Authorship：Lead author,　Last author,　Corresponding author   Language：English   Publishing type：Research paper (scientific journal)   Publisher：TAYLOR & FRANCIS INC  

The author previously proposed that glassy cesium-bearing microparticles [resulting uniquely from the Fukushima Daiichi nuclear power station (FDNPS) accident] may have been formed by melting and atomization of glass fibers (GFs) of the high-efficiency particulate air filter in the standby gas treatment system line due to the flame and blast during the hydrogen explosion in Unit 3. Assuming that this hypothesis is correct, Type A could contain or accompany carbon, which ignites spontaneously above 623 K, because of the limited time to be heated up, the inclusion of carbon in the binder applied on the GF surface, and the closely located charcoal filter. As previous studies have not identified carbon, the present analyses were performed with an electron probe microanalyzer to determine whether Type A contains carbon. The results show that Type A contained carbon originating from the binder. Some nonspherical particles were accompanied by Type A, and the film surrounding Type A contained more carbon, which is thought to originate from the charcoal filter. These results cannot be explained by the other mechanisms proposed so far and can be explained consistently only by the author’s proposed hypothesis. Although it may be premature to determine Type A formation mechanisms, this information enables one to limit the temperature conditions of Type A formation.

DOI： 10.1080/00295450.2021.1929767

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DOI： 10.1016/j.jnucmat.2019.151819

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© 2018, © 2018 American Nuclear Society. A simulation model was developed to estimate the areal (surface) deposition pattern of 129mTe after the Fukushima Daiichi nuclear power plant (FDNPP) accident. Using this model, the timing and intensity of the 129mTe release were reverse estimated from the environmental monitoring data. Validation using 137Cs data showed that the model simulated atmospheric dispersion and estimated surface deposition with relatively high accuracy. The estimated surface deposition pattern of 129mTe was consistent with the actual measured pattern. The estimated time and activity of 129mTe emissions indicated that 129mTe was predominantly emitted from FDNPP Unit 3.

DOI： 10.1080/00295450.2018.1521186

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Authorship：Lead author,　Last author,　Corresponding author   Language：English   Publishing type：Research paper (scientific journal)   Publisher：Atomic Energy Society, Japan, Journal of Nuclear Science and technology  

Insoluble Cs particles (Type A) were first observed in Tsukuba city in the morning of
15 March 2011. The particles were considered to have been formed in the reactor pressure vessel of Unit 2 by evaporation/condensation based on the measured 134Cs/137Cs ratio and the accident progression of each Unit. However, the particles were covered by almost pure silicate glass and showed a trace of quenching. This can be explained by other mechanisms, that is, the particles were formed by the melting of glass fibers of the high efficiency particulate air filter in the Standby Gas Treatment System owing to the fire due to hydrogen detonation in Unit 3 at 11:01 on March 14 and atomization due to the explosion, followed by quenching of the molten materials. Although the particles formed in this way were mostly dispersed to the sea by wind at that time, some of them were deposited on the lower elevation of the reactor building of Unit 3, and they could have been resuspended and released into the environment, by the flow owing to the generation of a large amount of steam as result of a restart of core cooling water injection at 02:30 on March 15.

DOI： 10.1080/00223131.2019.1583611

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Other Link： http://orcid.org/0000-0003-3342-3571

Authorship：Lead author,　Corresponding author   Language：English   Publishing type：Research paper (scientific journal)   Publisher：TAYLOR & FRANCIS LTD  

To investigate what happened in reality during the Fukushima Daiichi Nuclear Power Plant accident, the phenomena within reactor pressure vessel and the discussion of ties with the environmental monitoring measurement are very important. However, the previous study that treats phenomena of the both has not necessarily advanced up to the present time. The source terms predicted by simulation codes such as MELCOR has not yet been consistent with the reverse estimation by WSPEEDI code using environmental measurement data. This study investigated I-131 and Cs-137 release behaviors during the late phase of the accident to contribute to such examination using the I-131/Cs-137 ratio of the new source terms predicted by Katata. The I-131 release by the gas-liquid partition from the contaminated water in the 1F2 and 1F3 reactor buildings which was pointed out in the previous study was reevaluated using the new source terms. In addition, paying attention to the similarity of the core conditions between the Fukushima accident and the Phebus FPT3 experiment using the B4C control rods, the release of organic iodine (CH3I) during the 1F3 suppression pool venting, formation of CsBO2 and its release behavior were examined which have not yet been sufficiently studied so far.

DOI： 10.1080/00223131.2017.1323691

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Research Expert Committee on Evaluation of Severe Accident of AESJ (Atomic Energy Society of Japan) has developed thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and source term (ST) PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aim to explore the debris distribution and the current condition in the NPPs with high accuracy and to extract higher priority from the aspect of the sophistication of the analytical technology to predict the severe accident phenomena by the analytical codes. The ST PIRT is divided into 3 phases for time domain and 9 categories for spatial domain. The 68 phenomena have been extracted and the importance from the viewpoint of the source term has been ranked through brainstorming and discussions among experts. The present paper describes the developed ST PIRT list and summarizes the high ranked phenomena in each phase. (C) 2015 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.nucengdes.2015.02.005

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Authorship：Lead author,　Last author,　Corresponding author   Language：Japanese   Publishing type：Research paper (scientific journal)   Publisher：Atomic Energy Society of Japan  

Boron carbide (B4C) used for BWR or EPR absorbers could cause phenomena that never occur in PWR with silver-indium-cadmium absorbers during a severe accident. B4C would undergo a eutectic interaction with stainless steel and enhance core melt relocation. Boron oxidation could increase H2 generation, and the change of liberated carbon to CH4 could enhance the generation of organic iodide vCH3I). HBO2 generated during B4C oxidation could be changed to cesium borate (CsBO2) by combining it with cesium. This may increase cesium deposition into the reactor coolant system. There could be differences in the configuration, surface area, and stainless-steel to B4C weight ratio between the B4C powder absorber and pellet absorber. The present task is to clarify the effect of these differences on melt progression, B4C oxidation, and the iodine or cesium source term. Advancement of this research field could contribute to further sophistication of prediction tools for melt progression and source terms of the Fukushima accident, and the treatment of organic iodide formation in safety evaluation.

DOI： 10.3327/taesj.J14.021

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Authorship：Lead author   Language：English   Publishing type：Research paper (scientific journal)   Publisher：TAYLOR & FRANCIS LTD  

During the process of core cooling at Fukushima Daiichi nuclear power plants accident, large amount of contaminated water was accumulated in the basements of the reactor buildings at Units 1-4. The present study estimated the quantities of I-131 and Cs-137 in the water as of late March based on the press-opened data. The estimated ratios of I-131 and Cs-137 quantities to the core inventories are 0.51%, 0.85% at Unit 1, 74%, 38% at Unit 2 and 26%, 18% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere due to gas-liquid partition and contribute to increase in the release to environment. A lot of evaluations for I-131 release have been performed so far by the MELCOR calculation or the SPEEDI reverse estimation. The SPEEDI reverse predicted significant release until 26 March, while no prediction in MELCOR after 17 March. The present study showed that iodine release from accumulated water may explain the release between 17 and 26 March. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks for several days after the core melt.

