samples.

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Language：English  

In the severe accident at Fukushima Daiichi Nuclear Power Station, the seawater was pumped into the cores. When the seawater is vaporized on the hot fuel debris, the various salts like chlorides and sulphates are deposited. In this work, the reactivity between UO$_{2}$ and the sea salt was examined in the temperature range from 815 to 1198 $^{\circ}$C, to characterize the impacts of seawater on the chemical form of fuel debris. The pelletized mixture of UO$_{2}$ and the salt heated for 30 min in argon flow yielded mainly MgO and UO$_{2}$, whose lattice parameter is much smaller than that of stoichiometric UO$_{2}$. It was found that calcium decomposed from the sulphate could dissolve into UO$_{2}$ to form solid solution. However, by the analyses on the UO$_{2}$ pellets heated with the salt at 913 and 1002 $^{\circ}$C for 3 h, the reaction was found to be limited at the UO$_{2}$/salt deposit interface. The evolution of corrosive gases by the volatilization and thermal decomposition of the salt would be another aspect of importance for the structural materials and chemical form of the volatile fission products in the core.

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Event date： 2021.12

Language：English  

Venue：Karlsruhe (online)   Country：Germany  

The next step in the sequence of severe accidents tests performed in JAEA for clarification of the situation in the damaged 1F reactors was to investigate the BWR materials behaviour and the features of bundle degradation under steam-rich conditions and relatively high heating rate of 1 K/s. The presentation outlines the results of the CLADS-MADE-03 test.

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Event date： 2021.9

Language：English  

Country：Japan  

The work on analysis of metallic sim-debris have shown similarity between the test and real debris, which allowed the interpretation of the Unit 2 PCV debris origin and uncovered control blade degradation mechanism with formation of boron compounds in the metallic part of melt. Though the information was surely lacking, a rough estimation of the chemical composition of metallic debris and the place of the first solidification was made.

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Event date： 2020.11

Language：English  

Venue：online  

R\&Ds for the decommissioning of the Fukushima Daiichi Nuclear Power Station in Collaborative Laboratories for Advanced Decommissioning Science, JAEA, are introduced including the debris behaviour in lower plenum test, the control blade degradation behaviour test, the fuel debris characterisation test, and the OECD/NEA international joint projects activity.

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Event date： 2020.9

Language：English  

Country：Japan  

Since the severe accident at Fukushima Daiichi Nuclear Power Plant, development of accident tolerant fuel claddings such SiC has been intensively discussed. However, application of SiC as fuel claddings has raised some concerns over its performance under severe accident conditions in steam. In this study, investigation of steam oxidation for SiC at 1400-1600$^{\circ}$C under 0.1 and 3 g/min steam flow rate was performed using newly developed laser heating facility. Results of the investigation showed that the SiC experienced a mass gain behavior under 0.1 g/min steam flow rate. Whereas, it suffered a mass loss under 3 g/min steam flow rate. Based on the mass change data obtained in the investigation, reaction rate constants and activation energy of the process were calculated and reported.

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Event date： 2020.2

Language：Japanese  

Country：Japan  

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Event date： 2019.10

Language：English  

Venue：Karlsruhe   Country：Germany  

BWR has heterogeneous configuration of fuel assembly, especially control blade. Growth and reduction behavior of oxidation depend on steam fraction. The oxidation behavior affects the reaction between SUS and Zry in an early stage of accident progression. According to ANSYS-fluent using oxidation growth and reduction models depending on steam fraction, oxidation behavior drastically change above 1200$^{\circ}$C. Steam fraction on the surface of channel box covered by control blade decreases up to steam-starved condition around 1200$^{\circ}$C, which means oxidation growth rate decreases depending on steam fraction. Oxidized layer grows continuously even steam-starved condition using a model of oxidation behavior independent of steam fraction. According to experimental integral tests without pre-oxidation, a reaction between SUS-B$_{4}$C and Zry occurs under low steam flow rate because the reaction rate is higher in the case of thin oxidized layer. Almost full oxidization occurs on the surface of Zry under a sufficient steam flow condition, except for thin oxidation at a top part of the test body. The reaction between control rod and Zry occurs at the top part. In addition, an observation of oxidation behavior in an integral test with pre-oxidation show us that a thickness of oxidized layer in a part of Zry covered by control blade is thinner and the interaction between SUS-B$_{4}$C and Zry begins even pre-oxidation. The model will be implemented into JUPITER code developed by Nuclear Science and Engineering Center.

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Event date： 2019.10

Language：English  

Venue：Karlsruhe   Country：Germany  

Preliminary results on the CLADS-MADE-02 control blade degradation test related to the problem of 1F Unit 3 accident under sufficient steam conditions will be discussed. Control blade debris distribution obtained in oxidative environment, possibility of B$_{4}$C interaction with hot steam, consequences of its reaction with surrounding stainless steel and Zircaloy are very important for 1F decommissioning. Experimental information with in situ video of an melt progression will be presented at the 25th international QUENCH workshop famous for attracting experts dealing with nuclear safety.

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Event date： 2019.10

Language：English  

Venue：Karlsruhe   Country：Germany  

One of the promising methods of debris characterization - a Raman spectroscopy was utilized for investigation of simulated debris obtained after two control blade degradation tests CLADS-MADE-01 and CLADS-MADE-02. New results allowed to refine a mechanism of the B$_{4}$C degradation during the beginning phase of a severe accident. A partial transformation of B$_{4}$C granules into pure graphite promotes formation of particular phases in the debris based on the oxidative or reducing environment. This knowledge would provide new insights for understanding of the absorber blade degradation mechanism under specific accident conditions close to 1F Unit 2 and Unit 3 reactors.

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Event date： 2019.10

Language：English  

Country：Japan  

Not sufficiently investigated role of B$_{4}$C in BWR assembly degradation challenges success of Fukushima Dai-ichi (1F) decommissioning. The control blade degradation under Unit 2 and Unit 3 conditions of 1F is unclear because understanding of high-temperature B$_{4}$C interaction with BWR assembly materials with formation of new phases is still lacking. Using Raman spectroscopy of debris of control blade degradation tests it was established that B$_{4}$C degradation happens through graphitization both under steam-starved and full steam conditions.

