Language：English  

Accelerator-driven system (ADS), the coolant of which is lead-bismuth eutectic (LBE), has been designed by Japan Atomic Energy Agency. Estimating corrosion rate at the wall surface of LBE channel is an important issue in considering safety and the life of the entire structure. The corrosion rate depends on state of oxygen layers forming at the material surface. Therefore, this study aims to develop a method to evaluate the corrosion rate in ADS for the design study by estimation of the oxide layer growth and dissolution (OLGD) rates by means of numerical analysis. The OLGD rates, mass transfer rates of oxygen and iron between the material and LBE and advection-diffusion rates of them in LBE depend on each other. Therefore, in order to estimate OLGD rates, the three numerical analysis models should be coupled. For the advection-diffusion calculation, to use CFD code should be reasonable approach to analyze complex flow in ADS, while for the OLGD and the mass transfer calculation, to use some correlation equations should be reasonable because their scales are much smaller than the advection-diffusion. The present work has developed the analysis method of OLGD rates by using JUPITER code, which is CFD code developed in JAEA. In terms of the correlation equations of OLGD and mass transfer rates, existing models used in a previous study were used with modified.

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The Okamura explicit method (OEM) for depletion calculation was developed by modifying the matrix exponential method for dynamic nuclear fuel cycle simulation. The OEM suppressed the divergence of the calculation for short half-life nuclides, even for long time steps. The computational cost of the OEM was small, equivalent to the Euler method, and it maintained sufficient accuracy for the fuel cycle simulation.

DOI： 10.1080/00223131.2022.2131646

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

DOI： 10.1088/1748-0221/17/10/P10005

Web of Science

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Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides (MAs) included in high level wastes discharged from nuclear power plants. The ADS is a lead-bismuth cooled tank-type reactor with 800 MW thermal power. It is supposed that the ADS is safer than conventional critical reactors because it is operated in a subcritical state. The previous study performed the transient analyses for the typical ADS accidents such as unprotected loss of flow or beam overpower. It was shown that all calculation cases except loss of heat sink (LOHS) satisfied the no-damage criteria. To avoid the damage by LOHS, the ADS equips Direct Reactor Auxiliary Cooling System (DRACS) to remove the decay heat. The most important points of a DRACS operation are its reliability and to ensure the flowrate in a natural circulation state. This study aims to perform the CFD analysis of the natural circulation to clarify the flowrate in the ADS reactor vessel.

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Nuclear Material Balance code version 4.0 (NMB4.0) has been developed through collaborative R\&D between Tokyo Institute of Technology and JAEA. Conventional nuclear fuel cycle simulation codes mainly analyze actinides and are specialized for front-end mass balance analysis. However, quantitative back-end simulation has recently become necessary for considering R\&D strategies and sustainable nuclear energy utilization. Therefore, NMB4.0 was developed to realize the integrated nuclear fuel cycle simulation from front- to back-end. There are three technical features in NMB4.0: 179 nuclides are tracked, more than any other code, throughout the nuclear fuel cycle; the Okamura explicit method is implemented, which contributes to reducing the numerical cost while maintaining the accuracy of depletion calculations on nuclides with a shorter half-life; and flexibility of back-end simulation is achieved. The main objective of this paper is to show the newly developed functions, made for integrated back-end simulation, and verify NMB4.0 through a benchmark study to show the computational performance.

DOI： 10.1051/epjn/2021019

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

DOI： 10.1080/00223131.2019.1647889

Web of Science

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

Scopus

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

DOI： 10.1115/ICONE26-82012

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

DOI： 10.1016/j.pnucene.2018.02.007

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At the Kyoto University Critical Assembly, a series of reaction rate experiments is conducted on the accelerator-driven system (ADS) with spallation neutrons generated by the combined use of 100 MeV protons and a lead and bismuth target in the subcritical state. The reaction rates are measured by the foil activation method to obtain neutron spectrum information on ADS. Numerical calculations are performed with MCNP6.1 and JENDL/HE-2007 for high-energy protons and spallation process, JENDL-4.0 for transport and JENDL/D-99 for reaction rates. The reaction rates depend on subcriticality is revealed by the accuracy of the C/E (calculation/experiment) values. Nonetheless, the accuracy of the reaction rates at high-energy thresholds remains an important issue in the fixed-source calculations.

DOI： 10.1080/00223131.2017.1389314

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To investigate the applicability of the pulsed neutron source (PNS) method using a pulsed spallation neutron source (PSNS) for an on-line subcriticality monitoring system for an accelerator-driven system (ADS), a subcriticality experiment is conducted using Kyoto University Criticality Assembly (KUCA) in combination with the fixed-field alternating gradient (FFAG) accelerator. Reactivity values obtained from different traditional techniques, the area-ratio method and the $\alpha$-fitting method, are discussed with respect to the applicability to on-line subcriticality monitoring. The result shows that the area-ratio method robustly and accurately monitors subcriticality in shallow subcritical states of negative reactivity of up to a few dollars; however with this method, it faces problems with temporal fluctuations, spatial dispersion, and sensitivity to the proton-beam current with increasing depth of subcriticality. As a complement to this method, it is shown that the $\alpha$-fitting method alleviates such problems in deep subcritical state. Moreover, a proposed fitting technique using the maximum-likelihood estimation method based on the Poisson distribution is robust enough to be applicable for measuring negative reactivity of up to roughly nine dollars.

DOI： 10.1080/00223131.2016.1274687

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Sensitivity and uncertainty analyses of lead (Pb) isotope cross sections are conducted with the use of sample reactivity experiments at the Kyoto University Critical Assembly (KUCA). With the combined use of the SRAC2006 and MARBLE code systems, attempts are made to precisely examine the contributions of the reactions and energy regions of Pb isotope cross sections to reactivity based on the covariance data of JENDL-4.0. Moreover, the effect of decreasing uncertainty is discussed in terms of the accuracy of sample reactivity by applying the cross section adjustment method to the uncertainty analyses. From the results of the sensitivity and uncertainty analyses, the reliability of Pb isotope cross sections, such as the Pb isotope covariance data of JENDL-4.0, is compared with the JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1 libraries.

DOI： 10.1080/00295639.2016.1272976

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

DOI： 10.1080/00223131.2016.1179600

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DOI： 10.1080/00223131.2016.1140598

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Sample reactivity experiments on the uncertainty analyses of Pb nuclear data are carried out by substituting Al plates for Pb ones at the Kyoto University Critical Assembly, as part of basic research on Pb-Bi for the coolant. Numerical simulations of sample reactivity experiments are performed with the Monte Carlo calculation code MCNP6.1 together with four nuclear data libraries JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1, to examine the accuracy of cross-section uncertainties of Pb isotopes by comparing measured and calculated sample reactivities. A library update from JENDL-3.3 to JENDL-4.0 is demonstrated by the fact that the difference between Pb isotopes of the two JENDL libraries is dominant in the comparative study, through the experimental analyses of sample reactivity by the MCNP approach. In addition, JENDL-4.0 reveals a slight difference from ENDF/B-VII.0 in all Pb isotopes and $^{27}$Al, and from JEFF-3.1 in $^{238}$U and $^{27}$Al.

DOI： 10.1080/00223131.2015.1068716

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The JAEA has investigated an accelerator-driven system (ADS) to transmute minor actinides which will be partitioned from the high level waste. There are various inherent issues for the research and development on the ADS. This study introduces the current two activities to get a feasible and reliable ADS concept. For the feasibility, the beam window design is one of the most important issues. To mitigate the design condition of the beam window, namely to reduce the proton beam current, the subcritical core concepts with subcriticality adjustment mechanism were investigated. As the result, the proton beam current reduced from 20mA to 10mA by the installation of the control rods. For the reliability, the two accelerators concept was proposed and the beam trip frequency was estimated by the J-PARC LINAC data and Monte-Carlo method. Through these investigations, it was confirmed that the two accelerators concept was useful to realize reliable accelerator operation for the ADS.

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The present study focuses on transmutation of Pu and minor actinide in Japanese case without utilizing Pu as resource. Pu can be transmuted by two groups of technology: conventional ones without reprocessing of spent fuel from transmuter and advanced ones with reprocessing. Necessary number of transmuters, inventory reduction of actinide and impact on repository are revealed by nuclear material balance analysis. As a whole advanced technology performs better in transmutation efficiency, although required number of transmuters is larger.

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After the severe accident at the Fukushima-1 nuclear power plant, large amounts of contaminated stagnant water have accumulated in turbine buildings and their surroundings. This rapid communication reports calculation of the radionuclide inventory in the core, collection of measured inventory in the stagnant water, and estimation of radionuclide release ratios from the core to the stagnant water. This evaluation is based on data obtained before June 3, 2011. The release ratios of tritium, iodine, and cesium were several tens of percent, whereas those of strontium and barium were smaller by one or two orders of magnitude. The release ratios in the Fukushima accident were equivalent to those in the TMI-2 accident.

DOI： 10.1080/00223131.2014.946455

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With consideration of the phase-out option from NP utilization in Japan, an ADS for Pu transmutation is designed and scenario analysis is performed. The ADS is designed based on the existing ADS design for MA transmutation, and the 6-batch ADS is selected as a reference design for scenario analysis. In the scenario analysis, once-through scenario of LWR spent fuel is referred as a conventional scenario with a LWR-MOX utilization scenario. As the transmutation scenario, three cases of transmuters that are only-FR, only-ADS and both-FR+ADS are analyzed. The numbers of necessary transmuters are obtained as 15 to 32, and, the necessary period for transmutation as 180 to 240 years. Benefit on repository by reduction of Pu and MA is reduction of repository area by a factor of five and of decay time of toxicity by a magnitude of order. The FR+ADS scenario is considered to be a modest solution, though the ADS scenario is preferable if rapid transmutation is required.

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The Japan Atomic Energy Agency (JAEA) has the plan to construct Transmutation Physic Experimental Facility (TEF-P) under a framework of J-PARC (Japan Proton Accelerator Research Complex) project. TEF-P is a critical assembly which can load Minor Actinide (MA) fuels to perform reactor physics experiments for transmutation systems such as Accelerator-Driven System (ADS) or Fast Reactor (FR). The facility can also use proton beam from the J-PARC accelerator to investigate the controllability of ADS. Cur-rent status and activities for TEF-P are described.

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For safe storage of zeolite wastes generated by treatment of radioactive saline water at the Fukushima Daiichi Nuclear Power Station, properties of the Herschelite adsorbent were studied and its adsorption vessel was evaluated for hydrogen production and corrosion. Hydrogen production depends on its water level and dissolved species because hydrogen is oxidized by radicals in water. It is possible to evaluate hydrogen production rate in Herschelite submerged in seawater or pure water by taking into account of the depth effect of the water. The reference vessel of decay heat 504 W with or without residual pure water was evaluated for the hydrogen concentration by thermal hydraulic analysis using obtained fundamental properties. Maximum hydrogen concentration was below the lower explosive limit (4 \%). The steady-state corrosion potential of a stainless steel 316L increased with absorbed dose rate but its increase was repressed by the presence of Herschelite. At 750 Gy/h and $<$60$^{\circ}$C which were values evaluated at the bottom of the vessel of 504 W, the localized corrosion of SUS316L contacted with Herschelite would not immediately occur under 20,000 ppm of Cl$^{-}$ concentration.

DOI： 10.1080/00223131.2014.924446

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DOI： 10.1016/j.nds.2014.04.123

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Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by material balance analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the nuclear material balance analysis (NMB) code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels and an ability to estimate waste and repository. Three scenario-groups of once-through, Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. By construction of two full MOX- or ROX- reactors, Pu amount is reduced to about a half and isotopic vector of Pu is deteriorated as nuclear weapon especially in terms of spontaneous fission neutron. Effects by ROX are more significant than MOX in both amount and vector. Repository footprint and potential radio-toxicity is not reduced by MOX and ROX because heat and toxicity of MOX and ROX spent fuel is considerably high.

DOI： 10.1080/00223131.2014.846834

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DOI： 10.1080/00223131.2013.811954

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The transmutation of iodine-129 in accelerator driven system (ADS) is studied. The sodium iodide assembly loadings inside the core of ADS and in the surrounding core region are considered. The introduced concept of ADS with a power of 800 MWt is able to transmute 250 kg/y of minor actinides (MAs) and 46 kg/y of Iodine-129 that supports ten PWRs. The initial loading masses of MAs and I-129 in ADS were equal to 3810 kg and 824 kg, respectively.

DOI： 10.1016/j.anucene.2013.01.042

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DOI： 10.1016/j.anucene.2012.12.016

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DOI： 10.1016/j.nucengdes.2012.09.019

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DOI： 10.17146/aij.2012.163

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Frequent beam trips as experienced in existing high power proton accelerators may cause thermal fatigue problems in Accelerator-Driven System (ADS) components which may lead to degradation of their structural integrity and reduction of their lifetime. Thermal transient analyses were performed to investigate the effects of beam trips on the reactor components, with the objective of formulating the feasible ADS design and determining the requirements for accelerator reliability. From the aspect of the thermal stress and the annual plant availability, our results indicated that the acceptable beam-trip frequency was classified by three criteria, according to the beam trip duration, $\tau_b$ : $2\times10^4$ times per year for 0 $\leq\tau_b\leq$ 10 s, $2\times10^3$ for 10 s $<\tau_b\leq$ 5 min, and 42 for $\tau_b>$ 5 min. And the annual plant availability was estimated to be 70 \% or greater in cases where the beam trip frequencies decreased to the acceptable level. In order to consider measures to reduce the beam trip frequency of the high power accelerator for ADS, we compared the acceptable beam-trip frequency with the operational data of existing accelerators. The result of the comparison showed that the beam trip frequency with durations of 10 s or less was within the acceptable level. On the other hand, for the beam trip with durations of 10 s $<\tau_b\leq$ 5 min and $\tau_b>$ 5 min, it is necessary to reduce the beam trip frequency to about $\frac{1}{6}$ and $\frac{1}{35}$ to satisfy the acceptable level, respectively.