DOI： 10.1080/00223131.2014.881725

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Authorship：Lead author,　Last author,　Corresponding author   Language：English   Publishing type：Research paper (scientific journal)   Publisher：Informa UK Limited  

In the VEGA program on radionuclide release from irradiated fuel under severe accident conditions, 10 tests in total were performed at JAEA from 1999 to 2004 under inert and steam atmospheres including the highest pressure or temperature conditions. These tests showed the increase in release rate above 2,800K or at the fuel liquefaction and the decrease in release rate under elevated pressure, which was a first observation in the world. The data on low-volatility radionuclide release, release from MOX fuel, effect of fuel oxidation, and eutectic reaction with cladding on release were obtained from the tests. The mechanism of pressure effect on release was examined and a new release model with pressure effect was proposed. In addition, the pressure effect on source term evaluation and effectiveness of accident management measures were investigated. This article summarizes the major outcomes described above that have already been published and newly describes the validation of the proposed release model together with limitations of the VEGA program and future issues. © Atomic Energy Society of Japan.

DOI： 10.3327/jnst.48.85

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The radionuclide release from mixed-oxide fuel (MOX) under severe accident conditions was investigated in the VEGA program to provide the technical bases for safety evaluation including probabilistic safety assessment (PSA) for light water reactor (LWR) using MOX. The MOX specimens irradiated at Advanced Thermal Reactor (ATR) Fugen were heated up to 3,123K in helium at 0.1 and 1.0MPa. The release of volatile fission products (FP) was slightly enhanced below 1,623 K compared with that of UO2. The volatile FP release at elevated pressure was decreased as in the case with UO2. The total fractional release of Cs reached almost 100% while almost no release of low-volatile FP even after the fuel melting. The release rate of plutonium above 2,800 K increased rapidly although the amount was small. Since the existing models cannot predict this increase, an empirical model was prepared based on the data. The present study showed that there are no large differences in total fractional releases and inventories of important FP in PSA between UO2 and MOX. This suggests that the consequences of LWR using MOX are mostly equal to those using UO2 from a view point of risks. © 2005 Taylor &amp
Francis Group, Ltd.

DOI： 10.1080/18811248.2005.9726413

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Authorship：Lead author,　Corresponding author   Language：English   Publishing type：Research paper (scientific journal)  

The VEGA tests on radionuclides release from fuel under severe accident conditions showed that the Cs release rate at 1.0 MPa decreased by about 30% compared with that at 0.1 MPa. To explain this pressure effect, a numerical release model that considers the lattice diffusion in grains followed by the gaseous diffusion in open pores was developed. However, this model is not practical for the PSA analyses due to much computation time and therefore a simplified model called CORSOR-M with the release rate coefficient multiplied by (P≧1 atm) was derived from the numerical model. The multiplier comes from the pressure dependency of gaseous diffusion flux in pores at the pellet surface. The effect of pressure on source term was also estimated for a transient sequence at BWR with JAERI's THALES-2 code in which the simplified model was incorporated. Since the adequacy and applicability of CORSOR-M model were confirmed for the pressures up to 16 MPa through comparison with the VEGA tests and mechanistic models, it is proposed that the model be used for the source term analyses. © 2004 Taylor &amp
Francis Group, LLC.

DOI： 10.1080/18811248.2004.9726348

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Authorship：Lead author,　Corresponding author   Language：English   Publishing type：Research paper (scientific journal)  

In the case of severe accidents, the radionuclides release from fuel could mostly occur at high temperature under elevated pressure. The effect of temperature on the release has been clarified in many previous studies while the pressure influence has been scarcely investigated so far due to difficulty in the experimental operation. To investigate the effect of pressure on the release, two tests under the same conditions except for the system pressure were performed in the VEGA program at JAERI by heating up the irradiated UO2fuels up to 2,773 K in inert helium. The test results uniquely showed that the release rate of cesium for the temperatures below 2,773 K at 1.0 MPa could be suppressed by about 30% compared with that at 0.1 MPa. This article describes the outlines of the two tests and the observed effects of system pressure on cesium release as well as the results of various post-irradiation examinations. Moreover, the mechanisms and models that explain the pressure effect are discussed. © 2002 Taylor and Francis Group, LTD.

DOI： 10.1080/18811248.2002.9715258

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DOI： 10.1080/18811248.2002.9715185

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DOI： 10.1080/18811248.2001.9715115

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Language：English   Publishing type：Research paper (scientific journal)   Publisher：American Nuclear Society  

Aerosol revaporization in piping is being investigated in the WIND project at the Japan Atomic Energy Research Institute. The objectives of this study are to characterize the aerosol revaporization from piping surfaces under various thermal-hydraulic conditions and to obtain insights applicable to the validation of analytical models. Cesium iodide aerosol was introduced into the test section with a carrier gas. After quantifying the deposited mass of cesium and iodine, the test section was reheated to realize the revaporization. The revaporized materials were deposited onto another test section with an axial temperature gradient located downstream. Two runs (WAV1 and WAV2) were conducted. In WAV2, the influence of metaboric acid was examined. Most of the deposited cesium and iodine in the test section was revaporized and transported downstream. In WAV2, deposition density of cesium was much larger than that of iodine. It was supposed that a part of the cesium iodide that was deposited in the upstream test section reacted with boric oxide to form cesium metaborate.

DOI： 10.13182/NT01-A3186

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Authorship：Lead author,　Corresponding author   Language：English   Publishing type：Research paper (scientific journal)   Publisher：PERGAMON-ELSEVIER SCIENCE LTD  

A relatively small-scale aerosol deposition experiment (called WAVE) in a quartz glass pipe with a 90 degrees bend followed by a short straight section was performed at Japan Atomic Energy Research Institute to investigate the effect of pipe orientation on the cesium iodide (CsI) aerosol deposition. In these basic configurations, the section after the bend was either horizontal, upward and downward, respectively. The upward case showed 5 and 10 times larger CsI mass deposition in the vertical section of pipe than the horizontal and downward configurations, respectively. The experiments were analyzed by coupling a three-dimensional fluid-dynamic and an aerosol behavior codes. The calculations were in reasonable agreement with the measured aerosol mass deposition except for the upward case because the principal CsI deposition mechanism is thermophoresis which depends on the thermal gradient in gas and the gradient was well predicted by the fluid-dynamic code. In order to better predict the deposited mass in vertical section of upward case, Nusselt number which is used for thermophoretic deposition calculation had to be reevaluated carefully by considering the effect of secondary free convection in the vertical pipe which flowed opposite to the main stream. (C) 2000 Elsevier Science Ltd. All rights reserved.

DOI： 10.1016/S0021-8502(00)00038-0

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It has been pointed out that the reactor coolant system piping could fail prior to the melthrough of the reactor pressure vessel in a high pressure sequence of pressurized water reactor severe accidents. In order to apply to the evaluation of the piping failure which influences the subsequent accident progression, models for the strength of piping materials at high temperatures were examined. It was found that 0.2% proof stress and ultimate tensile strength above 1,073K obtained from tensile tests was reproduced by a quadratic equation of the reciprocal absolute temperature. Short-term creep rupture time and minimum creep rate at high temperatures were well correlated by the modified Norton's Law as a function of stress and temperature, which implicitly expressed the effect of the precipitation and the resolution of precipitates on the creep strength. The modified Norton's Law gave better results than the conventional Larson-Miller method. Relating applied stress vs. minimum creep rate and tensile properties vs. applied strain rate obtained from the creep and tensile tests, a temperature range where the dynamic recrystallization significantly occurred was evaluated.