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Event date： 2019.9

Language：Japanese  

Country：Japan  

In Japan, research and development (R\&D) program for establishing technical basis of ATF started in 2015 and is being conducted in cooperation with power plant providers, fuel venders and universities for making the most use of the experiences in R\&D, practical design, and evaluations of fuels and cores of commercial LWRs. The present paper reviews the progress of the development and summarizes subjects to be solved for ATF cladding, based on the attribute guide which was originally drawn up in the Japanese ATF R\&D program as it should be an indicator of proper ATF development guidance.

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Event date： 2019.9

Language：Japanese  

Country：Japan  

Based on the analysis of the samples collected in the PCV interior of Fukushima Daiichi NPP, potential issues regarding the fuel debris retrieval are summarized. The analytical targets, analytical method and obtainable knowledge that are necessary to evaluate each issue are discussed. Valuable information can be obtained even from small samples to understand the current situation of the plant and the accident progression. Although the amount of collectable samples is limited, we considered how to evaluate the whole fuel debris and how to utilize the analytical results for recriticality, burnup and safety issues.

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Event date： 2019.9

Language：English  

Country：Japan  

In this study, oxidation of SiC under various flow rates of steam was investigated at 1400$^{\circ}$C using laser heating. Results of the investigation indicated that the mass evolution of SiC underwent a mass gain with 0.1 g/min of steam flow rate. Whereas, the mass evolution experienced a mass loss with 3 g/min of steam. Based on the data of mass change under the investigated conditions, the parabolic and linear rate constants were calculated and reported.

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Event date： 2019.7

Language：Japanese  

Country：Japan  

With a view to investigating the tendency of segregation of constituents of fuel debris, liquefaction/solidification tests of simulated corium (UO$_{2}$, ZrO$_{2}$, FeO, B$_{4}$C and sim-FP oxides) were performed. Simulated corium was heated up to 2600$^{\circ}$C under Ar atmosphere and then cooled down with two different cooling processes; furnace cooling and slow cooling. Cross sectional images of both samples showed that the oxide layer and metallic layer were separately solidified. EDX analysis of oxide layer of samples revealed three oxide phases and Fe metal. Solubility of FeO in these oxide phases was mostly fixed to be 12$\pm$5at\% in all phases, which is in reasonable accordance with the value estimated from UO$_{2}$-ZrO$_{2}$-FeO phase diagrams. The tendency of re-distribution of oxide elements observed in the sample with slow cooling was similar to that of quick cooling, however, the grain-growth of one oxide phase, rich in Zr-oxide, occurred at the bottom region of the oxide layer. The composition of this Zr-oxide phase is comparable to the initial average composition. The condensation is considered to be caused by the connection of remaining liquid agglomerates during slow solidification.

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Event date： 2019.7

Language：English  

Venue：Naraha   Country：Japan  

The new data from video investigation of the 1F Unit 2 pedestal debris performed by TEPCO was analysed. The debris features as derived from visual appearance on the video compared with the debris obtained after the CLADS-MADE-01 test. Some speculative conclusions concerning the properties and possible nature of the debris can be made.

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Event date： 2019.7

Language：English  

Venue：Naraha   Country：Japan  

The problem of control blade degradation which involves boron carbide in the eutectic reaction is one of the key issues for understanding of the accident progression at the beginning phase of the severe accident. In the 1f Unit 2 postulated integral test conditions the interaction of B with the hot steam was found to be rather difficult because of melting and encapsulation of the B materials in the debris. Thus the amount of B release for BWRs in Japan can be significantly lower. The presentation discusses the possibility of unpredictable B accumulation in some unpredictable places in the reactor core.

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Event date： 2019.7

Language：English  

Venue：Naraha   Country：Japan  

Macroscopic segregation of molten core components occurs with slow cooling rate in the accident of Fukushima Daiichi Nuclear Power Plants. In order to investigate these segregation behavior, the solidification model for numerical simulation has been developed. In this model, solidification and microscopic segregation of molten corium are simulated with the Scheil model and thermal properties calculated by Thermo-calc. In this study, the validation of macrosegregation analysis of this model were performed. As the preliminary analysis, the calculation results were compared with corium solidification experiments. It was proved that this model can estimate the tendencies of macrosegregation.

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Event date： 2019.7

Language：English  

Venue：Naraha   Country：Japan  

JAEA and NSTec develop a thermodynamic database for fuel debris and concrete based on iron and concrete system, which has been developed for analysis of steel making process. Ionic two-sublattice model and cell model are examined for thermodynamic models of liquid oxide used in SA analysis. The current state and issues of the database used the models will be discussed.

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Event date： 2019.7

Language：English  

Venue：Naraha   Country：Japan  

Since the disaster at Fukushima Daiichi Nuclear Power Plant, development of accident tolerant fuel to manage severe accidents has been proposed. As temperature of fuel rods could reach to temperatures above 2000$^{\circ}$C during the accident, investigation of oxidation behavior at such temperatures for fuel claddings must be conducted. Nevertheless, there is no test facility capable for operating at such conditions. In this study, we developed a laser heating facility for steam oxidation of cladding materials at extreme temperatures. Main features of the facility are controlled environment, in situ observation, extreme fast heating and cooling rate and off gas analysis. As the heat is focused on the test sample only, the facility can be operated at temperatures above 2000$^{\circ}$C. Preliminary tests have been conducted for monolithic SiC at 1400-1800$^{\circ}$C for 1-7 h under 0.96 atm steam partial pressure. Results of the investigation showed that SiC underwent a mass loss which obeyed paralinear laws. Based on the mass change data, parabolic oxidation rate and linear volatilization rate of the process were obtained. The apparent activation energy of the oxidation process was calculated to be 96 kJmol $^{-1}$. Bubbling phenomenon was observed on SiC sample surface at 1800$^{\circ}$C.