DOI： 10.1080/00223131.2012.669239

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Language：Japanese   Publisher：Atomic Energy Society of Japan  

After the severe accident in the Fukushima-1 Nuclear Power Plant, large amount of contaminated stagnant water has been produced in turbine buildings and those surroundings. This rapid communication reports calculation of radionuclide inventory in the core, collection of measured inventory in the stagnant water, and estimation of radionuclide release ratios from the core to the stagnant water. The present evaluation is based on data obtained before June 3, 2011.

DOI： 10.3327/taesj.J11.040

CiNii Books

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Language：Japanese   Publisher：Atomic Energy Society of Japan  

DOI： 10.11561/aesj.2012f.0.208.0

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Zeolite adsorbents are to be used for decontamination of radioactive water in Fukushima Dai-ichi Nuclear Power Station. Evaluation of hydrogen production by water radiolysis during decontamination is important for safe operation. Thus hydrogen production from the mixture of zeolite adsorbents and seawater was studied because seawater was urgently used as a coolant for the fuels. The hydrogen yield from the mixture decreased at a high weight fraction of zeolites. However, the measured yield was higher than the yield expected from the direct radiolysis of seawater in the mixture, which would decrease proportional to the weight fraction of seawater. The result suggests that the radiation energy deposited to zeolite is involved in the hydrogen formation. From the results, the hydrogen production rate was evaluated to be 3.6 mL/h per ton of radioactive water before decontamination. After the process, it was evaluated to be 1.5 L/h per ton of waste adsorbents due to the high dose rate.

DOI： 10.3327/taesj.J11.021

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

DOI： 10.1016/j.anucene.2011.06.015

Scopus

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Japan Atomic Energy Agency (JAEA) has promoted the design study of the Accelerator-Driven System (ADS) cooled by lead-bismuth eutectic (LBE) to transmute minor actinide. The structural integrity of the fuel claddings is one of key issues in designing the ADS. For ADS, frequent beam trips as experienced in existing high power proton accelerators may cause thermal fatigue problems which may lead to degradation of the structural integrity and reduction of the lifetime of the fuel claddings. To estimate the structural integrity of the fuel claddings, the thermal stress of the fuel cladding was evaluated for following two states: the steady state and the transient state. For the steady-state analysis, the thermal stress was divided into two types: the primary stress and the secondary stress, which were caused by the inner pressure and the thermal load, respectively. These stresses were compared with the design stress limits. As a result, it was confirmed that their structural strength was sufficient under the conditions of design temperature and corrosion environment without irradiation. For the transient-state analysis, the number of acceptable beam trips was evaluated based on the total strain range due to the primary and secondary stress of the fuel cladding. As a result, the number of acceptable beam trips was evaluated to be over $1\times10^6$ times. If the fuel cladding was replaced once every two years, the acceptable beam-trip frequency was estimated to be greater than $5\times10^5$ times per year. This acceptable frequency was 20 times greater than the expected beam-trip frequency caused by the accelerator for the JAEA's reference ADS.

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For the feasibility of ADS, the study for the design of a beam window is important. The characteristics of the thermal-fluid behavior around the beam window with the profile of proton beam were simulated by CFD analysis. However, because of a lack of the experimental data for the validation of this numerical analysis, the estimation for the reliability of a beam window design was not enough. And the erosion of the material in the flowing LBE is also significant issue. However, the flow measurement of the general liquid metals was difficult. The flow measurement techniques for LBE flow has been developed by using UVP method. A developed system realizes the measurement of the velocity vector profiles including the information of the multiple velocity components simultaneously on time. And it was successfully applied to the actual LBE flow.

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A feasibility study on natural uranium spallation target in accelerator-driven system (ADS) for minor actinide (MA) transmutation was performed. As a result of comparative study of uranium and lead-bismuth (PbBi) targets in the bare case without blanket surrounding, it was found that uranium target had better neutron generation performance, but limited by the geometrical size due to high neutron absorption in $^{238}$U. In ADS for MA transmutation, uranium used as target instead of PbBi also absorbs neutrons passing the target area. More realistic concept of pin type uranium spallation target cooled by liquid PbBi was considered aiming at enhancing spallation target performance in terms of neutron generation efficiency and operation temperature. The uranium pin target design had nothing better effects on neutron balance of such system than a conventional PbBi target in ADS and it was concluded that uranium target was not suitable for the full-scale ADS.

DOI： 10.1016/j.pnucene.2011.05.019

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

DOI： 10.1016/j.jnucmat.2011.04.032

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Sensitivity calculations are carried out for some neutronic parameters on neutron multiplication in subcritical systems. Sensitivities of the subcriticality multiplication rate k$_{\rm sub}$ and the generation-wise neutron multiplication rate k$_{\rm i}$, which is a new neutronic parameter proposed in the present paper, are calculated, and these sensitivities are compared with the k$_{\rm eff}$ sensitivities which have been conventionally used for sensitivity works. The present sensitivity calculations show that the sensitivities of k$_{\rm sub}$ and k$_{\rm i}$ in small values of $i$ are significantly different from the k$_{\rm eff}$ sensitivities. This result indicates that the sensitivity analyses focusing only on the solution of the eigenvalue equation cannot provide full information on the core properties of subcritical systems, and that the sensitivities of the other neutronic parameters describing real neutron multiplication are essential.

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The cross sections for nuclide production in thin $^{56}$Fe and $^{\rm nat}$Cr targets irradiated by 0.04 to 2.6 GeV protons are measured by direct $\gamma$ spectrometry using two $\gamma$ spectrometers with the resolutions of 1.8 and 1.7 keV for the $^{60}$Co 1332 keV $\gamma$ line. As a result, 649 yields of radioactive residual product nuclei have been obtained. The $^{27}$Al($p$, $x$)$^{22}$Na reaction has been used as a monitor reaction. The experimental data are compared with the MCNPX (BERTINI, ISABEL), CEM03.02, INCL4.2, INCL4.5, PHITS, and CASCADE07 calculations.

DOI： 10.1134/S1063778811040168

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The cross sections for the production of $^{148}$Gd in $^{\rm nat}$W and $^{181}$Ta targets irradiated by 0.4-, 0.6-, 0.8-, 1.2-, 1.6-, and 2.6-GeV protons at the ITEP accelerator complex have been measured by direct $\alpha$ spectrometry without chemical separation. The experimental data have been compared with the data obtained at other laboratories and with the theoretical simulations of the yields on the basis of the BERTINI, ISABEL, CEM03.02, INCL4.2, INCL4.5, CASCADE07, and PHITS codes.

DOI： 10.1134/S1063778811040193

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The cross sections for nuclide production in thin $^{\rm nat}$W and $^{181}$Ta targets irradiated by 0.04 to 2.6GeV protons are measured by direct $\gamma$ spectrometry using two $\gamma$ spectrometers with the resolutions of 1.8 and 1.7 keV for the $^{60}$Co 1332-keV $\gamma$ line. As a result, 1895 yields of radioactive residual product nuclei have been obtained. The $^{27}$Al($p$, $x$)$^{22}$Na reaction has been used as a monitor reaction. The experimental data are compared with the MCNPX (BERTINI, ISABEL), CEM03.02, INCL4.2, INCL4.5, PHITS, and CASCADE07 calculations.

DOI： 10.1134/S1063778811040181

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The cross sections for nuclide production in thin $^{93}$Nb and $^{\rm nat}$Ni targets irradiated by 0.04 to 2.6 GeV protons are measured by direct $\gamma$ spectrometry using two $\gamma$ spectrometers with the resolutions of 1.8 and 1.7 keV for the $^{60}$Co 1332-keV $\gamma$ line. As a result, 1112 yields of radioactive residual product nuclei have been obtained. The $^{27}$Al($p$, $x$)$^{22}$Na reaction has been used as a monitor reaction. The experimental data are compared with the MCNPX (BERTINI, ISABEL), CEM03.02, INCL4.2, INCL4.5, PHITS, and CASCADE07 calculations.

DOI： 10.1134/S106377881104017X

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The impact of partitioning and/or transmutation (PT) technology on the high level waste management was investigated for the fast breeder reactor (FBR) fuel cycle. Three different fuel cycle scenarios involving PT technology were analyzed: (1) partitioning process only (separation of some fission products), (2) transmutation process only (separation and transmutation of minor actinides), and (3) both partitioning and transmutation. It was found that the introduction of FBR cycle without PT yields a smaller repository after 65 years of predisposal storage than the LWR after 50 years by a factor of up to 0.65. The impact of the partitioning-only process on the repository is not significant, but that of the transmutation-only process produces a reduction of area by a factor of 0.25 after 80 years of storage, and that of a coupled partitioning and transmutation by two orders of magnitude after 60 and 295 years of storage for the glass waste and the calcined waste with Sr-Cs, respectively.

DOI： 10.3327/jnst.47.1101

CiNii Books

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The conceptual design study for the beam window of the spallation target is one of the critical issues in the R\&D of the accelerator driven system (ADS). In this study, the investigation to create a feasible concept of the beam window for the ADS was performed by changing the proton beam profile from the gaussian distribution to the parabolic and the flat distributions. The detailed analyses were performed by considering the particle transport of protons and neutrons in the spallation target region, the thermal hydraulics of lead bismuth eutectic (LBE) around the beam window and the structural strength of the beam window. The calculation results presented the concept with the parabolic distribution would be the most feasible concept in the current ADS design condition.

DOI： 10.3327/jnst.47.953

CiNii Books

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To confirm the reliability of calculated reactor physics parameters for the nuclear transmutation systems, the uncertainty calculated by the covariance data prepared in JENDL-3.3 was compared with the differences of the reactor physics parameters in the Monte-Carlo calculation using different nuclear data libraries. The Accelerator Driven System (ADS) and the Minor Actinide (MA) loaded Fast Reactor (FR) were selected as the representative transmutation systems. The results showed that the uncertainties calculated by JENDL-3.3 covariance data were smaller than the differences of the reactor physics parameters among the nuclear data libraries. The cause of this discrepancy was that the covariance data of main nuclides and reactions in JENDL-3.3 were smaller than the relative differences in the cross sections among the nuclear data libraries. It is required to verify the uncertainty by integral experiments and to discuss the uncertainty utilization for the nuclear design accuracy.

DOI： 10.3327/jnst.47.521

CiNii Books

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

DOI： 10.1016/j.jnucmat.2009.10.040

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In Japan Atomic Energy Agency (JAEA), various R\&D for 800 MWt, lead bismuth eutectic (LBE) cooled ADS have been performed to transmute minor actinides (MA) discharged from spent fuel of commercial nuclear power plants. This study introduces the latest two activities for the ADS in JAEA. The first one is the uncertainty analysis for the subcritical core. Since uncertainties of the current MA nuclear data are supposed to be larger than those of other nuclides, it is necessary to improve the accuracy of the neutronics design. We discuss the effect of MA-loaded experiments for its improvement by using the cross-section adjustment procedure. The second activity is the structural analysis of the beam window. The discussions for the feasible concept are performed by using Finite Element Method.

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In the present neutronics design of the Accelerator-driven system (ADS) cooled by lead-bismuth eutectic (LBE), we investigated several methods to reduce the power peak and beam current, and estimated the temperature drops of the cladding tube and beam window. The methods are adjustment of inert matrix ratio in fuel in each burn-up cycle, multi-region design in terms of pin radius or inert matrix content, and modification of the level of the beam window position and the height of the central fuel assemblies. As the result, we optimized the ADS combined with the adjustment of inert matrix ratio in each burn-up cycle, multi-region design in terms of inert matrix content and deepened window level. The maximum temperatures of the optimized ADS at the surface of the cladding tube and the beam window were reduced by 91 and 38 $^{\circ}$C, respectively.

DOI： 10.3327/jnst.45.812

CiNii Books

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Irradiation damage to the beam window in the concept of 800 MWth accelerator-driven system is evaluated. Heat produced in the window is also evaluated. Transport of proton and neutron up to 3.0 GeV is calculated by both PHITS that is the Monte Carlo code for particles and heavy ions and TWODANT that is 2-dimentional deterministic transport code. The beam window is irradiated at the center of the accelerator-driven system with 20 MW proton beam power and neutron from the core during 300 full power days. Heat, displacement per atom, production rate of hydrogen and helium isotopes, and neutron and proton fields are estimated, assuming the Gaussian and flat beam profiles.

DOI： 10.1016/j.jnucmat.2008.02.045

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Feasibility was discussed for lead-bismuth cooled accelerator-driven system (ADS) to transmute minor actinides partitioned from high-level radioactive waste. Since lead-bismuth will cause considerable corrosion and erosion effects at high temperature, the fuel clad temperature must be kept as low as possible. Moreover, the most critical issue of ADS design is engineering viability of the high-power spallation target and the beam window. Thermal-hydraulic and structural analysis was carried out for both the fuel assembly and the beam window. In addition to the analysis in steady state, transient behaviors were also studied during typical transient and unprotected accidents. The results showed that engineering viability is reasonably achievable in nominal operation. For beam trip which will be the most frequent transient, the number of events to cause the failure of the beam window is estimated as more than $10^{5}$. For safety aspects of ADS during unprotected accidents, estimated results showed that ULOF would cause the most significant problem, if beam operation was kept. Therefore, high reliability of the beam shutdown is required for the ADS safety.