DOI： 10.1080/18811248.2000.9714925

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Language：English   Publishing type：Research paper (scientific journal)   Publisher：Elsevier Sequoia SA  

The tests on fission product (FP) behavior in piping under severe accidents are being conducted in the wide range piping integrity demonstration (WIND) project at JAERI to investigate the piping integrity which may be threatened by decay heat from deposited FPs. In order to obtain the background information for future WIND experiment and to confirm analytical capabilities of the FP aerosol analysis codes, ART and VICTORIA, the FP behavior in safety relief valve (SRV) line of BWR during TQUX sequence was analyzed. The analyses showed that the mechanisms that control the FP deposition and transport agreed well between the two codes. However, the differences in models such as diffusiophoresis or turbulence, the treatment of chemical forms and aerosol mass distribution could affect the deposition in piping and, consequently, on the source terms. The WIND experimental analyses were also conducted with a three-dimensional fluiddynamic WINDFLOW, ART and an interface module to appropriately couple the fluiddynamics and FP behavior analyses. The analyses showed that the major deposition mechanism for cesium iodide (CsI) is thermophoresis which depends on the thermal gradient in gas. Accordingly, the coupling analyses were found to be essential to accurately predict the CsI deposition in piping, to which little attention has been paid in the previous studies.

DOI： 10.1016/S0029-5493(99)00328-3

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In a severe accident of light water reactors, the reactor coolant system (RCS) piping might be subjected to thermal loads caused by the decay heat of the deposited fission products and the heat transfer from the hot gases, with an internal pressure in some accident sequences. Tests on the RCS piping failure were performed along with high temperature tensile and creep rupture tests including metallography to investigate the failure behavior. The prediction of the 0.2% proof stress by Arrhenius equation is in good agreement with the measured stress above 800 degrees C for served RCS piping materials. The modified Norton's Law for the short term creep rupture model agrees with the experimental values between 800 and 1,150 degrees C for type 316 stainless steel. The microstructural change was discussed with the effect of the very rapid formation and resolution of the precipitation on the strength at high temperature. The result of the piping failure tests which simulated the severe accident conditions, i.e.: in short-term at high-temperature, could support the plastic limit load prediction of the flow stress model using the 0.2% roof stress.

DOI： 10.1080/18811248.1999.9726282

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The aerosol deposition test series is being performed to investigate the deposition of FP vapor and aerosol onto the inner surface of reactor coolant piping during a severe accident of a light water reactor. Vapor and aerosol of CsI as an FP simulant was introduced into the horizontal test section pipes. No substantial decomposition of CsI was identified to occur both in the high temperature inert and superheated steam environments. The comparison between the results of the aerosol deposition test series and the thermo-fluiddynamic analysis with WINDFLOW implied that a profile of the CsI deposition due to vapor condensation and thermophoretic aerosol deposition was influenced hy local thermo-fluiddynamic conditions. Deposition velocities were evaluated for CsI vapor and aerosol based on the deposition characteristics of CsI and the thermo-fluiddynamic analysis.

DOI： 10.1080/18811248.1999.9726226

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Language：English   Publishing type：Research paper (scientific journal)   Publisher：ELSEVIER SCIENCE BV  

The fission products (FP) behavior analysis code ART developed at JAERI simulates a FP transport and deposition in a reactor coolant system and containment during severe accidents. As part of the code development and verification, several experimental analyses have been conducted. In the JAERI's WAVE experiment, the effect of nitrogen or steam carrier gases on the cesium iodide (CsI) behavior in piping was recently investigated. The ART analysis for nitrogen agreed with the experimental results by reflecting the detailed thermo-fluiddynamic calculation on the CsI aerosol behavior analysis. On the contrary, the analysis for steam did not agree well with the experimental results because observed enhancement of aerosol growth cannot be explained by existing models. Moreover, the newly developed empirical models on iodine chemistry in water were examined for the ACE/RTF 3B experiment. The analysis showed that those models have a fundamental analytical capability. (C) 1997 Elsevier Science B.V.

DOI： 10.1016/S0022-3115(97)00207-9

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Language：Japanese   Publisher：ATOMIC ENERGY SOC JAPAN  

Research on severe accident for light water reactors has been initiated after the accident of Three Mile Island in 1979 and accelerated all over the world after the Chernobyl accident in 1986. For the severe accident management Nuclear Safety Commission of Japan recommended electric utilities to investigate the possible measures in 1992. In response to this they proposed accident management strategies based on their investigations in 1994 to be voluntarily implemented in several years by reflecting findings from severe accident research. Since the severe accident phenomena largely involve the thermal-hydraulic aspects, it is of great importance to evaluate the thermal-hydraulic behaviors of severe accident in order to reduce the associated uncertainties. In the present article thermal-hydraulic behaviors during severe accident are described in terms of phenomenology, severe accident management, key issues, research methodology and remaining research items. The main phenomena include core melt progression, in-vessel core melt behaviors, molten core coolant interaction, molten core concrete interaction, direct containment heating, hydrogen behaviors, and fission product behaviors.

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In the FPT0 test of the PHEBUS/FP program, it was observed that the fraction of liquefied UO2 reached 50%, which is much larger than the expected maximum value of 20%. Most of the post-test analyses with various computer codes underpredicted the bundle temperature during a late phase and could not reproduce such a large core degradation. In most of the previous analyses, the shroud thermal conductivity evaluated based on the Pears' ZrO2 specific heat data and the thermal diffusivity measured by JAERI was used. However, recent thermal properly data books adopt a lower specific heat than measured by Coughlin and King's at high temperature. The present analyses with ICARE2 showed that the FPT0 bundle behavior could be mostly reproduced by using the shroud thermal conductivity based on Coughlin and King's. If the present calculation is assumed to be correct enough, the shroud thermal conductivity at high temperature could be smaller than the current evaluation based on the Pears' data. Since the shroud thermal conductivity has thus a strong effect on the bundle behavior, further measurement and evaluation of the thermal properties of the shroud are highly recommended. (C) 1997 Elsevier Science S.A.