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Event date： 2019.7

Language：English  

Venue：Naraha   Country：Japan  

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Event date： 2019.7

Language：English  

Venue：Kyoto   Country：Japan  

Research and development (R\&D) program for establishing technical basis of ATFs for light water reactor (LWR) started in 2015. Since then the R\&D is being conducted in cooperation with power plant providers, fuel venders, research institutes and universities for making the most use of the experiences in R\&D, practical design, and evaluations of fuels and cores of commercial LWRs. Among currently explored ATF candidate materials in the program, silicon carbide composite reinforced by SiC fiber (SiC/SiC) and FeCrAl steel strengthened by dispersion of fine oxide particles (FeCrAl-ODS) offer several attractive features including the remarkable high temperature capabilities and the slow kinetics of steam oxidation reactions. This presentation will give an overview of the progress in ATF development and review the current status of data availability and integrity for the properties and behaviors of ATF candidate materials, followed by discussion on the primary differences from zirconium alloy in the behaviors in the severe accident scenarios. Finally, subjects to be solved for practical use of ATF will be summarized.

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Event date： 2019.4

Language：Japanese  

Country：Japan  

Japan Atomic Energy Agency (JAEA) has been developing and compiling a thermodynamic database of nuclear fuels and reactor materials for multi-purposes, including the prediction of fuel degradation progression in severe accident conditions and the characterization of chemical properties of fuel debris. We have constructed a website. The phase diagrams, thermodynamic data and liquidus-solidus surfaces calculated using the database are introduced in the website.

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Event date： 2019.3

Language：Japanese  

Country：Japan  

Various numerical simulations were performed by JUPITER code in order to validate the reliability of its multi-physics models, which were developed for evaluating the melting and relocation behavior of the core materials. By comparing with the previous experimental results, we could conclude that JUPITER code is a useful tool on severe accident analysis.

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Event date： 2019.3

Language：Japanese  

Country：Japan  

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Event date： 2019.3

Language：Japanese  

Country：Japan  

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Event date： 2019.3

Language：English  

Country：Japan  

Silicon carbide has been considered as a potential candidate for fuel cladding. Many studies have been conducted to investigate the steam oxidation of SiC at temperatures below 1600$^{\circ}$C. However, the steam oxidation behavior of SiC at temperatures above 1600$^{\circ}$C remained unclear due to the lack of test facilities. In this study, we investigated the steam oxidation of SiC at temperatures ranging from 1400-1800$^{\circ}$C using a newly developed laser heating facility.

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Event date： 2019.3

Language：Japanese  

Country：Japan  

The high temperature interaction tests between UO$_{2}$ and Zr were performed at around 2173 K in stream of Ar, to make clear the UO$_{2}$ and Zircaloy cladding interaction and the mechanism of degradation. After the heat-treatment, the samples were subjected to surface microanalysis. The mechanism of fuel degradation was more clarified.

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Event date： 2019.3

Language：Japanese  

Country：Japan  

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Event date： 2019.3

Language：Japanese  

Country：Japan  

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Event date： 2018.11

Language：English  

Venue：Karlsruhe   Country：Germany  

The present study is focused on the issue of a control blade degradation, which is the first step during BWR severe accidents. In particular problem of heterogeneous core degradation, boron behaviour, control blade liquefaction and melt relocation were analysed. Melt relocation features: onset, progress, candling, blockage formation and its subsequent melting was directly observed in situ using video cameras. Carefully tailored experimental conditions of the CLADS-MADE-01 test, which include preliminary oxidation, transient heating and long steam starvation phase, enabled successful reproducing of an accident scenario close to one that assumed to happen at Fukushima-1 Unit 2.

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Event date： 2018.10

Language：English  

Venue：Matsue   Country：Japan  

Silicon carbide (SiC) has recently attracted much attention as a potential material for accident tolerant fuel cladding. High-temperature interaction of SiC at temperature range below 1600$^{\circ}$C has been studied by various authors. As expected for enhanced accident tolerant fuel cladding, high-temperature interaction of SiC must be also investigated at temperatures above 1600$^{\circ}$C. In this study, we investigate high-temperature interaction of SiC in steam at temperature range of 1700$\sim$1800$^{\circ}$C using laser heating furnace. The oxidation kinetics is then studied based on the weight change and observation on cross-sectioned surface of tested sample using field emission scanning electron microscope. Off-gas measurement of hydrogen (H$_{2}$) and carbon monoxide (CO) generated during the test was also conducted via sensor gas chromatography.

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Event date： 2018.10

Language：English  

Venue：Matsue   Country：Japan  

For a better understanding of the Fukushima-Daiichi Nuclear Power Plant accident, a reaction model of core melting process is very important. In this study, oxidized Zry plates were soaked into molten SS-B$_{4}$C mixture and heated into Ar gas at 1200$^{\circ}$C. Recession of ZrO$_{2}$ scale was observed in the both cases. An oxygen solid solution region in the Zry sample was observed below ZrO$_{2}$ scale, although it was not observed in Zry oxidized at 600$^{\circ}$C in high temperature steam conditions. When oxygen supply from the surrounding was halted, ZrO$_{2}$ scale can be reduced at high temperature such as 1200$^{\circ}$C by oxygen dissolution into Zry inside.

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Event date： 2018.10

Language：English  

Venue：Seattle  

Macroscopic segregation of molten core and melt components occurs with slow cooling rate in the accident of Fukushima Daiichi Nuclear Power Plants. In this study, solidification and microscopic segregation are simulated with the Scheil model and thermal properties calculated by Thermo-calc in order to investigate an influence of cooling conditions on macroscopic segregation. A macroscopic segregation behaviour has been calculated for UO$_{2}$-ZrO$_{2}$-FeO system, which are major oxides of molten core materials. According to calculated results, UO$_{2}$ and ZrO$_{2}$ was concentrated in initial solidification area. On the other hand, FeO were strongly concentrated in later solidification area. In addition, macroscopic segregation tends to be suppressed in the conditions of fast solidification rate and slow velocity of solidification interface.