DOI： 10.13182/NT08-A3929

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

DOI： 10.1016/j.pnucene.2007.11.017

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We studied how cooling in the predisposal storage period may affect the design of the emplacement area in a repository for radioactive wastes produced by a light-water-reactor nuclear system that uses partitioning and/or transmutation (PT) technology. Three different fuel cycle scenarios involving PT technology were analyzed: (1) partitioning process only (separation of some fission products), (2) transmutation process only (separation and transmutation of minor actinides), and (3) both partitioning and transmutation. The necessary predisposal storage periods for some predefined emplacement configurations were determined through transient thermal analysis. For each scenario, we also estimated the storage capacity required for dry storage. The contributions of PT technology on the storage and disposal were discussed holistically, and we noted that the coupled introduction of partitioning and transmutation processes can bring an appreciable reduction in waste management size.

DOI： 10.3327/jnst.45.84

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Benefit of implementing Partitioning and Transmutation (P\&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. The results showed that, by recycling minor actinides (MA), the placement area could be reduced by 17-29\% in the case of UO$_{2}$-LWR and by 63-85\% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of $^{241}$Am which was a long-term heat source. Further 70-80\% reduction of the placement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types.

DOI： 10.3327/jnst.44.398

CiNii Books

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For a dedicated transmutation system, Japan Atomic Energy Agency (JAEA) has been proceeding with the research and development (R\&D) on an accelerator-driven subcritical system (ADS). JAEA has started a comprehensive R\&D program since the fiscal year of 2002 to acquire knowledge and elemental technology for the validation of engineering feasibility of the ADS. In this paper, the outline and the results in the first three-years stage of the program are reported. For R\&D on accelerator, a prototype cryomodule was built and its good performance in electric field was examined. For R\&D on LBE, various technical data for material corrosion, thermal-hydraulics and radioactive impurity were obtained. For R\&D on subcritical core, engineering feasibility for the ADS was discussed using thermal-hydraulic and structural analysis not only in normal operation but also in transient situations. Reactor physics experiments for subcritical monitoring of ADS were performed at critical assemblies.

DOI： 10.3327/jnst.44.483

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Scopus

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This paper introduces application of the nuclear fuel cycle code FAMILY-21 which was developed in Fast reactor development division of Japan Atomic Energy Agency (JAEA). In the first part, benchmark comparison between FAMILY-21 and Nuclear Material Balance version 4.0 (NMB4.0) which is open-source code for integrated fuel cycle simulation developed in the sector of nuclear science research, JAEA is carried out. Second part shows preliminary application of FAMILY-21 for Japan.

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This work demonstrates the value of code verification as an initial step in utilizing fuel cycle simulation. Cyclus and NMB are open-source fuel cycle simulators that provide computational modeling of nuclear fuel cycle alternatives and were chosen by Argonne National Laboratory and Japan Atomic Energy Agency (JAEA), respectively, for a multi-year collaboration on fuel cycle benchmarks. Both are relatively new and can be improved after conducting a rigorous code-to-code comparison. Initial verification of these simulators was performed using a set of hypothetical scenarios for once-through and multi-recycle fuel cycles. The results of this work identify how differences in scenario definitions and the modeling methodologies of the two simulators lead to differences in results in material inventories, mass flows, and other important metrics for fuel cycle assessments.

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Adsorption columns filled with zeolite are used to treat contaminated water containing radioactive cesium generated by the Fukushima Daiichi Nuclear Power Station accident. As the contaminated water treatment progresses, the radioactive cesium in the adsorption column becomes highly concentrated, and the adsorption column becomes a high radiation source. To evaluate the radiation effects such as decay heat and radiolytic hydrogen production in the adsorption column, the concentration of radioactive cesium in the adsorption column is necessary, but since it is difficult to evaluate the concentration by measurement, it is estimated by simulation. In this research, a zeolite column adsorption dynamics simulation (Zeolite Adsorption Column: ZAC) code was developed to calculate the concentration of radioactive materials such as radioactive cesium in a zeolite filled adsorption column when they are injected into the column. The code was validated through comparison of calculation results with existing codes and experimental results of small column tests. This report presents the details of the model, the handling of the code, and the validity of the results for the developed code.

DOI： 10.11484/jaea-data-code-2024-002

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The aim and overall picture of partitioning and transmutation technology will be described, and the recent research and development status of JAEA for each of partitioning and transmutation will be outlined.

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This tutorial contains a role of accelerator-driven system (ADS) in the nuclear fuel cycle and necessity of nuclear data to realize the ADS. After an overview of Japanese nuclear fuel cycle and government direction, geological disposal concept of high-level waste (HLW) will be described. By partitioning problematic elements from the HLW and transmuting, utilizing or storing them, geological disposal can be changed. ADS plays a role of transmuting minor actinide (MA) separated from HLW to fission product (FP), which are less radio-toxic than MA. The principle of ADS will be introduced with technological issues, and finally utilization of nuclear data for R\&D on ADS will be introduced.

DOI： 10.11484/jaea-conf-2022-001

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The overall picture and the goals of partitioning and transmutation technology are described. Then, the main technologies currently being developed for partitioning and transmutation are respectively outlined. Finally, a comprehensive evaluation of the nuclear fuel cycle including partitioning and transmutation, conducted in the Atomic Energy Society of Japan is introduced.

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The quantitative prediction and analysis of the future nuclear energy utilization scenarios are required in order to establish the advanced nuclear fuel cycle. However, the nuclear fuel cycle consists of various processes from front- to back-end, and it is difficult to analyze the scenarios due to the complexity of modeling and the variety of scenarios. Japan Atomic Energy Agency and Tokyo Institute of Technology have jointly developed the NMB code as a tool for integrated analysis of mass balance from natural uranium needs to radionuclide migration of geological disposal. This user manual describes how to create a database and scenario input for the NMB version 4.0.

DOI： 10.11484/jaea-data-code-2021-016

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It has to be confirmed that the accelerator-driven system (ADS), which is dedicated to transmuting minor actinides, is subcritical in any state by measurements. In the previous research, we have proposed a procedure in which the core safely and efficiently approaches the target subcriticality before the operation. In this procedure, the reference value of the subcriticality at the initial state is measured by the area ratio method capable of the absolute value measurement. The area ratio method uses a pulsed neutron source. However, specific and practical parameters of the accelerator for the area ratio method have not been determined. In this study, we determined the accelerator parameters with the consideration of the uncertainties derived by the dead-time of the detector and the statistical error of the count ratio. In addition, we estimate the coating amount of the sample nuclide in the assumption of the use of the fission chambers.

DOI： 10.11484/jaea-research-2021-016

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Takeshita Laboratory, Tokyo Institute of Technology, has been developing Nuclear Material Balance code version 4.0 (NMB4.0) in collaboration with Japan Atomic Energy Agency (JAEA). This report summarized the outline and functions of NMB4.0.

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A radiation shielding analysis was performed for the structure located above the spallation target of an accelerator-driven system (ADS), assuming one cycle of an 800 MW thermal and 30 MW beam power operation. In this analysis, the Monte Carlo particle transport code PHITS and the activation analysis code DCHAIN-PHITS were used. The structures to be analyzed are a beam duct above the target, a beam transport room located above the ADS reactor vessel, beam transport equipment, and the room ceiling. For each structure, the radiation doses and radioactivities during and after the operation were estimated. Furthermore, the shielding structure of the ceiling was determined. As a result, it was found that the radiation dose at the site boundary would be sufficiently lower than the legal limit by applying the determined shielding structure. Moreover, under the condition of this study, it was shown that the effective dose rate around the beam transport equipment positioned above the target after the operation exceeded 10 mSv/h, and that the maintenance and replacement of the equipment in the room would require remote handling.

DOI： 10.11484/jaea-research-2021-012

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Zeolite adsorbents are to be used for decontamination of radioactive water in Fukushima Dai-ichi Nuclear Power Station. Evaluation of hydrogen production by water radiolysis during decontamination is important for safe operation. Thus hydrogen production from the mixture of zeolite adsorbents and seawater was studied because seawater was urgently used as a coolant for the fuels. The hydrogen yield from the mixture decreased at a high weight fraction of zeolites. However, the measured yield was higher than the yield expected from the direct radiolysis of seawater in the mixture, which would decrease proportional to the weight fraction of seawater. The result suggests that the radiation energy deposited to zeolites was involved in the hydrogen formation. From the results, the hydrogen production rate was evaluated to be 3.6 mL/h per ton of radioactive water before decontamination. After the process, it was evaluated to be 1.5 L/h per ton of waste adsorbents due to the high dose rate.

DOI： 10.15669/fukushimainsights.Vol.4.37

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Nuclear Material Balance code (NMB code) have been developed in Japan Atomic Energy Agency. The NMB code will be updated with the function of mass balance analysis at the backend process such as reprocessing, vitrification and geological disposal. In order to perform its analysis with high accuracy, it is necessary to expand the number of FP nuclides calculated in the NMB code. In this study, depletion calculation by ORIGEN code was performed under 3 different burn-up conditions such as spent uranium fuel from light water reactor, and nuclides were selected from 5 evaluation indexes such as mass and heat generation. In addition, the FP nuclides required to configure a simple burnup chain with the same calculation accuracy as ORIGEN in the NMB code was selected. As the result, two lists with different number of nuclides, such as "Detailed list" and a "Simplified list", were created.

DOI： 10.11484/jaea-data-code-2020-023

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As a part of partitioning and transmutation technology development to reduce the burden of radioactive disposal, an investigation of Accelerator-Driven System (ADS) has been performed by Japan Atomic Energy Agency. A beam window, which is an inherent structure of the ADS, is planned to be made from T91 steel and its shape is a thin hemisphere shell. However, it had never been tried to manufacture it out of T91 steel. In this investigation, we tried to manufacture miniature beam windows by cutting T91 steel, and to discuss the process, manufacturing accuracy and geometry measurement methods. As a result, considering a real scale ADS beam window, a figure error between designing and machining ones is estimated to be about 5%. Its effect would be very small to the structural strength.

DOI： 10.11484/jaea-technology-2020-026

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The Takeshita Laboratory at Tokyo Institute of Technology has started to develop a Nuclear Material Balance code (NMB code) in collaboration with Japan Atomic Energy Agency. This report summarized the results of the joint research conducted in 2019.

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In order to discuss the technological development and human resource development necessary for the future nuclear fuel cycle, various quantitative analyzes were conducted assuming a wide range of future nuclear power generation scenarios. In the evaluation of quantities, the future power generation of LWR and fast reactor, the amount of spent fuel reprocessing, etc. were assumed, and the amount of uranium demand, the accumulation of spent fuel, plutonium, vitrified waste etc. were estimated.

DOI： 10.11484/jaea-data-code-2020-005

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The JAEA has proposed the double-strata strategy which will introduce a dedicated minor actinide (MA) transmutation cycle using an accelerator-driven system (ADS). In the previous study, the ideal fuel condition was supposed in the neutronics design of the ADS. For example, impurities such as rare earth (RE) nuclides which would accompany with MA, were not assumed. However, these nuclides would accompany with the ADS fuel and the capture reaction of these nuclides deteriorates the neutron economy of the ADS core. This study investigates a new fuel composition based on the SELECT (Solvent Extraction from Liquid-waste using Extractants of CHON-type for Transmutation) process proposed by JAEA. By performing the neutronics calculation of the ADS with the new fuel composition, a feasibility of the new fuel composition will be investigated.

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In this project, long-lived fission products (LLFP) contained in conventional high-level radioactive wastes are separated and their life is reduced, and elements that can be used as resources are separated. By shortening the life of LLFP, it has been shown that it may be possible to dispose in intermediate depth of several tens of meters, meeting safety requirements, instead of geological disposal. In addition, for reassuring recycling of usable elements, possible exposure pathways were evaluated to estimate the safe concentration level of radioactivity.

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High level radioactive waste contains elements with various characteristics. It is possible to reduce the load on the disposal site by separating them according to those characteristics and appropriately dealing with them. In this project, we are working to shorten the life span of long-lived fission products (LLFP). When this technology is realized, high-level radioactive wastes will become new radioactive wastes with low radioactivity. As a result of investigation of disposal concept of new radioactive waste, it turned out that intermediate-depth disposal currently considered for low level radioactive waste may be suitable. Intermediate-depth disposal is a method of small-scale disposal in shallow locations as compared to geological disposal for conventional high-level radioactive waste. We conducted a safety assessment when this disposal is applied to new radioactive wastes, and found that it is possible to safely dispose of for the four LLFPs addressed by this project.

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Potential radiotoxicity defined as a summation of intake dose was estimated for each actinide element to suppose target of recovery ratio of minor actinide (MA). Importance of each element from the viewpoint of the radiotoxicity was evaluated from the evolution of the radiotoxicity and ratio to the total radiotoxicity. In all the 4 types of spent fuels examined, Am is the most important element. For instance, the potential radiotoxicity of Am accounts for 93\% of the total radiotoxicity of actinide elements in HLW produced by reprocessing of spent fuel from pressurized water reactor (PWR). Residual Pu after the recovery of 99.5\% in reprocessing still gives contribution that cannot be ignored in radiotoxicity. When the burn-up of the UO$_{2}$ fuel in PWR increased, the potential radiotoxicity of actinide elements increased almost in proportion to the burn-up, but in case of MOX fuel in PWR and minor-actinide-recycled MOX fuel in fast reactor, the radiotoxicity of actinide elements increased further. Much consideration is required for the recovery of actinide elements in HLW from different types of fuel.