DOI： 10.1016/S0029-5493(96)00009-X

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DOI： 10.3327/jaesj.39.123

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DOI： 10.1080/18811248.1996.9732006

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Language：English   Publishing type：Research paper (scientific journal)   Publisher：ELSEVIER SCIENCE SA LAUSANNE  

Two series of experiments to investigate melt-coolant interactions have been performed as part of the ALPHA program at JAERI. In the melt drop steam explosion experiments, melt simulating a molten core was dropped into a pool of water. Volume fractions of the melt, water and steam in the mixing region prior to the occurrence of spontaneous steam explosions were quantified. Other characteristics of melt-coolant interactions were evaluated for settling velocity of the melt in water, propagation and expansion velocities, energy conversion ratio and debris size distribution. It was found that the probability of the occurrence of spontaneous steam explosions could be reduced by using a melt dispersion device. Measurement of void fraction in the mixing region clearly showed that the melt dispersion device enhanced steam generation. However, one experiment indicated that the use of the dispersion device could possibly result in a more energetic steam explosion. It was found that the mixing of non-condensable gas in the steam phase of the mixing region during melt dispersion played an important role for the suppression of the spontaneous steam explosion. Knowledge of the parametric effects of melt mass, ambient pressure and water temperature was extended. In the melt coolability experiments, water was poured onto the melt to investigate melt-coolant interactions in a stratified geometry where water overlies on a melt layer. Melt eruptions which could induce an explosive interaction were observed when the subcooled water was poured through a pipe nozzle. The eruption was not observed when the water was near the saturation temperature or supplied through a spray nozzle. The explosive interaction in the stratified geometry was found to be much smaller in magnitude than the steam explosion in the melt drop configuration.

DOI： 10.1016/0029-5493(94)00883-Z

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During a station blackout of PWR, the pump seal will fail due to loss of the seal cooling. This particular transient-LOCA sequence designated as S3-TMLB’ analyzed by SNL with MELPROG/TRAC for Surry plant showed that the depressurization due to the pump seal LOCA would result in early accumulator injection and subsequent core cooling which lead to the delay of reactor pressure vessel (RPV) meltthrough. The present analysis was performed with SCDAP/RELAP5 to evaluate this scenario shown in the MELPROG/TRAC analyses. Additionally, the calculated results were compared with the similar experimental studies of JAERI’s ROSA-IV program. The present analyses showed that: (1) During S3-TMLB’, the loop seal clearing would occur and cause a slight delay of accident progression. (2) It is unlikely that the accumulator injection, which leads to the delay of RPV meltthrough by approximately 60 min, is initiated automatically during S3-TMLB’. Accordingly, an intentional depressurization using PORVs is recommended for the mitigation of the accident consequences. (3) The present SCDAP/RELAP5 analyses did not show significant delay of accident progression. It was found that non-realistic lower heat generation and higher core cooling models used in the MELPROG/TRAC analysis are attributed to this discrepancy. © 1995 Taylor &amp
Francis Group, LLC.

DOI： 10.1080/18811248.1995.9731740

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The WAVE experiments have been performed at JAERI to investigate the CsI deposition onto the inner surface of pipe wall under typical severe accident conditions. It was shown that relatively large amount of CsI was deposited at the upstream floor of the pipe and that larger amount of CsI was deposited on the ceiling than the floor at the downstream. Analyses of the experiments have also been conducted with the three-dimensional thermohydraulic code, SPRAC, and the radionuclide transport analysis code, ART. The experimental results were well reproduced with ART by using peripherally subdivided pipe cross section and associated representative thermohydraulic information from SPRAC prediction. It was clarified through the present experiment and analyses that major deposition mechanisms for the chemical form of CsI are thermophoresis and condensation. Accordingly, the coupling of the FP behavior and the detailed thermohydraulic analyses was found to be essential in order to accurately predict the CsI deposition in the pipe, to which little attention has been paid in the previous studies. © 1995, Atomic Energy Society of Japan. All rights reserved.

DOI： 10.3327/jnst.32.1047

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A small-scale penetration leak characterization test has been performed as a part of the ALPHA program at Japan Atomic Energy Research Institute (JAERI). Two series of experiments were performed using test sections which simulate relevant parts of an EPA (Electrical Penetration Assembly) used in Japanese PWR containments. One of the test sections simulates an alumina module and the other includes the silicone resin portion of the EPA. The test section was heated in a leak test vessel which simulated thermal-hydraulic conditions inside and outside of the containment in a severe accident. From the experimental results, it was concluded that although the silicone resin may melt at high temperature, the alumina module will remain intact under severe accident conditions. The EPA as a whole is estimated to maintain leak-tightness during a severe accident. It was found in the experiments that heat conduction along the metal portion of the test section had a strong influence on the melt progression of the resin. It was also found that the measured strain of the alumina module was predominantly caused by the elevated temperature. Therefore, the thermal load will be more of a threat to the EPA's integrity rather than the pressure load. © 1993.

DOI： 10.1016/0029-5493(93)90246-6

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Language：Japanese   Publishing type：Research paper (scientific journal)   Publisher：ATOMIC ENERGY SOC JAPAN  

A simple method that uses effective energy and average energy has been used for calculating gamma doses from radioactive plumes. Although the additive method, in which the contributions to gamma doseS due to the photons released from a noble nuclide are added together, should be used for this purpose, the computation is very time-consuming, so that a simplemethod is desirable in emergencies. The relationship between the calculated doses and the photon energy is complicated and no linear expression holds good. Thus, when the photon energy is widely dispersed, the simple method is not appropriate, We cornpared the results of the simple method with those of the additive method, and examined the applications of both am the limitations of the simple method. It was found that for most nuclides the simple method gave overestimates of within 20% for doses on downwind axis, and for a few nuclides, under. estimates by more than 90% were seen for doses apart from the downwind axis near a stack. We suggest a method of correction in which the photons released from a nuclide are grouped into two categories: those with an energy of less than 50keV and those with a higher energy. When the sum of the doses due to these two groups of photons was obtained, the errors were within 5%.

DOI： 10.3327/jaesj.29.1023

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The external exposure to gamma rays from a radioactive cloud, <i>i. e.</i> the cloudshine, is an important exposure pathway. A computer code, AIRGAMMA, was developed to calculate quickly the cloudshine doses by interpolating normalized doses provided on the basis of the Gaussian plume model. The normalized dose is defined as the dose on the ground when a gamma-emitting nuclide is released at a rate of 1Ci/hr or the amount of 1Ci and the wind speed is 1m/sec. The code approximately takes account of depletion of the dispersing materials for calculation of the cloudshine doses, but ignores effects of the building wake and the mixing layer on the cloudshine doses. The interpolation of the normalized dose would not cause errors larger than 1% for most of the cases. The time for calculation of a dose is less than 2 msec on the computer FACOM-M380. The applicability of the approximations mentioned above is discussed in this report.

DOI： 10.5453/jhps.20.33

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DOI： 10.11484/jaea-review-2016-011

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DOI： 10.11484/jaea-data-code-2015-001

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Copyright © 2015 by JSME. Nuclear Human Resource Development Center (NuHRDeC) of JAEA has conducted nuclear human resource development for more than 50 years since its establishment in 1958. NuHRDeC conducts international nuclear human resource development, so called "Instructor Training Program (ITP)", which is a training scheme launched in 1996 in order to support Asian countries seeking peaceful use of nuclear energy. The ITP consists of 1) Instructor Training Course (ITC) in Japan, 2) Follow-up Training Course (FTC) in own countries organized by instructors trained at ITC in Japan, and 3) Nuclear Technology Seminar for bringing up nuclear trainers and leaders in Asian countries. The purpose of ITP is to develop a self-sustainable training system in Asian countries, which disseminates the knowledge and technology in their countries. After completing ITC trainings at NuHRDeC, the trainees are obliged to set up FTC in each country. They create own 1 or 2 weeks course curricula and allocate local lecturers including themselves. Two or three Japanese experts join the FTC to give technical advices and support to local lecturers. The present specialized fields of ITC are 1) Reactor engineering such as reactor physics, thermal engineering and reactor safety, 2) Environmental radioactivity monitoring, and 3) Nuclear emergency preparedness. The main feature of ITC is that the curricula places emphasis on the practical exercise with well-equipped training facilities, experimental laboratories utilizing the simulators of research reactor, and the expertise of lecturers mostly from JAEA. As of FY2014, ITC is applied to 8 countries; Indonesia, Thailand, Vietnam, Bangladesh, Kazakhstan, Malaysia, Philippines and Mongolia. The total number of participants at ITC since 1996 is approximately 300 and the participation of FTC has been increased significantly year after year with more than 3,000 in total. This result indicates that the ITP system has been effectively contributed to fostering local trainers in Asian counties. Present paper summarizes the outlines, experiences and future prospects of ITP.