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Event date： 2018.9

Language：Japanese  

Country：Japan  

Macroscopic segregation of molten core and melted components occurs with slow cooling rate in the accident of Fukushima Daiichi Nuclear Power Plants. The solidification model considering macroscopic segregation of molten core and melted components was developed in order to predict distribution of core elements. In this study, solidification and microscopic segregation were simulated with the Scheil model using thermal properties calculated by Thermo-calc. We estimated segregation behavior of molten core and investigated an influence of cooling conditions on macroscopic segregation.

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Event date： 2018.9

Language：Japanese  

Country：Japan  

Since the accident at Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of Light Water Reactors (LWRs) became a topic of serious discussion, and the research and development for improving the safety LWRs has been activated in many countries. This project is one of the projects that entrusted from the Ministry of Economy, Trade and Industry Agency for Natural Resources and Energy. Using a domestic human source, an analysis tool, know-how and experience, the basic technique to apply an accident tolerant fuel material to an existing LWRs is developed in the project. It is started from October, 2015 and is carrying out this project continuously.

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Event date： 2018.9

Language：English  

Venue：Sendai   Country：Japan  

After the 2011 Great East Japan Earthquake and Tsunami, and the events that followed at the Fukushima Daiichi Nuclear Power Station (1F), global interest has expanded in exploring fuels with enhanced performance during such rare events, with accident-tolerant fuel (ATF) development programs starting in many countries. In Japan, several ATF programs have been launched for feasibility study, development of manufacturing technology, and establishment of technical basis since the 1F accident. ATF candidate materials considered in Japan are silicon carbide (SiC) composite and FeCrAl steel strengthened by dispersion of fine oxide particles (FeCrAl-ODS). SiC composite is very attractive material because of its lower hydrogen generation rate and lower reaction heat in comparison with conventional Zirconium (Zr) alloys. Therefore, practical uses for a fuel cladding of pressurized water reactor (PWR) and for the fuel cladding, channel box of boiling water reactor (BWR) are expected. On the other hand, FeCrAl-ODS steel is a promising material and is considered to apply to the fuel cladding of BWR.

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Event date： 2018.9

Language：Japanese  

Country：Japan  

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Event date： 2018.9

Language：Japanese  

Country：Japan  

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Event date： 2018.9

Language：Japanese  

Country：Japan  

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Event date： 2018.9

Language：Japanese  

Country：Japan  

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Event date： 2018.9

Language：Japanese  

Country：Japan  

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Event date： 2018.9

Language：Japanese  

Country：Japan  

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Event date： 2018.9

Language：English  

Country：Japan  

In this study, soaking experiments of pre-oxidized zry plates in molten mixture of stainless steel and 5 mol\% boron carbide were conducted at 1200-1250$^{\circ}$C. Thickness of oxide layer, ZrO$_{2}$ was measured before and after the soaking test to investigate the ZrO$_{2}$ thickness recession in the molten mixture. Results of the study indicated that the thickness recession of ZrO$_{2}$ was mainly caused by the diffusion of oxygen from oxide layer to metallic-Zr region.

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Event date： 2018.7

Language：English  

Venue：Tokyo   Country：Japan  

In order to obtain the knowledge on the phase relationship and stratification behavior of oxide debris under the slow solidification condition, liquefaction/solidification tests of simulated oxide debris were performed by using a high frequency induction furnace. As the oxides components from the core melt and structural materials, sintered pellets of UO$_{2}$-ZrO$_{2}$-sim-FPs and powder of FeO and B$_{4}$C were prepared. The powder mixture and crashed sintered oxides were heated to 2600$^{\circ}$C in Ar atmosphere in the first step (partial liquefaction), and then solidified at two different cooling rates; furnace cooling (No.1) and slow cooling (5$^{\circ}$C /min, No.2). For the corium microstructure study, solidified samples were subjected to elemental analysis by SEM/EDX. Cross sectional images of both samples showed that the oxide layer and metallic layer were separately solidified. EDX analysis of oxide layer of No.1 sample revealed four phases; (U$_{0.6}$Zr$_{0.31}$Fe$_{0.09}$)O$_{2}$, (Zr$_{0.51}$U$_{0.34}$Fe$_{0.15}$)O$_{2}$, ZrO$_{2}$ and Fe metal. The tendency of re-distribution of oxide elements observed in No.2 was similar to that of No.1, however, the grain-growth of (Zr,U)O$_{2}$ phase occurred at the bottom region of the oxide layer. In these results, the phases detected in both samples were in reasonable accordance with those evaluated by the UO$_{2}$-ZrO$_{2}$-FeO phase diagram.

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Event date： 2018.7

Language：English  

Venue：Tokyo   Country：Japan  

Current presentation gives a deep analysis of the progression of the 1F accident at Unit 2. Based on that analysis the explanation of the strategy for a severe accident test program of JAEA-CLADS to contribute to understanding of accident progression at 1F Unit 2 is given. First result of CLADS-MADE-01 (BWR control blade degradation test) is presented.

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Event date： 2018.7

Language：English  

Venue：Tokyo   Country：Japan  

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Event date： 2018.7

Language：English  

Venue：Tokyo   Country：Japan  

Macroscopic segregation of molten core and melt components occurs with slow cooling rate in the accident of Fukushima Daiichi Nuclear Power Plants. In this study, solidification and microscopic segregation are simulated with the Scheil model and thermal properties calculated by Thermo-calc in order to investigate an influence of cooling conditions on macroscopic segregation. A macroscopic segregation behaviour has been calculated for UO$_{2}$-ZrO$_{2}$-FeO system, which are major oxides of molten core materials in various conditions. According to calculated results, UO$_{2}$ and ZrO$_{2}$ were concentrated in initial solidification area. On the other hand, FeO was strongly concentrated in later solidification area. FeO was significantly segregated because FeO does not be dissolved in UO$_{2}$ and ZrO$_{2}$. In addition, macroscopic segregation tends to become stronger in the conditions of slow solidification rate and fast velocity of solidification interface.