DOI： 10.11484/jaea-data-code-2018-017

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Impact of reduction of source term on design and safety assessment of disposal concept for high level radioactive waste is considered. Reduction of source term in partitioning and transmutation technology is shown with impact on disposal concept. Moreover, cost and technological readiness is outlined.

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DOI： 10.11484/jaea-technology-2017-033

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The present article describes reduction technology of radioactive waste by accelerator that aims to reduce burden of underground disposal. If the technology will be realized, amount of waste will be decreased, period of confinement in underground will be shorten to several hundreds of years, and, the concept of waste disposal can be renewed.

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The JAEA has investigated 800-MWth Accelerator-Driven System (ADS) to transmute minor actinides. One of the most important issues for the ADS development is the design of a beam window which is a boundary of the accelerator and the sub-critical reactor. Since the past shape of the beam window is difficult to fabricate, a hemispherical shape one is considered as a new concept. To realize the hemispherical shape beam window, it is required to decrease the maximum proton beam current. For this purpose, a new concept of sub-criticality adjustment mechanism using control rod (CR) or burnable poison (BP) is investigated. The results of neutronics calculation show that the concepts with boron carbide and tantalum CRs have good property to adjust the sub-criticality. They have a possibility to maintain the proton beam current at 10 mA during the operation, which is a great benefit for the beam window design since the proton beam current becomes half of the previous design. However, it is required to consider CR drive mechanism to keep out of the beam window and related apparatus. For the concept with BP, it is observed that the maximum proton beam current is 17 mA. This value is better than the reference case. Moreover, the BP concept is achieved by just introducing BP without driver mechanism. These concepts are useful to adjust the sub-criticality and mitigate the design condition of the beam window.

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Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuel which has large decay heat. So it is necessary to consider the increase of the core temperature when the core cooling system is stopped. This change of the core temperature was evaluated by thermal conduction analysis. It was found that the impact of thermal insulation in the empty rectangular lattice matrix area was large. So it is necessary to verify reliability and accuracy of heat transfer effect used in this area. Testing equipment was fabricated to verify the accuracy of calculation model for the empty lattice matrix which was the free convection model of sealed fluid. By using this equipment, thermal distribution and one dimensional heat flow through the lattice were measured. It was observed that the actual equivalent thermal conductivity in the lattice was larger than the free convection model. It was also confirmed that the insertion of the aluminum block into the empty lattice could achieve the higher equivalent thermal conductivity. These results could be the useful data for the thermal conduction analysis.

DOI： 10.11484/jaea-technology-2015-052

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

The general-purpose reactor analysis code system, MARBLE, has been used to calculate neutron transport and burn-up calculations for Accelerator-Driven System (ADS). In the burn-up calculation of MARBLE, fission product (FP) nuclides had been treated as lump FP in the past. It meant that MARBLE was unable to treat residual nuclides such as rare-earth ones which would be generated by the fuel exchange of the ADS. To treat residual nuclides, ORIGEN2, which was one of the most famous burn-up calculation codes was implemented to MARBLE. By the implementation of ORIGEN2 code, it was available to treat FP nuclides by each nuclide and to consider the residual nuclides in the ADS burn-up calculation.

DOI： 10.11484/jaea-data-code-2015-016

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Nuclear transmutation got much interested as an effective option of nuclear waste management. Japan Atomic Energy Agency (JAEA) proposes the transmutation of minor actinides (MA) by accelerator-driven system (ADS) using lead-bismuth alloy (Pb-Bi). To obtain the data for ADS design, JAEA plans to build a Transmutation Experimental Facility (TEF) in the J-PARC project. TEF consists of two buildings, an ADS target test facility (TEF-T) with 400MeV-250kW Pb-Bi target, and a Transmutation Physics Experimental Facility (TEF-P), which set up a fast critical assembly driven by low power proton beam with MA fuel. In TEF-T, irradiation test for materials, and engineering tests for Pb-Bi target operation will be performed. Various research plans such as nuclear data measurements have been proposed and layout of the experimental hall are underway. In the presentation, roadmap to establish the ADS transmutor and latest design activities for TEF construction will be summarized.

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The Japan Atomic Energy Agency (JAEA) has been conducting the research and development (R\&D) on accelerator-driven system (ADS) as a dedicated transmutation system of long-lived radioactive nuclides. The ADS proposed by JAEA is a lead-bismuth eutectic (LBE) cooled fast subcritical reactor with thermal output of 800 MW. For ADS to play important roles in the nuclear fuel cycle, several critical issues have to be resolved. The critical issue in the design of the ADS is the engineering feasibility of the beam window for the high power spallation target. Although a feasible concept of the beam window for the ADS was proposed in our previous work. To mitigate the thermal load of the beam window not only in the normal operation but also in the transient situations, the R\&D for the mechanism of reactivity adjustment and the accelerator design have been performed in JAEA.

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It is supposed that an Accelerator-Driven System (ADS) is safer than conventional critical reactors since the ADS is driven by the external neutron source in the subcritical state. In this study, the transient analyses for the lead-bismuth cooled ADS proposed by JAEA were performed by using the SIMMER-III and RELAP5/mod3.2 codes to investigate the possibility of the core damage. Three accidents; the protected loss of heat sink, the protected overcooling and the unprotected blockage accident were calculated as the typical ADS accidents. Through these calculations, it was confirmed that all calculation results except the protected loss of heat sink satisfied the no-damage criteria. In the protected loss of heat sink, the cladding tube temperature reached at its melting temperature after 18-21 hours although the calculation condition was very conservative. It is required to design a safety system of the ADS to decrease the frequencies of the accidents and to ease the accidents.

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The effect of experiments using Transmutation Physics Experimental Facility (TEF-P) is analysed from the viewpoint of the reduction of uncertainties in reactor physics parameters (criticality and coolant void reactivity) of an accelerator-driven system (ADS). The analysis is conducted by the nuclear-data adjustment method using JENDL-4.0 on the assumption that ve types of reactor physics experiments (a total of 44 experiments) are performed in TEF-P: (1) criticality experiment, (2) lead void reactivity experiment, (3) reaction rate ratio experiment, (4) sample reactivity experiment, and (5) fuel replacement reactivity experiment. As the result, 1.0\% of uncertainty in criticality is found to be reduced to approximately 0.4\%, and effective experiments for the reduction of uncertainty in criticality and coolant void reactivity are shown to be fuel replacement reactivity experiments and lead void reactivity experiments, respectively. Although these effects depend largely on the composition and amount of minor-actinide (MA) fuels, it is found that a combination of different types of experiments and database of existing experiments is effective in reducing the uncertainties.

DOI： 10.11484/jaea-research-2014-033

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Environment impact, economy and proliferation resistance were estimated for nuclear fuel cycles involving transmutation by fast reactor and accelerator-driven system in equilibrium state. As the result, the transmutation scenario using only fast reactor was superior to the scenarios combined with accelerator-driven system in all estimation, but the differences were insignificant.

DOI： 10.11484/jaea-research-2014-032

CiNii Books

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To investigate an Accelerator-Driven System (ADS) with sub-criticality control mechanism such as control rods or burnable poison, the ADS3D code has been developed on MARBLE which is a next generation reactor analysis code system developed by JAEA. In the past neutronics calculation for the ADS, JAEA employed RZ calculation models to realize efficient investigations. However, it was very difficult to model sub-criticality control mechanisms in RZ calculation models. The ADS3D code system is available to calculate the transportation of protons and neutrons, the burn-up calculation and the fuel exchange in three-dimensional calculation models. It means this code system can treat ADS concepts with sub-criticality control mechanism and makes it possible to investigate a new concept of ADS.

DOI： 10.11484/jaea-data-code-2014-024

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Zeolite is used for adsorption of cesium in the treatment system for contaminated water generated at the Fukushima Daiichi Nuclear Power Plant. Spent zeolite enclosed in stainless steel vessels needs to be stored for a long term; therefore, the hydrogen production by radiolysis of water and corrosion of the vessel under zeolite coexistence needs to be clarified for the safe storage of spent zeolite. The kinetics of cesium adsorption on herschelite, which is used as a zeolite-like adsorbent at the Fukushima site, was investigated to estimate the cesium concentration in spent zeolite vessels. The distribution of cesium in the bed depends on the salt concentration of the feed solution. A simulation code was developed to calculate both the concentration and distribution of $^{134}$Cs and $^{137}$Cs in spent vessels, and the results of calculations were in good agreement with those of experiments.

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Partitioning and Transmutation technology is expected to be effective to mitigate the burden of the HLW disposal. The Japan Atomic Energy Agency (JAEA) has been conducting the research and development (R\&D) on accelerator-driven system (ADS) as a dedicated transmutation system of long-lived radioactive nuclides. For ADS to play important roles in the nuclear fuel cycle, several critical issues have to be resolved. To foster the R\&D for most of above mentioned technical issues, JAEA plans to build the Transmutation Experimental Facility (TEF) which consists of two buildings: the Transmutation Physics Experimental Facility (TEF-P) and the ADS Target Test Facility (TEF-T). TEF-P is a zero-power critical facility which is operated with a low power proton beam to research the reactor physics and the controllability of ADS. Using this TEF-P facility, many experimental studies are planned.

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An Accelerator-Driven System (ADS) has been investigated in Japan Atomic Energy Agency (JAEA) to transmute minor actinides discharged from nuclear power plants. The ADS proposed by JAEA is a lead bismuth eutectic (LBE) cooled-tank-type ADS. It has been known that there was a major upgrade for the cross section data of lead isotopes from JENDL-3.3 to JENDL-4.0 and the upgrade affects to the neutronic design of the ADS. This study aims to measure replacement reactivity from aluminum plates to lead plates at KUCA (Kyoto University Critical Assembly) to know which nuclear data library, JENDL-3.3 or JENDL-4.0 is reasonable for the lead isotopes. As the result, the replacement reactivity from the aluminum plates to the lead ones was measured as the positive value and the calculation results indicated that the lead nuclear data in JENDL-4.0 might be more reasonable than those in JENDL-3.3.

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Attractiveness as a nuclear material was estimated for fresh and spent fuels in nuclear fuel cycles involving transmutation by fast reactor (FR) and accelerator-driven system (ADS) in equilibrium state. As the result, attractiveness of plutonium in FR and neptunium in FR and ADS was relatively high. Moreover, difference of the attractiveness was observed among fuel cycles, but all plutonium can not be exempted from safeguards, and it is important to introduce other factors of proliferation resistance such as legal framework.

DOI： 10.11484/jaea-research-2013-010

CiNii Books

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At the end of the nuclear power generation in the future, there will be considerable amount of nuclear material that will be no longer burnable in the commercial nuclear power reactors. In the transmutation strategy, Pu has to be transmuted prior to MA because of its amount and usability as the nuclear weapon. For the country without the FBR technology, Pu must be transmuted in the LWR dedicated for the transmutation in forms of the MOX fuel or the ROX fuel. If the transmutation systems such as fast reactor or accelerator-driven system are introduced in the nuclear fuel cycle, very small amount of Pu and MA are contained in the high-level waste. In the present study, above mentioned transmutation methods were investigated quantitatively in terms of necessary number of transmutation systems and period after closing the nuclear power with the use of the analysis code for nuclear material flow. Moreover, the comparison of repository were made along the transmutation scenarios.

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In JAEA, research and development on an Accelerator-Driven System (ADS) have been conducted in order to transmute minor actinides (MA) contained in high-level radioactive wastes. One unit of ADS with thermal power of 800 MW can transmute MA generated in 10 units of light water reactors with electric power of 1 GW. In JAEA, we are studying variety of technical areas such as a superconducting proton accelerator, a spallation target including a beam window and a subcritical core including neutronics design and safety assessment. Moreover, construction of the Transmutation Experimental Facility is being planned as a phase-2 program of the J-PARC project to perform basic experiments for transmutation technology.

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Since the accident at Fukushima Daiichi Nuclear Power Station, a large amount of radioactive contaminated water has been generated to cool damaged reactor cores. Adsorption of cesium with zeolite-like media was employed for treatment of this contaminated saline water. As spent zeolite media are highly radioactive, their safe storage is a pressing issue. Japan Atomic Energy Agency has extensively conducted R\&D on the management of secondary wastes produced by the operation of the treatment system. Subjects on the safe storage of spent zeolites include the analysis of their characteristics and the evaluation of effectiveness of the present safety measures in consideration of decay heat emission and hydrogen generation by water radiolysis as well as durability of vessels exposed to saline. Preliminary results obtained are described in the present paper.

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The maximum likelihood analysis method for pulse neutron source experiment (PNS-ML) has been developed in experiments using KUCA to monitor sub-criticality of an accelerator-driven system (ADS). In this technique, the most likely sub-criticality can be deduced from a series of counted data from a detector. In the present study, both sub-criticality and neutron generation time are deduced from the same counted data.

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This document describes the analytical results of the amount of the radioactive nuclides in the Fukushima-Daiichi Nuclear Power Plant on March 31, 2011 and the following period with the use of the ORIGEN2 code. The results are given for the irradiated uranium pellet and the cladding tube of zirconium alloy in the core and the spent fuel storage pools of the respective reactors. The evaluated values are weight, radioactivity, heat generation, photon generation and neutron generation rate.

DOI： 10.11484/jaea-data-code-2012-018

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Nitride fuel cycle for transmutation of long-lived minor actinides (MAs) has been developed in Japan Atomic Energy Agency (JAEA) under the double-strata nuclear fuel cycle concept. This study aims at developing the process flow diagram with the material balance sheet of the pyrochemical reprocessing of spent nitride fuel for ADS to evaluate the technological feasibility of the fuel cycle. Three process flow diagrams were proposed in this work: (1) the currently-proposed process with the molten salt electrorefining of spent nitride fuel, (2) the process with the molten salt electrorefining of the actinide-cadmium alloy after the chemical dissolution of spent nitride fuel, and, (3) the process combining the chemical dissolution of spent nitride fuel and the multi-stage counter current extraction. Moreover, the material balance was evaluated for the process (1) by use of the calculated and experimental data of the nitride fuel for ADS.