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During core cooling at Fukushima Daiichi NPP accident, large amount of contaminated water was accumulated in the basements of reactor buildings at Units 1 to 4. The estimated ratios of I-131 and Cs-137 quantities in water to the core inventories are 0.51\%, 0.85\% at Unit 1, 74\%, 38\% at Unit 2 and 26\%, 18\% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere. Many evaluations for I-131 release have been performed so far by MELCOR or the reverse estimation with SPEEDI. The SPEEDI reverse predicted significant release until March 26 while no prediction in MELCOR after March 17. The present study showed that iodine release from accumulated water due to radiolytic conversion from I$^{-}$ to I$_{2}$ and gas-liquid partition of I$_{2}$ may explain the release between March 17 and 26. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks.

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NRC recently prepared the draft of revised NUREG-1465 for high burnup or MOX fuel and is peer-reviewing it by US, French specialists including myself. In the draft, the containment source terms were evaluated by the MELCOR code with implementation of the Booth model adjusted based on the IRSN's VERCORS test results. However, only the diffusion coefficient for Cs was changed and the original class scale factor of the ORNL-Booth model was used and therefore there was no large difference in the results between the newly revised and existing NUREG-1465. It was proposed that the class scale factor be reevaluated so that the final release rates of radionuclide other than Cs in the VERCORS tests might correspond to the calculation by the Booth model with change of only the diffusion coefficient for Cs.

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In VEGA program on FP release from fuel during severe accidents, 10 tests were conducted under the highest pressure and/or temperature conditions. Tests with PWR fuel at 1.0MPa showed first that Cs release was suppressed by about 30\% compared with that at 0.1MPa. This was reproduced by 2-stage diffusion model in UO$_{2}$ grains \& pores, and a simplified 1/P**0.5 CORSOR-M model. In BWR and MOX fuel tests, however, this effect was not observed clearly due to higher fuel temperature during normal operation and differences in test conditions. The pressure effect may affect PWR source terms and AM measures such as intentional depressurization. Analyses with THALES-2 suggested that the depressurization has many advantages such as delay in accident progression and mitigation of source terms at early CV failure despite increase in FP release into RCS. The effect of pressure on consequences needs to be evaluated systematically for various accident sequences with AM measures.

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In VEGA program on radionuclide release from fuel under severe accident condition, totally 10 tests were performed under inert and steam atmospheres including the highest pressure or temperature conditions from 1999 to 2004. These tests showed that increase in release rate at high temperature around UO$_{2}$ melting point and decrease in release rate under elevated pressure that was a first observation in the world. The data on low-volatile radionuclide release, release from MOX fuel, the effect of fuel oxidation and eutectic reaction with cladding on radionuclide release were obtained through the tests. In addition, the effect of obtained results on the source term evaluation and effectiveness of the accident management measures were examined. This manuscript summarizes the above described outcomes of VEGA program that have been already submitted to academic societies or international conferences.

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The TMI accident in 1979 motivated an interest in LWR source terms and resulted in the production of a supplement to the first state of the art report (SOAR) which concentrated on LWR aerosol issues. The second SOAR dealt with primary system FP release and transport that covers vapor the condensation on aerosols and aerosol agglomeration. The present third SOAR was prepared focusing on aerosol behavior in both the primary circuit and in containment such as mechanical resuspension, impact of chemistry, re-vaporization of deposits, charge effect, removal by spray, hydrogen-burn effects on suspended aerosols, penetration of aerosols through leak paths and so on. A large number of probabilistic safety analysis (PSA level 2) plant studies have been performed around the world, frequently involving aspects of aerosol behavior. This report provides some examples, including sensitivity studies that demonstrate the impact of aerosol-related processes.

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Language：English   Publisher：AMER SOC MECHANICAL ENGINEERS  

Based on the basic policy for introduction of Risk Informed Regulation (RIR) expressed by the Nuclear Safety Commission (NSC) of Japan in 2003, NSC set up the taskforce in April 2004 to take an initiative for coordinated collaboration of related organizations for developing scheme of RIR. Since then the taskforce has reviewed the efforts of related organizations and discussed the issues on further utilization of risk information in Japan. According to the final report prepared by the taskforce in September 2007, the risk consideration in related organizations has made a progress in line mostly with the NSC's basic policy. For example, the regulatory guide for seismic design was revised in 2006 including combination of deterministic and probabilistic approaches. The regulatory body will start a new inspection system in 2008 that considers the risk informed safety classification of structures, systems and components. However, the followings were identified as important issues in the future: i) promotion of advanced and preliminary trials, ii) comprehensive applications using PSA results for both of the representative and individual plants, iii) preparation of acceptance guidelines for safety and performance goals in risk informed decision making, iv) improvement of PSA quality, v) revision of regulatory guides considering risk information, vi) comprehensive evaluation and promotion by NSC for further utilization -of risk information, vii) enhancement of infrastructure such as PSA experts and database, viii) promotion of safety research, ix) application to seismic design, x) Introduction of risk informed approaches into nuclear fuel cycle facilities, and xi) promotion of risk communication, etc.

DOI： 10.1115/ICONE16-48569

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Language：Japanese   Publisher：ATOMIC ENERGY SOC JAPAN  

DOI： 10.3327/jaesj.48.242

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The Nuclear Safety Commission (NSC), Japan set up the taskforce on introduction of risk informed regulation (RIR) into nuclear safety regulations in April 2004. Since then the taskforce (Chairperson: Prof. Genki Yagawa, Toyo university) has reviewed the status of risk considerations at related organizations and discussed the issues for developing RIR in Japan. Recently, the taskforce prepared the interim report on the review results and discussions, and NSC approved it in December 2005. The report described that the risk consideration in related organizations in Japan has made progress in line mostly with the NSC's basic policy for introduction of RIR expressed in 2003. However, the following topics were identified as important issues for further promotion of RIR introduction: policy for utilization of risk information considering Japanese features, usage of safety goals and performance objectives in RIR, decision-making process using risk information, pilot program, PSA quality, improvement of safety examination guidelines considering risk information, utilization of risk information in nuclear fuel cycle facilities and risk communication. Copyright © 2006 by ASME.