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Event date： 2018.4

Language：Japanese  

Country：Japan  

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Event date： 2017.10

Language：English  

Venue：Karlsruhe   Country：Germany  

Two control blade degradation test facilities which were designed for the investigation of high temperature degradation of the nuclear reactor elements, have been set up in Tomioka-site. One was a test facility for small-scale samples and another one was test facility for large-scale samples. In situ video of the experiment and mass-spectrometric analysis of the output gases can be carried out during tests, and heating rate can be controlled up to 1 K/s. The working area of the large-scale facility consists of the isolated heating area 1.2 m height with the water cooled lower part to achieve the natural temperature gradient conditions as during severe accident. Steam rate in large-scale facilities can be controlled in the range of 15-100 g/min, and a test piece for the control blade degradation tests can be used to simulate a part of the control blade and fuel assembly. Chemical reaction between stainless steel/B$_{4}$C and Zry is affected by the oxide layer on Zry. The oxidation behavior depends on steel flow rate and tendency of control blade degradation possibly has threshold condition. According to pretest, critical condition for reaction between molten control rod and channel box can be recognized. Further analysis and tests in various steam rate will be conducted to evaluate the critical condition.

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Event date： 2017.9

Language：Japanese  

Country：Japan  

Since the accident at Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of Light Water Reactors (LWRs) became a topic of serious discussion, and the research and development (R\&D) for improving the safety LWRs has been activated in many countries. In Japan, the R\&D project on accident tolerant fuel and other components (ATFs) of LWRs, which is sponsored and organized by the Ministry of Economy, Trade and Industry (METI), has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R\&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys in LWRs.

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Event date： 2017.9

Language：English  

Venue：Mito   Country：Japan  

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS) carried out several international cooperation to gather the knowledge and expertise on decommissioning and accident management. The cooperation fields are radioactive waste management in the decommissioning, characterization of fuel debris, investigation on core degradation process and remote monitoring technology etc.. In this presentation current activities and results on the international cooperation will be reported.

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Event date： 2017.9

Language：Japanese  

Country：Japan  

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Event date： 2017.9

Language：Japanese  

Country：Japan  

A research project has been started for thermophysical property evaluation of eutectic melt composed of B$_{4}$C as control rod material and stainless steel (SS), eutectic reaction experiments, material analysis, physical modelling of eutectic reaction and reactor application analyses in order to simulate the B$_{4}$C-SS eutectic melting and relocation behaviors in core disruptive accident of sodium-cooled fast reactor. This paper reports the project overview and the first year (JFY 2016) outcomes.

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Event date： 2017.7

Language：English  

Venue：Tokyo   Country：Japan  

Japan Atomic Energy Agency (JAEA) has to establish the scientific basis for the decommissioning of the Boiled Water Reactors (BWR) at Fukushima Daiichi Nuclear Power Station and provide the information on the nuclear fuel behavior during the nuclear accidents. To address these problems JAEA has developed several test facilities which combine unique features for the R\&D related to Severe Accident (SA) propagation behavior for BWRs. The technical data and main features of the facilities will be introduced to the International audience to give an understanding of JAEA research abilities and possible collaboration in the severe accidents research.

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Event date： 2017.7

Language：English  

Venue：Tokyo   Country：Japan  

In this study, soaking experiments of Zry plates in molten SS-B$_{4}$C were conducted to investigate the recession rate of Zry plates in molten SS-B$_{4}$C at 1175-1225$^{\circ}$C. The molten SS-B$_{4}$C mixture was prepared by heat-treatment at 1400$^{\circ}$C for 12 h of a powder mixture containing 5 mol\% B$_{4}$C and stainless steel. After decreasing temperature to the testing temperatures, Zry plates were soaked in the molten mixture for 0-60 min. After soaking tests, the Zry plates were pulled out of the molten mixture. The remaining thickness of Zry plates after the tests were measured by using optical microscope. To understand the influences of alloying elements in Zry and SS, three other sets of experiments (Zr plates in Fe-B$_{4}$C, Zr plates in SS-B$_{4}$C and Zry plates in Fe-B$_{4}$C) were also conducted. Results of the study indicated that the thickness of Zry plates decreased linearly with increase of soaking duration. The recession rate of Zry plates was increase with increase of testing temperatures. The value of apparent activation energy on the thickness recession rate of Zry plates was determined to be 333 kJ/mol. The alloying elements in SS showed a positive effect on suppression of Zry corrosion. In contrast, the alloying elements in Zry showed a negative effect on the process.

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Event date： 2017.7

Language：English  

Venue：Iwaki   Country：Japan  

JAEA is conducting two projects on OECD/NEA. One is PreADES (Preparatory Study on Analysis of Fuel Debris) aims at ensuring processes are in place for the debris retrieval and analysis. The other is TCOFF (Thermodynamic Characterization of Fuel Debris and Fission Products based on Scenario Analysis of Severe Accident Progression at Fukushima-Daiichi NPS (1F)) aims at improving of thermodynamic databases for core and FPs. Outlines are introduced at the poster.

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Event date： 2016.10

Language：Japanese  

Country：Japan  

In TEPCO's Fukushima Daiichi Nuclear Power Station (1F), the cooling capability was lost due to tsunami caused by the Great East Japan Earthquake. It is considered that the Zirconium (Zr) alloy fuel cladding was oxidized in the heated core and the subsequent temperature escalation due to the oxidation caused core melting. As lessons learned from the 1F accident, development of advanced fuel and core components with enhanced accident tolerance becomes the greater concern. We have started R\&D to establish the technical basis to implement the advanced fuel components to existing LWRs. The R\&D is conducted in cooperation with power plant providers, fuel venders, research institutes and universities who have been involved in the development of the advanced fuel components. In this presentation, we will introduce the R\&D program which are carried out under the Project on Development of Technical Basis for Safety Improvement at Nuclear Power Plants by Ministry of Economy, Trade and Industry (METI) of Japan.