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Accuracy of nuclear data, especially for minor actinides (MAs), is important in neutronics design of advanced reactors for MA transmutation such as an Accelerator-Driven System (ADS). Authors introduce the results of a benchmark activity in the IAEA CRP that revealed a discrepancy among k-effectives by different nuclear data libraries is as large as 3\% even for an initial criticality before burn-up. In order to reduce this uncertainty, Transmutation Physics Experimental Facility (TEF-P) with a significant amount of MA is being proposed in the J-PARC project. An expected reduction of the uncertainty of criticality, void reactivity and Doppler coefficient by the TEF-P was numerically evaluated using the cross-section adjustment procedure based on the JENDL3.3. The uncertainty would be reduced from 1.3\% to 0.6\% if experimental results of the TEF-P and other past critical experiments were utilized to adjust nuclear data.

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

An analytical study of minor actinide (MA) transmutation systems was conducted using JENDL-4.0, with a comparison to JENDL-3.3 in terms of reactor physics parameters (criticality, void reactivity and the Doppler reactivity) and those uncertainties. As objects of the analyses, Accelerator driven system (ADS) and MA loaded fast reactor (FR) were assumed. It was found that there were considerable changes for both systems. As the results of the sensitivity and uncertainty analysis, we found that the difference of the parameters of ADS is due mainly to the inelastic scattering cross sections of lead isotopes and several reactions of $^{241}$Am. For FR, a large difference of the void reactivity uncertainty results primarily from the covariance data of the inelastic cross section of $^{23}$Na.

DOI： 10.11484/jaea-research-2011-036

CiNii Books

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Language：English  

R\&D on the transmutation of long-lived minor actinides (MA) by the accelerator-driven system (ADS) using nitride fuels is underway at Japan Atomic Energy Agency. In regard to reprocessing technology, pyrochemical process has several advantages in case of treating spent fuel with large decay heat and fast neutron emission, and recovering highly enriched N-15. In the pyrochemical reprocessing, plutonium (Pu) and MA are dissolved in LiCl-KCl eutectic melts and selectively recovered into liquid cadmium (Cd) cathode by molten salt electrorefining. The recovered Pu and MA are converted to nitrides by the nitridation-distillation combined method, in which the Cd alloys containing Pu and MA are heated in nitrogen gas stream. The authors have investigated its elemental technologies such as electrorefining and renitridation. On the other hand, development of the process flow diagram with the material balance sheet of the pyrochemical reprocessing of spent nitride fuel for ADS is in progress. This paper summarized recent progress of the study which aims to prove the technological applicability of pyrochemical process to the nitride fuel cycle for transmutation of MA.

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To continue the utilization of the nuclear fission energy, the management of the high-level radioactive waste (HLW) 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 (R\&D) on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. The ADS proposed by JAEA is a lead-bismuth eutectic (LBE) cooled fast subcritical reactor with thermal output of 800 MW. For ADS to play important roles in the nuclear fuel cycle, several critical issues have to be resolved. Items of R\&D are divided into three technical areas peculiar to the ADS: (1) superconducting linear accelerator (SC-LINAC), (2) LBE as spallation target and core coolant, and (3) subcritical core design and technology. For these technical areas, various R\&D activities are progressing in JAEA. In this paper, the present status of the R\&D activities in JAEA is reported.

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Waste management of the fast breeder reactor (FBR) fuel cycle with and without partitioning and transmutation (P\&T) was investigated by focusing thermal constraints in storage and disposal facilities. The result showed that transmutation of minor actinides (MAs) is essentially effective to reduce the waste emplacement area in repository, and a combination of P\&T can provide a compact disposal with a smaller emplacement area than the conventional repository design by two orders of magnitude. Cost analysis revealed that the cost for storage and disposal is comparable among the conventional light water reactor, FBR without P\&T, and FBR only with MA-transmutation. The cost of disposal for FBR fuel cycle with P\&T is significantly reduced by an order of magnitude from the others, while that of storage does not increase.

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Language：English  

It is supposed that the Accelerator Driven System (ADS) is safer than other critical systems such as fast reactors since the ADS is driven by the external neutron source in the subcritical state. In this study, the possibility of Core Disruptive Accident (CDA) for the ADS was investigated by detailed safety analyses. Safety analyses for Unprotected Transient Over Power (UTOP) and Unprotected Loss of Flow (ULOF) were carried out by using the SIMMER-III code. The calculation results showed that the maximum temperatures of the nitride fuel and the clad tube (316SS) in the accidents were lower than their melting points (2780$^{\circ}$C and 1400$^{\circ}$C, respectively). It was also found by Level 1 PSA that the frequency of those accidents would be exceedingly small. Although the creep rupture of the clad tube might happen in Beam Over Power (BOP), which was a kind of UTOP, and ULOF, it was concluded that the ADS had very little possibility of CDA or the re-criticality accident.

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Frequent beam trips as experienced in existing high power proton accelerators may cause thermal fatigue problems in ADS components which may lead to degradation of their structural integrity and reduction of their lifetime. Thermal transient analyses were made on the thermal responses of four parts of the reactor components; the beam window, the cladding tube, the inner barrel and the reactor vessel. Our results indicated that the acceptable frequency of beam trips ranged from 50 to $2 \times 10^4$ times per year depending on the beam trip duration. In order to consider measures to reduce the frequency of beam trips on the high power accelerator for ADS, we compared the acceptable frequency of beam trips with the operation data of existing accelerators. The result of this comparison showed that for typical conditions the beam trip frequency for durations of 10 seconds or less was within the acceptable level, while that exceeding five minutes should be reduced to about 1/30 to satisfy the acceptable level.

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After the first check and review on the partitioning and transmutation (PT) technology by the Japan Atomic Energy Commission (JAEC) in 2000, significant progress was made in respective research areas of the partitioning, fuel fabrication, transmutation and fuel recycling in Japan. The second check and review on the PT technology was made by the JAEC in 2008-2009. The final report issued in April, 2009, mentions that the significance of the PT technology is in three points: reduction of the potential hazard, mitigation of the requirement for geological repository site, and enhancement of the options in the design of the whole system of waste disposal. The current technology levels of the PT for both FBR and ADS were evaluated. The PT technology in general is still in the basic research because of the lack of experimental data for minor actinides (MA). It was, therefore, strongly recommended to accumulate the experimental data for MA as a common basis for both FBR and ADS.

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The regression method for pulse neutron source (PNS) experiment has been developed to monitor sub-criticality of an accelerator-driven system (ADS). This method is applicable for PNS experiment with bad experimental conditions, such as high repetition of neutron source that makes the alpha-fitting method difficult to determine delayed neutron level. In the present study, the regression method was verified using the KUCA-A core with various sub-critical levels induced by DT neutrons with several numbers of repetitions. As a result, the discrepancy of $k$$_{eff}$ and count rates between experiment and the regression method was observed for the low repetitions. Further investigation for background is necessary in next study.

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

The potential radiotoxicity of spent nuclear fuel from nuclear power plants is defined as a summation of dose of individual nuclides in case of intake. In the present data base, the dominant nuclides for the potential radiotoxicity were selected from radioactive nuclides in the spent fuel, and dose coefficients to be used to convert radioactivity to dose were tabulated for the selected nuclides. Moreover, time evolution of the potential radiotoxicity was listed taking progenies into account for each actinide. The database enables simple estimation of time evolution of the potential radiotoxicity using an initial weight composition of the spent fuel. The potential radiotoxicity of a spent fuel from the light water reactor is shown as an example of the estimation. The database includes nuclides originated in the spent fuel and does not include those from decommissioning of the nuclear and allied plants.

DOI： 10.11484/jaea-data-code-2010-012

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

The design of an accelerator-driven system was modified and its feasibility was investigated on the basis of new data and knowledge for corrosion by lead-bismuth eutectic and irradiation behavior of candidate materials. The neutronics and thermal design was carried out and a modified core concept was established. As a result of evaluation for the integrity of fuel cladding tubes and a beam window, it was confirmed that their structural strength was sufficient under the conditions of design temperature and corrosion environment without irradiation. Though irradiation effects on material properties were considered not to be so serious under the practical conditions of the ADS according to existing irradiation data, it was needed to accumulate further experimental data for more detailed evaluation. According to level-1 PSA and dynamic analysis of beyond-design-basis accidents, it was shown that there was little possibility of core disruption and re-criticality accidents for ADS.

DOI： 10.11484/jaea-research-2010-012

J-GLOBAL

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Construction of the Transmutation Experimental Facility (TEF) is planned within the framework of J-PARC project. In the current schedule, we are planning to install the critical assembly with a 30kW beam dump at the first phase and the lead-bismuth target facility with superconducting LINAC will be built at the latter phase. Based on this construction plan, JAEA collected the preliminary Letters Of Intent (LOI) for TEF. More than 30 research programs were proposed. According to these proposals, the re-arrangement of facility layout and the new experimental equipment design are now underway. One of the important developments is to prepare for the application of minor actinide fuels to critical assembly. The development of the technology to handle the high-background minor actinide fuel is being performed. Another important issue is the pulse operation of critical assembly under subcritical condition. The laser charge exchange technique and measurement system for pulse operation are also under development. The latest facility concept reflecting these efforts and user requests based on LOI will be introduced.

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Language：English  

To illustrate the benefit of the partitioning and transmutation (PT) technology, several types of the waste management and the geological disposal concepts incorporating PT was discussed in terms of the repository size required to emplace the wastes. It was found that the transmutation of $^{241}$Am, which is a long-term heat source, is effective to prevent the inflation of the repository size expected in the case of plutonium utilization. If we intend to reduce the repository size to a larger extent the partitioning of Sr and Cs followed by their long-term (100 - 300 years) storage should be adopted together with the transmutation of MA.

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The management of americium (Am) and Curium (Cm) in radioactive waste is one of the important issues for sustainable nuclear energy production. A concept of accelerator-driven system using liquid molten-salt fuel is proposed for the effective transmutation of Am and Cm. In the concept, Am and Cm are separated from the spent fuel and converted to chloride molten-salt fuel. Adopting such liquid-type fuel, many defects on Am and Cm such as high radioactivity and heat release and chemical incompatibility can be overcome. The system is an accelerator-driven system consists of a 10MW-class cyclotron, tungsten disk spallation target and subcritical core. The concept of molten-salt fueled accelerator-driven system and dedicated fuel cycle will be introduced.

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The investigations of the beam window which is one of the most critical components for the Accelerator Driven System (ADS) were performed. In the previous study, it was found that the buckling failure was the most severe failure mode for the beam window. Hence, the shape and the thickness of the beam window were optimized to prevent the buckling failure. The buckling analyses with initial imperfections were also performed to identify the level of the factor of safety (FS). The results showed that FS of 3 was conservative enough to ensure the integrity of the beam window. It was also shown that the ellipse shape concepts with the thickness of 2.0-2.4[mm] at the top and the thickness of 2.0-4.0[mm] at the transient part were acceptable under the current ADS design parameters.

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

In this study, the error analysis for transmutation systems was performed. As the transmutation system, Fast Reactor (FR) and Accelerator Driven System (ADS) were employed and the errors caused by the nuclear data for these systems were estimated by using the covariance data prepared in JENDL-3.3. The effect of MA-loaded critical experiments was also investigated by using the cross section adjustment procedure. It was considered that the MA-loaded critical experiment would be performed at Transmutation Physics Experimental Facility (TEF-P) which has been planned its construction in the second phase of J-PARC project. The analysis results presented quantitatively that the errors caused by the nuclear data for these systems were improved by the MA-loaded experiments. The MA-loaded experiment is one of the most useful methods to improve the error caused by the nuclear data which relates to the reliability and the economic efficiency of the transmutation systems.

DOI： 10.11484/jaea-research-2009-033

J-GLOBAL

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

It is supposed that the Accelerator Driven System (ADS) is safer than other critical systems such as Fast Reactors since the ADS is driven by the external neutron source in the subcritical state. In this study, the investigation to confirm the possibility of Core Disruptive Accident (CDA) for the ADS was performed by Level 1 PSA and detailed safety analyses. Abnormal events for the ADS were sorted by Level 1 PSA and calculation cases for the safety analysis were discussed. Based on these results, safety analyses for Unprotected Transient Over Power (UTOP) and Unprotected Loss of Flow (ULOF) were carried out. These results showed that there were hardly possibilities of CDA for the ADS. In the beyond design basis accidents (UTOP and ULOF), it was considered that the creep rupture of the clad tube would happen. However, the frequency of the accidents was exceedingly small. It was concluded that the ADS had very little possibility of CDA or the re-criticality accident.

DOI： 10.11484/jaea-research-2009-024

J-GLOBAL

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

Frequent beam trips as experienced in existing high power proton accelerators may cause thermal fatigue problems in ADS components which may lead to degradation of their structural integrity and reduction of their lifetime. Thermal transient analyses were performed to investigate the effects of beam trips on the reactor components. Our results indicated that the acceptable frequency of beam trips ranged from 50 to $2\times10^{4}$ times per year depending on the beam trip duration. In order to consider measures to reduce the frequency of beam trips on the high power accelerator for ADS, we compared the acceptable frequency of beam trips with the operation data of existing accelerators. The result of this comparison showed that the beam trip frequency for durations of 10 seconds or less was within the acceptable level, while that exceeding five minutes should be reduced to about 1/30 to satisfy the thermal stress conditions.