DOI： 10.1115/ICONE14-89449

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Language：Japanese   Publisher：日本原子力学会  

DOI： 10.3327/jaesj.47.755

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DOI： 10.11484/jaeri-research-2005-001

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DOI： 10.11484/jaeri-tech-2004-050

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Authorship：Lead author   Language：Japanese   Publishing type：Lecture material (seminar, tutorial, course, lecture, etc.)  

File： JAERI-Conf-2003-014.pdf

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Language：Japanese   Publisher：ATOMIC ENERGY SOC JAPAN  

<p> 軽水炉シビアアクシデント時の炉心状態を模擬するため, 世界最高温度, 圧力条件を達成する実験装置VEGAを設計, 製作し, 雰囲気圧力が放射性物質の放出挙動に及ぼす影響を初めて実験的に定量化した。また, その結果に基づいて, 圧力の影響を説明するモデルを提案した。これらの成果に対し, 第35回 (平成14年度) 日本原子力学会賞論文賞が授与された。受賞対象となった研究は, 工藤 保氏, 中村武彦氏をはじめとする日本原子力研究所の共同研究者とともに行ったものであるが, 本稿では, 筆者が個人の立場で, 同研究との出会い, 思い入れ, 苦労した点, 今後の展開等について紹介する。</p>

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DOI： 10.11484/jaeri-tech-2003-009

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DOI： 10.11484/jaeri-research-2001-055

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DOI： 10.11484/jaeri-tech-2001-011

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DOI： 10.11484/jaeri-tech-2001-029

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The integrity of reactor cooling system piping during severe accidents of light water reactors is being investigated at Japan Atomic Energy Research Institute. The investigation is composed of piping failure tests, material property measurement and creep analyses. In the piping failure tests, piping made of stainless steels (SUS316 or CF8M), carbon steel or Inconel 690 is loaded with a constant internal pressure and an elevated temperature. Deformation features and time to failure have been obtained under various pressure and temperature conditions. The creep analyses were performed by ABAQUS code being furnished with creep equations based on the material property measurement. A three -dimensional analysis with shell elements for the test with small-diameter pipe (114.3 mm in an outer diameter) made of nuclear grade SUS316 significantly underes timated the deformation, resulting in a longer time to failure. Two -dimensional analyses with solid and shell elements indicated that this discrepancy would have resulted from the use of shell elements. A two-dimensional analysis with solid elements predicted well the time to failure of a thin steam generator U -tube (22.23 mm in an outer diameter) made of Inconel 690.

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DOI： 10.11484/jaeri-conf-2000-015

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The main goal of International Standard Problem (ISP) is to increase confidence in the validity and accuracy of the tools, which were used in assessing the safety of nuclear installations. Moreover, the exercises enable code users to gain experience and demonstrate their competence. The ISP No. 41 exercise, computer code exercise based on a Radioiodine Test Facility (RTF) experiment on iodine behavior in containment under severe accident conditions, is one of such ISP exercises. The codes used by the participants were LIRIC (AECL), MELCOR-I (SNL), IMPAIR (PSI, Siemens, GRS, and JAERI) and IODE (CIEMAT, IPSN and NRIR). This report presents a detailed description of the RTF tests used for the exercise, a brief description of the models/codes used and the modelling process and the description and interpretation of the results.

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DOI： 10.11484/jaeri-tech-2000-015

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Language：Japanese   Publisher：日刊工業出版プロダクション  

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The tests on fission product (FP) behavior in piping under severe accidents are being conducted in the wide range piping integrity demonstration (WIND) project at JAERI to investigate the piping integrity which may be threatened by decay heat from deposited FPs. In order to obtain the background information for future WIND experiment and to confirm analytical capabilities of the FP aerosol analysis codes, ART and VICTORIA, the FP behavior in safety relief valve (SRV) line of BWR during TQUX sequence was analyzed. The analyses showed that the mechanisms that control the FP deposition and transport agreed well between the two codes. However, the differences in models such as diffusiophoresis or turbulence, the treatment of chemical forms and aerosol mass distribution could affect the deposition in piping and, consequently, on the source terms. The WIND experimental analyses were also conducted with a three-dimensional fluiddynamic WINDFLOW, ART and an interface module to appropriately couple the fluiddynamics and FP behavior analyses. The analyses showed that the major deposition mechanism for cesium iodide (CsI) is thermophoresis which depends on the thermal gradient in gas. Accordingly, the coupling analyses were found to be essential to accurately predict the CsI deposition in piping, to which little attention has been paid in the previous studies.

DOI： 10.1016/S0029-5493(99)00328-3

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DOI： 10.11484/jaeri-research-99-067

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DOI： 10.11484/jaeri-research-99-066

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DOI： 10.11484/jaeri-tech-99-013

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DOI： 10.11484/jaeri-tech-99-036

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The ISP No.41 exercise resulted from a recommendation at the fourth iodine chemistry workshop held at PSI, Switzerland in 1996 for an International Standard Problem on iodine behavior models. The test selected for the comparison was a Radio Test Facility experiment. It was conducted in a stainless steel vessel at 25$^{\circ}$C and at a dose rate of 1.4 kGy/h$^{-1}$ to evaluate the effect of pH on irradiated aqueous solutions containing CsI. This paper discusses the results of ISP 41 exercise, with a primary focus on the evaluation and comparison of calculated results, and what they demonstrate about the aqueous iodine reaction subset within each model. The paper will also discuss the relative importance of mass transfer, surface adsorption and aqueous chemistry, and the sensitivity of each of the models to these phenomena. Finally, it will assess the applicability of the ISP 41 exercise to qualitative validation of the iodine models, and provide recommendations for continuing model evaluation.

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The WIND (Wide range piping INtegrity Demonstration) project is being performed at JAERI in order to demonstrate the integrity of LWRs piping under the severe accident conditions. The hot tensile tests of the piping materials made of stainless steel (type 316 and CF8M) or carbon steel (STS 410), have been performed in order to identify the high-temperature strength and ductility. The yield strength and the ultimate tensile strength above 800°C decreases in the order, type 316> CF8M>STS 410. The deformation and failure behaviors of these straight piping have been investigated under an elevated temperature and a constant internal pressure between 5 and 15 MPa. The deformation and failure conditions of these piping were qualitatively in good agreement with the results of the specimen tests.

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DOI： 10.11484/jaeri-tech-98-026

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The aerosol behavior tests are being performed in WIND project at Japan Atomic Energy Research Institute (JAERI) to investigate FP aerosol behaviors in a reactor coolant piping system. It was identified that the deposition of CsI was remarkable within mid and outlet parts of the test section. Thermo-fluiddynamic analyses with WINDFLOW code suggested that sharp radial temperature decrease in the carrier gas was developed at the vicinity of the cooling surface.

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Research on severe accident for light water reactors has been initiated after the accident of Three Mile Island in 1979 and accelerated all over the world after the Chernobyl accident in 1986. For the severe accident management Nuclear Safety Commission of Japan recommended electric utilities to investigate the possible measures in 1992. In response to this they proposed accident management strategies based on their investigations in 1994 to be voluntarily implemented in several years by reflecting findings from severe accident research. Since the severe accident phenomena largely involve the thermal-hydraulic aspects, it is of great importance to evaluate the thermal-hydraulic behaviors of severe accident in order to reduce the associated uncertainties. In the present article thermal-hydraulic behaviors during severe accident are described in terms of phenomenology, severe accident management, key issues, research methodology and remaining research items. The main phenomena include core melt progression, in-vessel core melt behaviors, molten core coolant interaction, molten core concrete interaction, direct containment heating, hydrogen behaviors, and fission product behaviors.