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Event date： 2016.7

Language：Japanese  

Country：Japan  

At the fuel assembly degradation stage in severe accidents, chemical features of the intermediate products are expected to be changed depending upon the accident progressions. These differences are originated from the differences in oxygen potential and temperature, and are highly important for evaluating the relocation and stratification progress of the fuel debris. Two types of sim-test with the different oxygen potentials were performed to investigate these tendencies. The chemical features of the intermediate materials used in the tests were determined from the observations for the control blade and channel box degradation in our previous study. The present results indicate that the U concentration in the metallic layer is largely varied depending upon the oxygen potential of the atmosphere. Also, when the B$_{4}$C-Fe alloy, as of a typical intermediate product, coexists with UO$_{2}$ and Zr, the apparent red-ox reaction rate between UO$_{2}$ and Zr are mitigated.

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Event date： 2016.5

Language：English  

Venue：Awaji   Country：Japan  

After Fukushima Daiichi nuclear accident in 2011, enhancing the accident tolerance of light water reactor fuels became a very important issue and currently the development of accident-tolerant fuel (ATF) are in progress. SiC is an attractive candidate of ATF cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. SiC reacts much less than Zircaloy with steam, the generation of heat and hydrogen gas would be extremely suppressed. Thermodynamic evaluation on chemical interactions between UO$_{2}$ and SiC were performed on various possible conditions of oxygen potential and temperature in severe accident. SiC is converted to SiO$_{2}$ by reaction with O$_{2}$. SiC is also converted to SiO$_{2}$ by reaction with H$_{2}$O. However, the fraction of the sub-reaction for forming SiO increases than in the case of reaction with O$_{2}$ when comparing the results at the same temperatures and the same oxygen potentials.

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Event date： 2016.3

Language：Japanese  

Country：Japan  

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Event date： 2015.11

Language：English  

Venue：Kiev   Country：Ukraine  

Outline of R and D activities of CLADS

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Event date： 2015.11

Language：English  

Venue：Karlsruhe   Country：Germany  

To simulate each process of the core melt progression in detail and contribute the improvement of severe accident analysis codes, we have performed the melt relocation experiments by using simplified fuel support pieces of the BWR. In addition, we are developing a numerical simulation code, JUPITER, to simulate the melt relocation in a fuel assembly scale. In this study, the JUPITER was applied to numerical simulations of the melt behavior of the complicated fuel support structures and the melt relocation experiment. As a result, the JUPITER is able to simulate melt relocation behavior in the complicated structures stably, and showed good agreement with the experimental results qualitatively. Therefore, it was confirmed that the JUPITER has potential to provide accurate understanding of the melt relocation behavior. In the near future, in order to simulate the more realistic behavior of the melt relocation, we will implement the eutectic reaction model to the JUPITER and also proceed the experimental analysis.

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Event date： 2015.10

Language：English  

Venue：Karlsruhe   Country：Germany  

In this study, control blade degradation test under the conditions of axial temperature gradient and high temperature increase rate in Ar-atmosphere was performed as preliminary test in steam atmosphere to investigate the relocation of control blade melt and interaction with channel box under the conditions. It was found that the molten products redistribute to axial direction and interaction between S.S./B$_4$C-melt and Zry-4 progresses significantly in the condition that a few micron oxide layer exists on Zry-4. Moreover, as preliminary test in steam atmosphere, reaction test between the materials of control blade and channel box was carried out to investigate the effect of oxide layer formed on the surface of Zry-4 in steam atmosphere. It was found that in the condition of supplying oxygen continuously, the liquefaction interaction between Zry and S.S./B$_4$C-melt is considered to be prevented.

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Event date： 2015.8

Language：English  

Venue：Karlsruhe   Country：Germany  

Regarding METI-project entitled advanced multi-scale modeling of fuel degradation in severe accident, the present status and recent topics will be overviewed.

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Event date： 2015.7

Language：Japanese  

Country：Japan  

Present status of METI-project on multi-scale modeling on fuel degradation mechanism will be introduced.

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Event date： 2015.6

Language：English  

Venue：Vienna   Country：Austria  

This presentation overviews recent progress and future R\&D plan on the pyrochemical treatment of spent nitride fuel in JAEA. Some processes for pyrochemical treatment of nitride fuels have similarity to those of metal fuel for FR. Though main processes such as electrolysis of nitride fuels, refabrication of nitrides from the recovered actinides, and chemical dissolution of an anode residue have been studied especially for nitride fuel treatment, our concerns also include the subjects which are common to the treatment of metal fuel for FR.

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Event date： 2015.5

Language：Japanese  

Country：Japan  

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Event date： 2015.3

Language：Japanese  

Country：Japan  

The phase relations of the U-Fe-O system and the phase stability of the ternary compound UFeO$_{4}$ were investigated. The thermodynamic modelling has been performed for the U-Fe-O ternary system based on the obtained data and reported phase diagram of UO$_{2}$-FeO$_{1+x}$ system.

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Event date： 2015.3

Language：Japanese  

Country：Japan  

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Event date： 2014.10

Language：English  

Venue：Idaho Falls  

Uranium-free nitride fuel was chosen as the first candidate for transmutation of long-lived minor actinides (MA) using accelerator-driven system (ADS) under the double strata fuel cycle concept by Japan Atomic Energy Agency (JAEA). The advantages of nitride fuel are good thermal properties and large mutual solubility among actinide elements. A pyrochemical process has been proposed as the first candidate for reprocessing of the spent MA nitride fuel, because this technique has some advantages over aqueous process, such as the resistance to radiation damage, compactness of the facility, margin of criticality, and recycling feasibility of N-15. This presentation overviews recent progress and future R\&D plan on the pyrochemical treatment of spent nitride fuel in JAEA.