DOI： 10.11484/jaea-research-2009-023

J-GLOBAL

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Frequent beam trips as experienced in existing high power proton accelerators may cause thermal fatigue problems in ADS components which may lead to degradation of their structural integrity and reduction of their lifetime. Thermal transient analyses were performed to investigate the effects of beam trips on the reactor components, with the objective of formulating ADS design that had higher engineering possibilities and determining the requirements for accelerator reliability. These analyses were made on the thermal responses of four parts of the reactor components; the beam window, the cladding tube, the inner barrel and the reactor vessel. Our results indicated that the acceptable frequency of beam trips ranged from 43 to $2.5\times10^4$ times per year depending on the beam trip duration to keep the plant availability 70\%. In order to consider measures to reduce the frequency of beam trips on the high power accelerator for ADS, we compared the acceptable frequency of beam trips with the operation data of existing accelerators. The result of this comparison showed that the beam trip frequency for durations of 10 seconds or less was within the acceptable level, while that exceeding five minutes should be reduced to about 1/35 to satisfy the plant availability conditions.

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Gas measurements by $\gamma$ spectroscopy in the MEGAwatt PIlot Experiment (MEGAPIE) project has led to the determination of main radioactive isotopes released by the LBE. Comparison with calculations performed with several validated codes supplies important volatile elements release fraction estimation in a spallation target. In addition, calculations with MCNPX2.5.0, FLUKA and SNT codes coupled with evolution programs have been performed in order to study the activation of the target and structural materials. The induced database is relevant for safety and radioprotection during operation, for the post-irradiation experiments and for target dismantlement.

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

The design acceptance of the beam window which is one of the most critical components for the Accelerator Driven System (ADS) was investigated. From the simplified overall assessment for nuclear power plant, it was clarified that the buckling failure was the most severe failure mode for the beam window. The parametric survey for the thickness of the ellipse model was carried out to prevent the buckling failure by using the Finite Element Method code, FINAS. The buckling analyses with initial imperfections were also performed to discuss the factor of safety (FS). The results showed that FS of 3 was enough conservative to ensure the integrity of the beam window. It was also shown that the ellipse shape concepts with the thickness of 2.0$\sim$2.4 [mm] at the top and the thickness of 2.0$\sim$4.0 [mm] at the transient part were acceptable under the current ADS design parameters.

DOI： 10.11484/jaea-research-2008-026

CiNii Books

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In Japan, the partitioning and transmutation (PT) technology is being studied and developed to reduce the burden of the high-level radioactive waste (HLW) management. To demonstrate clearly the benefit of the PT technology on the waste management of future nuclear fuel cycles, the repository area necessitated to dispose of the HLW was discussed quantitatively for spent fuels from UO$_{2}$-LWR, MOX-LWR and MOX-FBR. Four options of separation process were assumed in the analysis: (1) Conventional PUREX reprocessing, (2) Transmutation of minor actinide (MA), (3) Partitioning of FP, and (4) PT for both MA and FP. The results showed that MA transmutation would be necessary to keep the emplacement area for MOX fuel at the same level as that for UO$_{2}$ fuel. The adoption of PT for both MA and FP was effective to further reduce the repository area independently on the fuel type, the reactor type and the cooling period.

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Scopus

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

Japan Atomic Energy Agency (JAEA) promotes research and development of Accelerator-Driven system (ADS) to reduce the burden for conditioning and disposal of the high level radioactive waste by transmuting minor actinide (MA). In the present neutronics design, we investigated several methods to reduce the power peak, and showed the reduction of the temperature of cladding tubes and influence to the beam current. These methods are adjustment of inert matrix content in fuel in each burn-up cycle, multi-region design in terms of plutonium enrichment, pin radius and inert matrix content, modification of the level of the beam window position and height of the central fuel assemblies. As the result, the maximum temperature at the surface of fuel cladding tubes can be reduced by 110 $^{\circ}$C by combining the adjustment of inert matrix content in each burn-up cycle and multi-regin design in terms of pin radius or inert matrix content.

DOI： 10.11484/jaea-research-2007-025

J-GLOBAL

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JAEA has been conducting the R\&D on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. The ADS can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The R\&D activities in JAEA include neutronics, thermal-hydraulic and structural design, transient analysis in case of beam trips, development of spallation target using Pb-Bi, research on the materials, and cost estimation. In addition, the high-intensity proton accelerator project J-PARC is under construction. One of the main experimental facilities of the J-PARC is the Transmutation Experimental Facility (TEF), which is assigned as Phase-II of the project.

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

The double-strata fuel cycle consisting of the commercial fuel cycle and Partitioning-and-Transmutation (P-T) cycle is proposed to reduce the burden for conditioning and disposal of the high level radioactive waste (HLW). The P-T cycle can be adapted to various fuel cycles for the commercial power plants. In the present report, introduction scenarios of P-T cycle were obtained based on mass balances analysis of actinide nuclides for several types of commercial nuclear fuel cycle during the next 200 years. Possible contributions to radioactive waste management were also quantified. The SCENARIO code was developed to calculate the mass balance.

DOI： 10.11484/jaea-research-2006-081

J-GLOBAL

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Event date： 2025.3

Language：Japanese  

Country：Japan  

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Event date： 2025.3

Language：Japanese  

Country：Japan  

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Event date： 2025.2

Language：Japanese  

Country：Japan  

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Event date： 2025.2

Language：Japanese  

Country：Japan  

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Event date： 2025.1

Language：English  

Venue：Tokyo   Country：Japan  

Since the public release of NMB4.0 (Nuclear Material Balance analysis code version 4.0) in March 2022, a nuclear fuel cycle simulator jointly developed by Institute of Science Tokyo (Tokyo Institute of Technology at the time of release) and Japan Atomic Energy Agency, we have engaged in continuous development and support. Over the past three years, the user community has grown to encompass over 30 institutions and more than 200 members. Based on many user feedbacks, 12 updates have been implemented, with the current version being 4.1.2. NMB4.0 has been actively employed in various research projects, international and domestic benchmark, and participation in expert working group at international organizations. In addition to the traditional backend process scenario analysis functions in NMB4.0, recent updates have strengthened capabilities in frontend processes, cycle robustness, actinide management and cost evaluations. This presentation will provide an overview of the recent research and development efforts surrounding NMB4.0, along with introducing new initiatives aimed at advancing the future of nuclear fuel cycle innovation.

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Event date： 2024.9

Language：Japanese  

Country：Japan  

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Event date： 2024.3

Language：Japanese  

Country：Japan  

Database of nuclear power plant and dry reprocessing required for the metal fuel fast reactor cycle was maintained in the nuclear fuel cycle simulator NMB4.0. In addition, the introduction scenario of metal-fueled fast reactors was verified using sample scenarios.

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Event date： 2024.3

Language：Japanese  

Country：Japan  

Many codes have been developed in the world to quantify future nuclear scenarios for the purpose of developing nuclear energy utilization strategies and R\&D goals. However, quality assurance and verification of the accuracy of the codes are not yet fully implemented, with only simple benchmark intercomparisons by the OECD/NEA. In the various evaluation codes, many models are used for each process of the nuclear fuel cycle and nuclear reactors, and there are differences in the definitions of inputs and outputs and databases, so a general verification method has not been established. Therefore, we developed a verification and validation (V\&V) method for quantity evaluation codes to contribute to quality assurance and accuracy verification in the future.

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Event date： 2024.3

Language：Japanese  

Country：Japan  

In the nuclear fuel cycle simulator NMB4.0, developed in collaboration between Tokyo Tech and JAEA, a spent fuel management analysis function has been extended to specify the capacities of multiple interim storage facilities and transportations between them. This presentation introduces the extended function and a result of the reproduction of the spent fuel transportation in Ikata Nuclear Power Plant for the validation of the function.

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Event date： 2024.3

Language：Japanese  

Country：Japan  

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Event date： 2024.3

Language：Japanese  

Country：Japan  

NEUChain (New Chain), a research and development project aiming at the digitization of the nuclear fuel cycle, is being implemented. The basic concept of NEUChain, its application, and the use of distributed ledger technology in the nuclear field will be discussed.

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Event date： 2023.10

Language：English  

Venue：Boulogne-Billancourt   Country：France  

In the case of the direct disposal of MOX spent fuel, the temperature of the buffer material surrounding the waste seems to exceed the limitation due to the decay heat higher than that of UO$_{2}$ spent fuel. To satisfy the limitation, we investigated the impacts of changing some design parameters, and several direct disposal concepts for MOX spent fuel were obtained.

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Event date： 2023.9

Language：Japanese  

Country：Japan  

In the nuclear fuel cycle simulator NMB4.0 being developed at JAEA, a model to numerically calculate the composition of new fuel when recovered uranium is re-enriched has been implemented. Using this function, we evaluated various quantities of re-enrichment of recovered uranium in the case of future nuclear power use in Japan, using the U-235 fraction of depleted uranium and other parameters. In this presentation, we will report the calculation method used in the function and the results of the evaluation of various quantities.

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Event date： 2023.9

Language：Japanese  

Country：Japan  

In order to study the future nuclear power generation and nuclear fuel cycle including geological disposal in Japan, various quantity evaluation codes are used to calculate the time variation of the quantities. The NMB code published by Tokyo Institute of Technology and JAEA was used to evaluate various quantities using the example of the Ikata Nuclear Power Plant, and the code was validated. Simple future projections were also made.

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Event date： 2023.3

Language：Japanese  

Country：Japan  

Since the decay heat of plu-thermal spent fuel is large, it is expected that the disposal conditions for direct disposal such as the emplacement pitch will be more severe than those of UO$_{2}$ spent fuel. In this study, heat transfer analysis around the waste was performed using the finite element method to investigate the disposal conditions under which the maximum buffer material temperature is less than the limit value. As a result, several disposal concepts were obtained, such as extending the cooling period and increasing the area per waste to about 13 times of that for UO$_{2}$ fuel at a disposal depth of 1000 m.

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Event date： 2023.3

Language：Japanese  

Country：Japan  

In the design of Accelerator-Driven System (ADS), a beam window is one of the structures used under severe conditions. Since the maximum temperature of the beam window at rated operation will be more than 500$^{\circ}$C, a creep damage evaluation has been required. Therefore, we have developed a coupled analysis system on ANSYS Workbench to evaluate the creep strain quantitatively. In this system, temperature distribution of the beam window is calculated by the coupled analysis of particle transport and thermal hydraulics analyses, and then is used as an input data for a creep analysis. Calculation result by this analysis system showed that the creep strain after the rated operation was less than 0.1\%.

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Event date： 2023.3

Language：Japanese  

Country：Japan  

We propose Partitioning and Transmutation Complex (PTComplex) aiming for early realization of Partitioning and Transmutation technology. The PTComplex consists of the pyrochemical reprocessing facility and the high intensity proton accelerator. Molten salt in a tank of the pyrochemical reprocessing facility is irradiated with spallation neutrons generated by high intensity protons, and minor actinides in the molten salt are transmuted.

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Event date： 2023.3

Language：Japanese  

Country：Japan  

In order to investigate the management of plu-thermal spent fuel, we compared the various cycle quantities of one-time recycling and two-time recycling, assuming long-term storage, direct disposal, and reprocessing of spent nuclear fuel. As a result, it was found that the volume handled in the front-end (U-mining and enrichment process) is less than that in the twice-recycled process, while the volume handled in the once-recycled process is less than that in many other processes, such as the back-end process.

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Event date： 2022.10

Language：Japanese  

Country：Japan  

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Event date： 2022.9

Language：Japanese  

Country：Japan  

In the design of a reactor using lead-bismuth eutectic (LBE) as a coolant, estimating corrosion rate of structural materials is an important issue in considering safety and the life of the entire structure. Oxide layers formed on the surface of T91 steel have corrosion inhibition effect, however, if they dissolve, corrosion can progress rapidly. Growth and dissolution rates of oxide layers depend on temperature and oxygen concentration, especially under static LBE, a semi-mechanical model of these rates has been developed. However, knowledge of them in flow filed is limited. This study aims to estimate them in flowing LBE by CFD. In this report, results of analysis by ANSYS Fluent code implementing the existing these rates model are described.

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Event date： 2022.9

Language：Japanese  

Country：Japan  

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Event date： 2022.3

Language：Japanese  

Country：Japan  

The integrated nuclear fuel cycle simulation from front- to back-end processes is required for the establishment of the advanced nuclear energy system. However, the existing nuclear fuel cycle simulators, which have been developed by various research institutes, specialized in front-end scenario analysis, and the scope of back-end simulation is limited. In addition, most of the simulators are not open to he public and have not been established as an basic system that can be used by various stakeholders. Tokyo Institute of Technology and JAEA have jointly developed a nuclear material balance analysis code version 4.0 (NMB4.0), which models the entire nuclear fuel cycle from natural uranium mining to nuclide migration for geological disposal. In order to establish the code as a domestically produced nuclear fuel cycle simulator, it has been released free of charge. In this presentation, the general functions and calculation performance of NMB4.0 will be introduced.

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Event date： 2022.3

Language：Japanese  

Country：Japan  

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Event date： 2022.3

Language：Japanese  

Country：Japan  

This study analyzes the nuclear fuel cycle when the partitioning and transmutation technology is introduced in the reprocessing of LWR spent fuel in the latter half of this century. In Japan, LWRs will continue to be used for power generation for the time being, and the treatment and disposal of uranium and plutonium spent fuel will be an issue. Assuming that the partitioning and transmutation of minor actinides becomes feasible in the reprocessing of these spent fuels in the middle of this century, it is shown that the burden of high-level waste disposal after that can be greatly reduced.