DOI： 10.3327/jaesj.39.739

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Thermal and structural responses of reactor coolant piping under elevated temperature and pressure are being investigated in piping integrity tests of Wide Range Piping Integrity Demonstration) project at Japan Atomic Energy Research Institute with an aim to evaluate the piping integrity during a severe accident. It was confirmed through the temperature measurement on the outer surface of the test pipe that thermal responses of the pipe were influenced by the internal pressure due to a variation in the magnitude of natural convection.

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The WIND (Wide range piping INtegrity Demonstration) project performed at Japan Atomic Energy Research Institute investigated fission product (FP) aerosol behavior in reactor piping and the integrity of reactor piping under severe accident condition. In the piping integrity test, a straight stainless steel pipe was used to simulate a partial fraction of reactor piping under severe accident conditions. Test analyses were performed using ABAQUS code and the best conditions to investigate the behavior of straight pipe against thermal and pressure loads. The scoping piping integrity test results and the analysis results using ABAQUS are described.

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In WIND (Wide range piping INtegrity Demonstration) project at Japan Atomic Energy Research Institute (JAERI), a scoping test was performed to investigate aerosol behavior in reactor piping under severe accident conditions. In the experiment, CsI aerosol was injected into a stainless steel straight pipe of 2 m in length and 0.1 m in diameter. The experiment showed that aerosol deposition mass was larger at downstream than that at upstream and aerosol was mostly deposited as chemical form of CsI. VICTORIA code well reproduced these trends of experiment. However, the calculated thermochemical equilibrium at upstream did not well explain the observed difference in deposited mass between cesium and iodine. The code also overpredicted aerosol diameter which depends on aerosol agglomeration rate and on condensation or evaporation of species at aerosol surface. These models in VICTORIA were found to be further improved through the WIND experiment analyses.

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A computer code, WINDFLOW, for a three-dimensional thermo-fluiddynamic analysis of a single phase gaseous flow in a reactor coolant piping has been developed at JAERI. A hybrid grid system composed of triangular and quadrangular cross-sectional cells was introduced in WINDFLOW to enhance the modeling flexibility. WINDFLOW was applied to an analysis of tests on thermo-fluiddynamics within a piping performed in WIND project. Formation of a cross-sectional secondary flow resulted from the natural convection was predicted. Appearance of a sharp cross-sectional temperature gradient at the vicinity of the piping wall was confirmed in the analysis. The predictions on the gaseous temperature distribution were compared with the test results. It was found that WINDFLOW tended to underestimate the axial temperature decrease and could qualitatively well reproduce radial temperature distribution observed in the tests.

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The ART (Analysis of Radionuclide Transport) Mod2 code has been developed at JAERI as a module of the THALES-2 code for probabilistic safety assessment (PSA). ART Mod2 calculates the fission products (FP) behavior in the reactor coolant system and in the containment under severe accident conditions. The analytical capability of the code in predicting the FP transport and deposition in pipings was assessed against the WAVE (Wide range Aerosol model VErification) and FALCON experiments. Moreover, the model on iodine chemistry in containment sump was tested against the Radionuclide Test Facility (RTF) 3B experiments conducted in the ACE (Advanced Containment Experiment) program. These assessment studies showed that the coupling between thermohydraulics and FP behavior, and the appropriate treatment of chemical reaction are of great importance to accurately evaluate the source term.

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Authorship：Lead author,　Last author,　Corresponding author   Language：English   Publishing type：Research paper, summary (international conference)  

File： THALES2STCPMELCOR.pdf

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DOI： 10.11484/jaeri-m-92-061

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DOI： 10.11484/jaeri-m-92-055

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DOI： 10.11484/jaeri-m-92-035

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Language：English   Publisher：KERNFORSCHUNGSZENTRUM KARLSRUHE GMBH  

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DOI： 10.11484/jaeri-m-91-175

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DOI： 10.11484/jaeri-m-89-213

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DOI： 10.11484/jaeri-1297

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Event date： 2019.9

Language：Japanese   Presentation type：Oral presentation (general)  

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Event date： 2017.11

Language：English  

Venue：Las Vegas  

During the Fukushima Daiichi NPP accident, I-131 release could have occurred from a large amount of contaminated water in the basements of Units 2 and 3 reactor buildings because of the gas-liquid partition of I-131 and steam generation from the accumulated water by decay heat. The chemical form of certain fraction of released Cs could have been CsBO$_{2}$, which was formed by reaction of CsOH with the boron originated from the B$_{4}$C absorbers. The chemical form could affect not only the Cs-137 source term but also the environmental transport behavior. Te-129m release behavior showed that the release from the containment vessel could be affected by the difference in failure location between the top or middle height of drywell without pool and the bottom of suppression pool where the pool scrubbing is expected. These findings have never been considered in most of the existing severe accident codes such as MELCOR that have been developed based on the findings of the TMI-2 accident.

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Event date： 2017.9

Language：Japanese  

Country：Japan  

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Event date： 2016.2

Language：English  

Venue：Kanpur   Country：India  

Highlights of WAVE and WIND experiments on FP aerosol behavior in piping under severe accident conditions performed at JAEA in 1990's are presented. It was confirmed that the FP gas/aerosol deposition due to thermophoresis or condensation can be mostly predicted well by conventional models. In WAVE experiments using a horizontal pipe with thermal gradient along the main stream or an upward oriented pipe, conventional deposition calculation based only on thermo-hydraulics of main stream could not reproduce well the measurements and a detailed deposition calculation was needed considering the secondary flow in the pipe cross section. In WIND experiment, effect of HBO$_{2}$ on Cesium deposition was investigated and the deposition mass increased due to formation of CsBO$_{2}$. The experimental results and chemical analyses are presented. This presentation could contribute to peaceful use of nuclear energy and its human resource development in India.

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Event date： 2014.11

Language：English  

Venue：Ulsan   Country：Korea, Democratic People's Republic of  

During core cooling at Fukushima Daiichi NPP accident, large amount of contaminated water was accumulated in the basements of reactor buildings at Units 1 to 4. The estimated ratios of $^{131}$I and $^{137}$Cs quantities in water to the core inventories are 0.51\%, 0.85\% at Unit 1, 74\%, 38\% at Unit 2 and 26\%, 18\% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere. Many evaluations for I-131 release have been performed so far by MELCOR or the reverse estimation with SPEEDI. The SPEEDI reverse predicted significant release until March 26 while no prediction in MELCOR after March 17. The present study showed that iodine release from accumulated water due to radiolytic conversion from I$^{-}$ to I$_{2}$ and gas-liquid partition of I$_{2}$ may explain the release between March 17 and 26. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks.