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Event date： 2014.10

Language：English  

Venue：Clearwater  

Outline of METI project on Advanced multi-scale modeling and experimental study for fuel degradation in severe accident is introduced as plenary lecture of NuMat 2014. Individual models, databases and test facilities developed in the project are introduced.

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Event date： 2014.9

Language：English  

Venue：Sendai   Country：Japan  

Although Fe-B-C ternary system is a dominant phase diagram when considering the control blade degradation, phase relation data around eutectic composition were not sufficient. The phase relations of three Fe-B-C samples were analyzed by XRD and SEM/EDX and solidus were determined by DTA. The solidus detected in the present study was maintained at about 1400 K for all three samples, although preliminary calculation using conventional thermodynamic database estimated that solidus varies with roughly 50 K difference. The difference might be originated from the insufficient evaluation on the phase stability of Fe$_{3}$(B,C) in the database. The present results give useful information on improvement of iron-boron-carbon phase diagram.

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Event date： 2014.9

Language：Japanese  

Country：Japan  

Research program on fission product (FP) chemistry has been launched towards the improvement of evaluation method of source term and FP release and transport behaviors. The purpose is to construct a database that is composed of FP chemical forms and their fractions at each stage of the nuclear power plant. The FP chemical behavior is modeled based on the experimentally determined FP chemical forms. The constructed FP chemical model will be verified and validated through the analysis of integrated tests for FP release and transport.

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Event date： 2014.9

Language：Japanese  

Country：Japan  

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Event date： 2014.9

Language：Japanese  

Country：Japan  

The outline and the present status of METI project entitled Advanced Multi-scale Modeling and Experimental Tests on Fuel Degradation in Severe Accident Condition are overviewed.

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Event date： 2014.6

Language：Japanese  

Country：Japan  

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Event date： 2014.4

Language：Japanese  

Country：Japan  

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Event date： 2014.4

Language：Japanese  

Country：Japan  

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Event date： 2014.3

Language：Japanese  

Country：Japan  

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Event date： 2014.3

Language：Japanese  

Country：Japan  

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Event date： 2014.3

Language：Japanese  

Country：Japan  

Present status on information exchange of accident torelant fuel in a frame of Japan-US CNWG is reviewed.

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Event date： 2014.3

Language：English  

Venue：Tokyo   Country：Japan  

JAEA conducts various R\&D to obtain the information for decommissioning of the Fukushima Daiichi NPP, better understanding of the accident progression, evaluation of effectiveness of accident mitigation measures, etc. The R\&D covers thermal hydraulic behavior, fuel rod degradation process, failure behavior of pressure vessel, molten materials relocation in the lower plenum region of BWR during accidents, fuel debris characterization and computer code development. The present paper introduces the outline of those R\&D.

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Event date： 2014.3

Language：Japanese  

Country：Japan  

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Event date： 2013.11

Language：English  

Venue：Shanghai   Country：China  

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Event date： 2013.11

Language：English  

Venue：Karlsruhe   Country：Germany  

The Japan Atomic Energy Agency (JAEA) has started new severe accident research programs focusing the issues where large uncertainties remain including the phenomena specific to Fukushima-Daiichi NPP accident: the influence of the injection of sea water into core, the failure behavior of fuel bundles and RPV in BWR, and the source term. JAEA has started acquiring basic data, performing analyses, developing models and preparing test apparatus on the sea water effects, the behavior of B$_{4}$C, the characteristics of corium debris and the fission product release behavior from fuel for the evaluation of accident progression with increased accuracies.

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Event date： 2013.11

Language：English  

Venue：Fukushima   Country：Japan  

In order to conduct decommissioning of the Fukushima-Daiichi NPPs properly and reliably, increase of accuracy is required in estimation of the accident progress and status inside the reactors. Knowledge of various phenomena is still lacking to improve models and computer codes for melt formation and relocation in BWR cores. It is also required to obtain information on fuel debris characteristic for the processes of fuel debris retrieval in the Fukushima Daiichi NPP. The present paper introduces the related activities and recent results at JAEA.

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Event date： 2013.11

Language：English  

Venue：Kyoto   Country：Japan  

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Event date： 2013.10

Language：English  

Venue：Chicago  

A METI project is ongoing to develop multi scale and multi physics models and databases for elucidating fuel degradation mechanism in severe accident under collaboration among JAEA, CRIEPI, NSSMC, and universities. In JFY2012, technology investigation and model calculation were performed. Base MPFM was developed to study liquefaction condition in mesoscale. Combination between VOF and FLUENT shows high technical potential to simulate candling. A copling among VOF, FLUENT and MARC shows the high technical potential to simulate real scale fuel assembly degradation. As for thermodynamic database, various concerns were extracted from model calculation using conventional database like NUCLEA. In the project, not only developing models and databases but also constructing test facilities are planned. In JFY2012, preliminal tests were performed using conventional test apparatuses and basic concepts for three test facilities to be developed were extracted.

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Event date： 2013.9

Language：Japanese  

Country：Japan  

In a severe accident of light water reactors, interaction layers are formed between UO$_{2}$ pellet and Zry cladding in the beginning stage. The melting temperature of a part of layers are rather lower than that of Zry cladding. In the present study, a base MPFM model was developed for understanding solid-solid diffusion and layer formation between UO$_{2}$ and Zr, which are major constituents of pellet and cladding, respectively. Then, liquefaction condition of each layer was discussed using MPFM model. Model calculation indicates that meta-stable liquid phase is possibly formed in intermediate stage of mutual diffusion even at 1773 K, which is far lower than melting temperature of Zry. Also, in case of mutual diffusion between UO$_{2}$ and Zr, spinodal separation of each phase potentially accerelates the layer formation.