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Event date： 2022.3

Language：Japanese  

Country：Japan  

Nuclear power generation will play an important role in achieving carbon neutral in 2050. In order to realize sustainable nuclear power utilization in the future, the rational operation of each process that consists of nuclear fuel cycle and the reduction of radioactive waste are required. Tokyo Institute of Technology (Tokyo Tech) and Japan Atomic Energy Agency (JAEA) have been conducting integrated research in the field of reprocessing and geological disposal, focusing on the NMB code for analyzing various nuclear energy utilization scenarios. In this presentation, the background of this study and the main research results will be presented.

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Event date： 2021.9

Language：Japanese  

Country：Japan  

In order to study the future nuclear energy utilization scenarios, it is necessary to analyze the front-end, reactor operation, and back-end of the nuclear fuel cycle in an integrated manner. On the other hand, the nuclear fuel cycle simulation codes that have been developed in various institutes are specialized in front-end scenario analysis, and it is not to provide user flexible analysis of the back-end simulation. Therefore, Tokyo Institute of Technology and Japan Atomic Energy Agency have jointly developed NMB4.0, a computational platform for integrated and flexible analysis of front-end to back-end scenarios of nuclear fuel cycle. In this presentation, the general functions and computational performance of NMB4.0 were reported.

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Event date： 2021.9

Language：Japanese  

Country：Japan  

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Event date： 2021.3

Language：Japanese  

Country：Japan  

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Event date： 2021.3

Language：Japanese  

Country：Japan  

Tokyo Institute of Technology and Japan Atomic Energy Agency are jointly developing a nuclear material balance code (NMB code), which is a mass balance analysis code for nuclear fuel cycle. The burnup calculation of this code is performed for each reactor or fuel batch using the matrix exponential method. Since the number of burnup calculations is very large depending on the analysis conditions, reducing the calculation cost has been an issue. In particular, to solve the short half-life nuclides by the matrix exponential method, the time step must be sufficiently shorter than the half-life and the calculation must be performed to a high order, which requires computational cost. Therefore, the accelerated exponential method was devised as a method to calculate the short half-life nuclides with enough accuracy for nuclear fuel cycle simulation without incurring computational cost.

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Event date： 2021.3

Language：Japanese  

Country：Japan  

One of the necessary conditions for sustainable reactor training is the diversity of needs for educational training. Educational training is currently open to university undergraduates, graduate students, and nuclear reactor engineers, but if researchers from nuclear reactor development companies and national research institute were to receive educational training, it could contribute greatly to human resource development. In this presentation, we will discuss the effectiveness of practical training for these researchers through the experience of R\&D on accelerator-driven system (ADS) at JAEA.

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Event date： 2021.3

Language：Japanese  

Country：Japan  

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Event date： 2021.3

Language：Japanese  

Country：Japan  

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Event date： 2021.3

Language：Japanese  

Country：Japan  

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Event date： 2020.10

Language：Japanese  

Country：Japan  

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Event date： 2020.5

Language：English  

Venue：Chiba   Country：Japan  

There are many inherent issues in R\&Ds of LBE cooled ADS. To address these issues effectively, we have launched a $\Psi$ (PSi, Proton accelerator-driven Subcritical virtual system) project which construct a virtual ADS on computer. This presentation will introduce an overview and part of details for the plant design part. One of the most important issues for the ADS is a design of a beam window (BW) which is a boundary of an accelerator and a subcritical core. The BW will be used in the severe condition. In the project, a design system for the BW is developed to achieve an efficient design. The design system consists of the FLUENT and ANSYS codes and performs a coupled analysis of thermal hydraulics and structural ones although the irradiation effect is not considered in the analysis. LBE flow in the reactor vessel is also important for a design of coolant system. For the design of coolant system such as PRACS, LBE flow in the reactor vessel has been simulated by the FLUENT code in three-dimensional calculation model.

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Event date： 2020.5

Language：English  

Venue：Chiba   Country：Japan  

Japan Atomic Energy Agency has reorganized past R\&Ds and launched PSi project (Proton accelerator-driven Subcritical virtual system) for R\&D on ADS (Accelerator-Driven System). The PSi project aims efficient R\&D toward industrialization of partitioning and transmutation by Accelerator-Driven System (ADS) by constructing "virtual" ADS on computer. The project is planned for around 10 years and final goal is to be ready for final design and approval application of demo-scale ADS. Overview of the project and some progresses will be introduced in this presentation.

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Event date： 2020.3

Language：Japanese  

Country：Japan  

In order to design an accelerator driven system (ADS), a beam window, which divides between areas of an accelerator and a subcritical core, is one of important structures. JAEA made a 3D model of a beam window and surrounding lead bismuth thermal flow field and developed a coupled analysis system of thermal hydraulics and structure in ANSYS Workbench. In the system a heat generation distribution from PHITS code is read, and steady thermal hydraulic of LBE coolant and stress of the beam window are calculated by Fluent code. Then its eigenvalue buckling is evaluated. In 3D analyses, non-axisymmetric evaluations are considered and radial deformations are found in low buckling modes.

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Event date： 2019.12

Language：Japanese  

Country：Japan  

Takeshita Laboratory, Tokyo Institute of Technology, has been improving the Nuclear Material Balance code (NMB code) developed by Japan Atomic Energy Agency based on the knowledge of the nuclear fuel cycle simulation conducted so far. The previous NMB code was specialized in the evaluation of mass balance of actinide and lacked the function of back-end scenario analysis. In this improvement, 3 main points were researched and developed that (1) selection of FP nuclides to be introduced into the NMB code, (2) development of a nuclear data library preparation tool for the NMB code, and (3) burnup calculation and its error evaluation. In this presentation, we reported the results of the studies in (1) to (3).

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Event date： 2019.10

Language：English  

Venue：Antwerp   Country：Belgium  

JAEA investigates lead-bismuth eutectic alloy (LBE) as a candidate of coolant and spallation target of Accelerator Driven System (ADS). Construction of an experimental facility for LBE spallation target is planned within the J-PARC project for feasibility study of ADS's beam window, which is exposed to high temperature LBE flow and strong proton and neutron irradiation environment. As a demonstration, an LBE target mock-up loop "IMMORTAL" was constructed to confirm the operation of a primary cooling system and to implement integration tests of developing components of LBE technologies. The thermal-hydraulic characteristic of LBE is very important, however, studies about it are still limited compared with that of light water. Therefore, in this research, a calculation model of IMMORTAL's primary loop was made and the thermal-hydraulic behavior was numerically simulated by using the RELAP5-3D code. Then, simulation results and experimental data are compared to validate heat transfer coefficient correlations and properties of LBE.

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Event date： 2019.10

Language：English  

Venue：Antwerp   Country：Belgium  

Japan Atomic Energy Agency has reorganized past R\&Ds and launched PSi project (PSi: Proton accelerator-driven Subcritical virtual system) for R\&D on ADS (Accelerator-Driven System). The PSi project aims efficient R\&D toward industrialization of partitioning and transmutation by Accelerator-Driven System (ADS) by constructing "virtual" ADS on computer. The project is planed for around 10 years and final goal is to be ready for final design and approval application of demo-scale ADS. The project consists of four R\&D fields: reactor physics and nuclear data, ADS plant and safety, lead-bismuth eutectic (LBE) and material, and, accelerator. Each field includes computation code for efficient engineering design, construction of model assembled in the code, and experiments to verify and validate the model. The experiment is not large-scale mock-up but small one using existing facility in Japan and the world.

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Event date： 2019.9

Language：Japanese  

Country：Japan  

In order to discuss the technological development and human resource development necessary for the future nuclear fuel cycle at the research committee on "Feasibility of Nuclear Fuel Cycle," various quantitative analyzes were conducted assuming a wide range of future scenarios. In the evaluation of quantities, the future power generation of LWR and fast reactor, the amount of spent fuel reprocessing, etc. were assumed, and the amount of uranium demand, the accumulation of spent fuel, plutonium, vitrified waste etc. were estimated.

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Event date： 2019.6

Language：English  

Venue：Mol   Country：Belgium  

JAEA investigates lead-bismuth eutectic alloy (LBE) as a candidate of coolant and spallation target of Accelerator Driven System (ADS). Construction of an experimental facility for LBE spallation target is planned within the J-PARC project for feasibility study of ADS's beam window, which is exposed to high temperature LBE flow and strong proton and neutron irradiation environment. As a demonstration, an LBE target mock-up loop "IMMORTAL" was constructed to confirm the operation of a primary cooling system and to implement integration tests of developing components of LBE technologies. One of important subjects is the thermal-hydraulic characteristic of LBE, however, studies about it are still limited compared with that of light water. In this study, numerical thermal-hydraulic simulation of IMMORTAL was implemented by using RELAP5-3D code, and the calculation results were evaluated. A calculation model of the primary LBE loop of IMMORTAL was configurated. In terms of heat balance, except for a module heater (MH) and two heat exchangers (HXs), the adiabatic assumption is considered. The MH have 18 heater rods, but in this model the heat source was simulated as if a rod having the same heat transfer area as the actual. In addition, since RELAP5-3D code is unable to describe directly a configuration of the HXs, positive and negative virtual power source terms were installed to the model. Comparing to the measured values, it overestimated the temperatures in the primary loop by around 20 K. It seems to be caused by the uncertainties in the LBE-side Nusselt correlation and by a possible underestimation of the fin efficiency.

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Event date： 2019.3

Language：Japanese  

Country：Japan  

Treatment of high-level radioactive wastes produced by reactor operation is are very important problem for nuclear energy. A new nuclear data library JENDL/ImPACT-2018 has been developed as a part of the ImPACT project aiming at reduction and resource recycling of LLFPs through nuclear transmutation. Due to possible creation of a lot of secondary products including unstable nuclei, the nuclear reaction data have been evaluated with improving predictability of theoretical calculations.

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Event date： 2019.3

Language：Japanese  

Country：Japan  

In this project, research and development of elemental technologies has been promoted for separation and transmutation. In this research, we integrated HLW volume reduction and resource recycling system and evaluated the overall material balance. In addition, we examined the disposal method of generated waste and the possibility of recycled materials.

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Event date： 2017.9

Language：English  

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

In the ImPACT project, transmutation technology of long-lived fission products (LLFPs) is being developed in addition to minor actinides (MAs) transmutation studied in other programs. If MAs, LLFPs, and heat-generating fission products (FPs), which are Sr-90 and Cs-137, are removed from high-level waste (HLW), drastic benefit is anticipated on the disposal of HLW. In the present study, we tried to estimate two impacts through transport analysis of radionuclide: (1) disposal of the waste after transmutation in the deep underground repository designed for the conventional HLW, and (2) disposal of the waste after transmutation in the intermediate depth disposal designed for the low level radioactive wastes such as hull and end-pieces of the spent fuel assembly. As the result, the reduction of public dose in analysis (1) was observed, and the dose was enough small in analysis (2).

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Event date： 2017.3

Language：Japanese  

Country：Japan  

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Event date： 2016.12

Language：Japanese  

Country：Japan  

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Event date： 2016.10

Language：English  

Venue：Sendai   Country：Japan  

The Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides which will be partitioned from the high level waste. There are various inherent issues for the research and development on the ADS. The recent two activities to realize a feasible and reliable ADS concept are introduced in this paper. For the feasibility, the beam window design is one of the most important issues. To mitigate the design condition of the beam window, namely to reduce the proton beam current, the subcritical core concepts with subcriticality adjustment rods were investigated. For the reliability, the double-accelerator concept was proposed to reduce the beam-trip frequency.

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Event date： 2016.3

Language：Japanese  

Country：Japan  

This presentation shows fabrication and test results of a testing facility for fuel cooling that is a component of the testing facility for remote-handling of highly-radioactive MA fuels in the transmutation physics experimental facility (TEF-P) planned in the J-PARC. Evaluation formula of pressure drop and temperature increase used in the design of TEF-P was validated by the test, and, feasibility of cooling concept was confirmed.

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Event date： 2016.3

Language：Japanese  

Country：Japan  

Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuels which have large radioactivity. So it is necessary to remote handling for loading and taking these MA fuel pins to the TEF-P core. The testing equipment for fuel loading was fabricated with dummy fuel pins and the mock-up core, and performed the loading and taking test. Useful design data was obtained by the fabrication, and it was confirmed to load/take dummy fuel pins from the mock-up core without failure.

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Event date： 2016.3

Language：Japanese  

Country：Japan  

Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuel which has large decay heat. So it is necessary to consider the increase of the core temperature when the core cooling system is stopped. This change of the core temperature was evaluated by thermal conduction analysis. It was found that the impact of thermal insulation in the empty rectangular lattice matrix area was large. Testing equipment was fabricated to verify the accuracy of calculation model for the empty lattice matrix which was the free convection model of sealed fluid. By using this equipment, thermal distribution and one dimensional heat flow through the lattice were measured. It was observed that the actual equivalent thermal conductivity in the lattice was larger than the free convection model. It was also confirmed that the insertion of the aluminum block into the empty lattice could achieve the higher equivalent thermal conductivity. These results could be the useful data for the thermal conduction analysis.

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Event date： 2016.2

Language：English  

Venue：Jukkasjarvi   Country：Sweden  

The Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides (MAs) which will be partitioned from the high level waste. The ADS investigated by JAEA is a lead-bismuth eutectic (LBE) cooled subcritical reactor with 800MW thermal power. One of the most important issues for the ADS design is the design of a beam window which is a boundary of the accelerator and the subcritical reactor. To investigate new beam window concepts, a design condition for the beam window was revised by changing core design to decrease the proton beam current. New core designs with a subcriticality adjustment mechanism such as control rod or burnable poison, have been investigated.