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Event date： 2014.11

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Venue：Busan   Country：Korea, Democratic People's Republic of  

Nuclear Human Resource Development Center (NuHRDeC) of JAEA has conducted nuclear human resource development for more than 50 years since its establishment in 1958. At present, NuHRDeC conducts international Nuclear Human Resource Development "Instructor Training Program (ITP)", which is training scheme launched in 1996 under contract with MEXT in order to support Asian countries seeking peaceful use of nuclear energy. The purpose of ITP is to develop a self-sustainable training system in Asian countries, which disseminates the knowledge and technology on nuclear engineering in their countries. NuHRDeC has expanded the ITP considering change of needs after the Fukushima accident and development stages of these countries. As of FY2014, ITP is applied to eight Asian countries. Present paper summarizes the recent efforts of NuHRDeC and the encountered issues.

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Event date： 2014.9

Language：Japanese  

Country：Japan  

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Event date： 2014.2

Language：English  

Venue：Vienna   Country：Austria  

To assess the extent of prospective doses to the people to be returned after they start again their daily life in the contaminated areas, we have developed a probabilistic method for estimating the dose distribution taking into account the realistic contamination conditions, and population-specific lifestyle habit. The measurements and surveys were performed with the cooperation of various population groups living in Fukushima Prefecture. The prospective dose distributions of the people to be returned were assessed based on these data. In some areas, the doses received by the people to be returned are less than those received by the continuously living areas. Measurements and assessments of doses to the people to be returned are important to ensure the safety of the people. Results of this study are not taken into account the contribution of doses from ingestion pathway. To assess precisely the consequences from radiation exposures, further considerations are needed.

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Event date： 2014.1

Language：Japanese  

Country：Japan  

During the process of core cooling at Fukushima Daiichi Nuclear Power Plants accident, large amount of contaminated water was accumulated in the basements of the reactor buildings at Units 1 to 4. Most of iodine dissolved into water becomes I$^{-}$ and some of I$^{-}$ species could be changed to I$_{2}$. Certain fraction of iodine in water could be released to atmosphere due to gas-liquid partition and contribute to increase in the release to environment. A lot of evaluations for I-131($^{131}$I) release have been performed so far by the MELCOR calculation or the SPEEDI reverse estimation. The SPEEDI reverse predicted significant release until March 26, 2011 while no prediction in MELCOR after March 17. The present study showed that iodine release from accumulated water due to gas-liquid partition may explain the release between March 17 and 26.

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Event date： 2011.2

Language：English  

Venue：Washington, D.C.  

In VEGA experimental program on radionuclide release from irradiated fuel under severe accident condition, 10 tests in total were performed at JAEA from 1999 to 2004 under inert and steam atmospheres including the highest pressure or temperature conditions. These tests showed that increase in release rate at high temperature around UO$_{2}$ melting point and decrease in release rate under elevated pressure that was a first observation in the world. The mechanism of pressure effect on release was examined and the release model with pressure effect was proposed. The data on low-volatile radionuclide release, release from MOX fuel, the effect of fuel oxidation and eutectic reaction with cladding on release were obtained through the tests. In addition, the effect of obtained results on source term evaluation and effectiveness of the accident management measures were investigated. This presentation includes the above described outcomes, limitations of VEGA program, and future issues.

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Event date： 2009.6

Language：Japanese  

Country：Japan  

Utilization of risk information for ensuring nuclear safety has recently made a progress in USA and European countries with development of PSA technology. The Nuclear Safety Commission published the basic policy for introduction of risk informed regulation (RIR). The purposes are further enhancement of rationality, consistency and transparency in the nuclear safety regulation and appropriate allocation of limited resources. In 2003, the special committee proposed the interim safety goals and derived the performance goals for NPPs in 2006. The regulatory body (NISA) also has enhanced utilization of risk information. A pipe rupture incident occurred at Hamaoka Unit-1 in 2001 and NISA referred PSA to confirm the adequacy of the corrective actions. It is also important to enhance understanding between public and people concerned with atomic energy. This lecture describes recent utilization of risk information as described.

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Event date： 2009.5

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Venue：Jeju   Country：Korea, Democratic People's Republic of  

In March 2009, the safety importance classification guide was partly revised. The guide was originally prepared for design phase but had been also referred to the construction and operation phases because the reliability of safety function should be maintained at all phases according to their importance. The maintenance significance had been practically determined using examples of the SSC classification attached to the guide. In January 2009, utilization of risks was initiated in the new maintenance program but conservative evaluation between risk significance and the SSC classification is being taken as the maintenance significance. The present revision clarified that utilization of risks is effective for realizing the safety function in operation phase. That is, if the safety function is maintained, maintenance level for SSC can be determined using risks separately from examples in the guide. Utilization of risks would enable further effective maintenance or inspection.

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ATRふげんで照射されたMOX燃料を大気圧下で燃料の融点を超える3123Kまで昇温したVEGA-M1実験を実施し、放射性物質の放出を、γ線計測、ICP-AES、α線計測により評価した。実験における放射性物質の放出割合及びUO2燃料実験に基づくORNL-Boothモデルを用いた解析による放出割合の評価から、CsとSbは実験及び解析ともにほぼ全量放出を示した。また、解析はBa、Sr、Ruの放出を保守側に評価し、Uの放出及び2800K以下のPu放出はほぼ適切に予測した。しかしながら、同モデルはPuO2の蒸気圧があまり増加しない2650K以下のデータに基づいて作成されたためか、2800K以上のPu放出を過小評価した。Pu放出割合の実験値に基づいて実験式を作成し、原研のソースターム解析コードTHALES-2に組み込んでソースタームを計算したところ、BWRの代表的なシビアアクシデント数例については、Puに関するめやす線量を用いた被ばく評価を行う場合の立地評価事故の仮定（格納容器への瞬時Pu1%放出）は十分な保守性を有していることを確認した。

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原研では、核分裂生成物（FP）の放出機構解明とソースターム高精度予測を目的として、シビアアクシデント条件下における燃料からのFP放出挙動を調べるVEGA実験を行っている。同実験で観測された圧力効果を基に、その機構について検討した結果、高圧では、従来、律速と考えられていた結晶粒内の個体拡散に加え、燃料開気孔中のガス拡散も影響しており、2段階の拡散計算を行う必要が明らかになった。しかしながら、2段階拡散方程式は非定常で数値的に解く必要があることから、確率論的安全評価用に、より簡易な手法の定式化を行った。その結果、従来の放出率速度係数を計算する式に1/√Pを乗じることにより圧力効果が近似できることを見出した。また、導出した圧力影響放出モデルを原研のソースタームコードTHALES-2に組み込んで、TQUXシーケンスにおけるCsI放出に対する圧力影響について調べた所、圧力容器破損前の放出量が減り、冷却系への沈着量が減少する結果、冷却系から再蒸発して圧力容器内に浮遊する量が減少し、格納容器破損時に圧力容器から格納容器を経て環境中へ放出する量が低下した。

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56GWd/tUまで照射された高燃焼度BWR燃料を原研の研究炉JRR-3で再照射して短半減期核種を生成させ、その燃料を水蒸気雰囲気下で2773Kまで昇温したVEGA-6実験を行い、シビアアクシデント条件下における短半減期核種を含む放射性物質の放出データを取得した。

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