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Event date： 2013.7

Language：Japanese  

Country：Japan  

Estimation of the accident progress and status inside the reactor pressure vessels (RPV) and primary containment vessels (PCV) is required for appropriate conductance of decommissioning in the Fukushima-Daiichi NPP. For that, it is necessary to obtain additional experimental data and revised models for the estimation using computer codes with increased accuracies. The Japan Atomic Energy Agency (JAEA) has selected phenomena to be reviewed and developed, considering previously obtained information, conditions specific to the Fukushima-Daiichi NPP accident, and recent progress of experimental and analytical technologies. As a result, research and development items have been picked up in terms of thermal-hydraulic behavior in the RPV and PCV, progression of fuel bundle degradation, failure of the lower head of RPV, and analysis of the accident. This paper introduces the selected phenomena to be reviewed and developed as well as the JAEA's corresponding activities.

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Event date： 2013.5

Language：English  

Venue：San Sebasti\'an   Country：Spain  

Phase field simulation model and thermodynamic database are being developed in JAEA for elucidating chemical mechanism of fuel degradation. The base simulation model and thermodynamic database developed by 2012 indicated that the chemical form of the first liquid generated in fuel and cladding diffusion couples is U-Zr alloy and the oxygen concentration in the liquid gradually increases with temperature. Then, in the high temperature region, the liquid is divided into tow melts, such as metal-melt and oxide-melt. These phenomena predicted from modeling agree well with experimental observations. Regarding solidification of fuel melt, spinodal potentially happens and phase separation is easily progressing. Also, Fe-B-C melt is formed from eutectic reaction between B$_{4}$C and the cladding. It potentially accelerates the fuel degradation.

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Event date： 2013.4

Language：Japanese  

Country：Japan  

Construction of thermodynamic database is one of the important issues for evaluating physical-chemical properties of fuel debri generated during a severe accident at Fukushima-Daiichi Nuclear Power Plant. The paper introduces the present status in JAEA for developing the database.

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Event date： 2012.12

Language：Japanese  

Country：Japan  

Outline of fuel degradation in severe accident of LWR is introduced to students of Nagaoka University of Technology.

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Event date： 2012.10

Language：English  

Country：Japan  

Since a fuel degradation in severe accident is caused by very complicated physical-chemical reactions, huge efforts are necessary to elucidate the detail mechanisms, such as accumulation of experimental data and construction of dedicated modeling. Efficient approach is indispensable. In the seminar, fundamental approaches in JAEA for these purposes are introduced, such as (1) thermodynamic database for degraded fuel, (2) dedicated modeling for various single steps, (3) thermodynamic evaluation of effect of sea-water or mixing between degraded fuel and degraded control rod, (4) segment-scale experiments for single step, and (5) plan of V-V experiments based on Fukushima-PIRT.

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Event date： 2012.10

Language：English  

Venue：Osaka   Country：Japan  

In the pyrochemical reprocessing of spent metallic fuels, electrorefining is conducted in a molten salt of LiCl-KCl eutectic (59:41 mol\%) with dissolved actinide chlorides (AnCl$_{3}$) at 773 K. Spent metallic fuels are dissolved in the molten salt at the anode, and actinide metals are deposited at the cathode. The concentration of AnCl$_{3}$ in LiCl-KCl eutectic is assumed to be 2 mol\%. However, the concentration of AnCl$_{3}$ becomes higher in the vicinity of the anode during electrorefining, and the solubility of AnCl$_{3}$ may affect the rate of anodic dissolution. Therefore, it is important for the process operation to clarify the phase diagrams of LiCl-KCl-AnCl$_{3}$ systems. In the present study, LiCl-KCl-UCl$_{3}$ ternary phase diagram in the range of 0$\sim$20 mol\% UCl$_{3}$ is investigated by a combination of X-ray diffraction method (XRD) and differential thermal analysis (DTA).

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Event date： 2012.9

Language：Japanese  

Country：Japan  

In Fukushima Dai-ichi Nuclear Power Plant accident, significant amount of sea water was introduced in reactor pressure vessel and it possibly caused various chemical reaction with degraded fuels. Since B$_{4}$C control rod was used in Fukushima Dai-ichi reactors, the chemical behavior between fuel and control rod is different from TMI-accident. Thermodynamic evaluation was performed to estimate tendency of these reactions. As for the reaction between degraded fuel and sea water, the chemical forms of Cs, Sr, Mo, Ru and Te are changed and the volatility is increased, in which various impurities contained in sea water play important roles. Regarding the chemical reaction between degraded fuel and B, valiation in oxygen potential in degraded core becomes far complicated in case of BWR. Especially, formation of intermetallic compounds among U, Pu and Zr with B and its oxidation plays important roles for the redistribution of B. Regarding the corrosive gases, the formation of HCl or H$_{2}$S becomes easier even in lower temperature when heating sea water in higher oxygen potential condition.

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Event date： 2012.9

Language：Japanese  

Country：Japan  

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Event date： 2012.8

Language：Japanese  

Country：Japan  

Regarding fuel degradation in a severe accident, role of thermodynamic database and kinetic modeling is introduced, and some examples are also introduced as well as the related international collaboration.

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Event date： 2012.6

Language：English  

Venue：Berkeley  

Thermodynamic database and kinetic modelling will play important roles not only in the recovery process of Fukushima Dai-ichi Nuclear Power Plant, but also in the modification process of accident management. Model evaluation can help to image rough views of accident sequences happened under "black-out". Database on high temperature corium, reaction between corium and B$_{4}$C control rod, effect of sea-water, and molten corium concrete interaction are especially necessary. In case of accident management, the most important issue is to obtain consensuses on "How reactor cores are degraded?" among various kinds of specialists. Visualization based on dedicated modelling acts as an useful tool to relate various sciences. Recent progress on database and modelling is discussed.

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Event date： 2012.3

Language：Japanese  

Country：Japan  

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Event date： 2012.3

Language：Japanese  

Country：Japan  

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Event date： 2011.9

Language：Japanese  

Country：Japan  

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Event date： 2011.3

Language：Japanese  

Country：Japan  

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Event date： 2011.3

Language：Japanese  

Country：Japan  

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Event date： 2010.3

Language：Japanese  

Country：Japan  

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Event date： 2007.9

Language：Japanese  

Country：Japan  

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