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Event date： 2015.9

Language：Japanese  

Country：Japan  

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Event date： 2015.9

Language：Japanese  

Country：Japan  

To continue the utilization of the nuclear fission energy, the management of the high-level radioactive waste (HLW) is one of the most important issues to be solved. Partitioning and Transmutation (P\&T) 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 research and development on P\&T in Japan are basing on two kinds of concepts: one is the homogeneous recycling of minor actinide (MA) in commercial fast breeder reactors and the other is the dedicated MA transmutation, so-called "double-strata" strategy, using an accelerator-driven system (ADS). The objective of this research program is to enhance the technology revel from basic research phase to semi-engineering research phase for the main component of the double-strata P\&T concept, MA partitioning process and the ADS technologies. For the ADS, the simulation experiments in the critical assembly connecting with the proton accelerator will be carried out to accumulate the reactor experimental results to solve R\&D issues in the reactor physics aspect of the ADS. The results of this R\&D program will be reflected on the assessment of the technically feasibility for each technology.

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Event date： 2015.9

Language：Japanese  

Country：Japan  

It is supposed that an Accelerator-Driven System (ADS) is safer than critical reactors because the ADS is operated in a subcritical state. However, if loss of heat sink would occur, there is a possibility of core damage in the ADS core by the decay heat. This study aims to investigate Decay Heat Removal System (DHRS) which can remove the decay heat with the passive operation. This presentation will introduce the progress of transient analysis code preparation and calculation results which will be used for the DHRS design.

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Event date： 2015.9

Language：Japanese  

Country：Japan  

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Event date： 2015.3

Language：Japanese  

Country：Japan  

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Event date： 2015.3

Language：Japanese  

Country：Japan  

To investigate an Accelerator-Driven System (ADS) with sub-criticality control mechanism such as control rods or burnable poison, the ADS3D code has been developed on MARBLE which is a next generation reactor analysis code system developed by JAEA. In the past neutronics calculation for the ADS, JAEA employed RZ calculation models to realize efficient investigations. However, it was very difficult to model sub-criticality control mechanisms in RZ calculation models. The ADS3D code system is available to calculate the transportation of protons and neutrons, the burn-up calculation and the fuel exchange in three-dimensional calculation models. It means this code system can treat ADS concepts with sub-criticality control mechanism and makes it possible to investigate a new concept of ADS.

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Event date： 2014.12

Language：Japanese  

Country：Japan  

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Event date： 2014.12

Language：English  

Venue：Hefei   Country：China  

The effectiveness of reactor physics experiments using Transmutation Physics Experimental Facility (TEF-P) in J-PARC was analysed from the viewpoint of the reduction of uncertainties in the calculated reactor physics parameters (criticality and coolant void reactivity) of the ADS proposed by JAEA. The analysis was conducted by the nuclear-data adjustment method using JENDL-4.0 on the assumption that several types of reactor physics experiments using minor actinide bearing fuel will be performed in TEF-P. It was found that a combination of various experiments and database of existing experimental data was effective in reducing the uncertainties. As the typical result, 1.0\% of uncertainty in calculated criticality value can be reduced to 0.4\%. As an elemental device development for TEF-P, we developed an online subcriticality monitor. Through experiments using Kyoto University Critical Assembly (KUCA) and the FFAG proton accelerator, it was confirmed that this device could be applicable to measure deep subcriticality around 10{\$} below the critical condition.

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Event date： 2014.10

Language：English  

Venue：Karlsruhe   Country：Germany  

It is supposed that an Accelerator-Driven System (ADS) is safer than conventional critical reactors since the ADS is driven by the external neutron source in the subcritical state. In this study, the transient analyses for the lead-bismuth eutectic (LBE) cooled ADS proposed by JAEA were performed by using the RELAP5/mod3.2 code to investigate the possibility of the core damage. In this research, three accidents; the protected loss of heat sink, the protected overcooling and the unprotected loss of flow were considered as the typical ADS accidents. Through these calculations, it was confirmed that all calculation results except the protected loss of heat sink satisfied the no-damage criteria. In the protected loss of heat sink, if the PRACS (Primary Reactor Auxiliary Cooling System) would not operate correctly, the cladding tube temperature would reach at its melting temperature after 18-21 hours although the calculation condition was very conservative. The transients for the beam trip were also calculated to derive the temperature changes at the important components such as the beam window, the cladding tube and the reactor vessel. These results will be employed for the lifetime estimation of these components.

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Event date： 2014.10

Language：English  

Venue：Karlsruhe   Country：Germany  

Japan Atomic Energy Agency (JAEA) proposes transmutation of minor actinides (MA) by an accelerator-driven system (ADS) using lead-bismuth eutectic (LBE) as a spallation target and a coolant. To obtain the data required for ADS design, JAEA plans to build a Transmutation Experimental Facility (TEF) within the framework of J-PARC project. TEF consists of two buildings, an ADS target test facility (TEF-T), in which a 400 MeV - 250 kW LBE spallation target will be installed, and a Transmutation Physics Experimental Facility (TEF-P), in which a MA-fuelledfast critical/subcritical assembly driven by low power proton beam will be set up. As a major research items of TEF-T, irradiation test for candidate ADS structural materials, engineering tests for LBE target operation and experiments to specify the effective lifetime of proton beam window will be performed. Secondary neutrons from the LBE target are possible to be used for multi-purpose application. The roadmap to establish the ADS transmutor and latest design activities for TEF construction will be summarized.

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Event date： 2014.10

Language：English  

Venue：Bregenz   Country：Austria  

As one of experimental facilities of J-PARC (Japan Proton Accelerator Research Complex), we have a plan to construct the Transmutation Experimental Facility (TEF). TEF consists of two facilities: the ADS Target Test Facility (TEF-T) and the Transmutation Physics Experimental Facility (TEF-P). TEF-T equips a liquid lead-bismuth spallation target bombarded by a 400 MeV - 250 kW proton beam in which candidate proton beam window materials are to be irradiated. TEF-P equips a critical/subcritical assembly to investigate physical and dynamic properties of the accelerator-driven system by using a low power (10W) proton beam. Uranium, Plutonium and minor actinide fuels are planned to be loaded into the assembly. Recent progress in R\&D efforts toward construction of J-PARC TEF will be presented.

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Event date： 2014.9

Language：English  

Venue：Kyoto   Country：Japan  

A burnup calculation is carried out for an accelerator-driven system (ADS) proposed by the Japan Atomic Energy Agency (JAEA) with the fourth version of JENDL, JENDL-4.0 and the previous one, JENDL-3.3. Considerable differences are seen in burnup reactivity between the nuclear data libraries for an initial phase (first burnup cycle) and an equilibrium phase (tenth burnup cycle). The differences in these values are investigated using two methods: a method by replacing a nuclear data library by nuclide and a sensitivity analysis technique. Among many contributors to them for the both phases, we identify major ones; (i) the initial phase: fission cross section and fission neutron multiplicity of $^{238}$Pu, capture cross section of $^{241}$Am, and(ii) the equilibrium phase: capture cross section of $^{244}$Cm and $^{241}$Am, and inelastic scattering cross section of $^{206,207}$Pb. The uncertainty analysis shows that uncertainties in the bunrup reactivity deduced from the JENDL-4.0 covariance data are comparable in magnitude to the differences between the nuclear data libraries, and are dominated by nuclear data parameters of $^{238}$Pu. Finally, we show the necessity of uncertainty evaluation of the branching ratio of $^{241}$Am capture reaction.

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Event date： 2014.9

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  

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Event date： 2014.3

Language：Japanese  

Country：Japan  

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Event date： 2013.12

Language：Japanese  

Country：Japan  

An Accelerator-Driven System (ADS) has been investigated in Japan Atomic Energy Agency (JAEA) to transmute minor actinides discharged from nuclear power plants. The ADS proposed by JAEA is a lead bismuth eutectic (LBE) cooled-tank-type ADS. It has been known that there was a major upgrade for the cross section data of lead isotopes from JENDL-3.3 to JENDL-4.0 and the upgrade affects to the neutronic design of the ADS. For instance, $k_{eff}$ value calculated by JENDL-3.3 was 0.97 and the value calculated by JENDL-4.0 was 1.00 and the main cause of this difference was the cross section data of the lead isotopes. This study aims to measure replacement reactivity from aluminum plates to lead plates to validate the nuclear data of the lead isotopes.

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Event date： 2013.11

Language：English  

Venue：Kyoto   Country：Japan  

The Japan Atomic Energy Agency (JAEA) has been conducting research and development on an accelerator-driven system (ADS) to transmute minor actinides (MAs) included in high-level waste (HLW) into short-lived nuclides. The ADS proposed by JAEA is a lead bismuth eutectic (LBE) cooled-tank-type ADS. The system consists of a high-intensity proton accelerator with 1.5GeV beam energy, a subcritical core with 800MW thermal power and an LBE spallation target placed at a center of the subcritical core. Since the ADS is a hybrid system of an accelerator and a nuclear reactor, there are various inherent issues in the research and development of the ADS. In this study, two important issues are introduced from the viewpoint of the ADS plant investigation and the latest results are presented.

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Event date： 2013.9

Language：Japanese  

Country：Japan  

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Event date： 2013.9

Language：Japanese  

Country：Japan  

The establishment of the plutonium burner system with a High Temperature Gas-cooled Reactor requires the establishment of the method for manufacturing plutonium fuel which has higher nuclear proliferation resistance than the previous method and higher efficiency of plutonium burn-up than LWRs. We discussed on the nuclear proliferation resistance focusing on the plutonium fuel with inert matrix made of yttria-stablized zircona (YSZ). We also discussed on the efficiency of plutonium burn-up by performing three-dimensional burn-up calculations.

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Event date： 2013.9

Language：Japanese  

Country：Japan  

In the scenario of the denuclearization for the energy sources, a fast reactor (FR) would not be assumed. Therefore, the countries which have decided the denuclearization for the energy sources would have a large amount of surplus plutonium. This study aims to investigate the possibility of an accelerator-driven system (ADS) to reduce the surplus plutonium.

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Event date： 2013.9

Language：Japanese  

Country：Japan  

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Event date： 2013.9

Language：Japanese  

Country：Japan  

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Event date： 2013.9

Language：Japanese  

Country：Japan  

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Event date： 2013.3

Language：Japanese  

Country：Japan  

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Event date： 2013.3

Language：Japanese  

Country：Japan  

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Event date： 2013.3

Language：Japanese  

Country：Japan  

Japan Atomic Energy Agency has extensively conducted R\&D on the management of secondary wastes produced by the operation of the system based on the Fukushima Daiichi research and development roadmap. In this paper, to comprehend the temperature distribution and hydrogen diffusion inside the spent zeolite vessel, we report thermal-hydraulic analytical results.

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Event date： 2013.3

Language：Japanese  

Country：Japan  

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Event date： 2012.12

Language：Japanese  

Country：Japan  

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Event date： 2012.12

Language：English  

Venue：Kumatori   Country：Japan  

The transient analyses for the lead-bismuth cooled accelerator-driven system (ADS) were performed with the use of the SIMMER-III and RELAP5/mod3.2 codes to investigate the possibility of the core damage. Five accidents; the beam window breakage, the protected loss of heat sink, the beam overpower, the unprotected loss of flow and the unprotected blockage accident were analyzed as the typical accidents in the ADS. Through these calculations, it was confirmed that all calculation results except the protected loss of heat sink satisfied the no-damage criteria. In the protected loss of heat sink, the cladding tube temperature reached at the melting temperature after 20h although the calculation condition was very conservative. It is required to design a safety system of the ADS to decrease the frequencies of the accidents and to ease the accidents.

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Event date： 2012.10

Language：English  

Venue：Huangshan   Country：China  

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Event date： 2012.9

Language：Japanese  

Country：Japan  

The transient analysis of PLOHS (protected loss of heat sink) for the lead-bismuth cooled Accelerator-Driven System (ADS) was performed with the use of the RELAP5/mod3.2 codes to confirm the possibility of the core damage. The calculation result showed that the cladding tube temperature decreased immediately by the shutdown of the accelerator. After the shutdown, the temperature increased moderately by the loss of heat sink and the temperature reached at the melting temperature of the cladding after 20h. The result indicates that there is little possibility of the core damage but the possibility is not zero. As the future work, it is required to investigate a possibility of the critical accident after the core damage.

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Event date： 2012.9

Language：Japanese  

Country：Japan  

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Event date： 2012.9

Language：Japanese  

Country：Japan  

The proton and neutron irradiation of various structural materials using lead-bismuth eutectic alloy spallation target will be performed at ADS target test facility (TEF-T) which plans to be built within J-PARC project. The irradiation performance of TEF-T was evaluated. The target is arranged as 15 cm in diameter and 60 cm long cylindrical geometry with 2-mm-thick SUS316 casing. DPA and the amount of helium generation by irradiation of proton beam (400 MeV - 250 kW) with the Gaussian beam profile (20 $\mu$A/cm$^{2}$ of peak current density) for 4500 hours were analyzed by Phits 2.24 code. The proton and neutron irradiation of 7.5 DPA are obtained at the maximum, and the irradiation around 7 DPA is possible in the area around 20 mm $\times$ 30 mm. To improve the ADS design, possibility of irradiation by higher proton beam density will be examined.

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Event date： 2012.9

Language：Japanese  

Country：Japan  

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Event date： 2012.9

Language：Japanese  

Country：Japan  

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Event date： 2012.9

Language：Japanese  

Country：Japan  

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Event date： 2012.9

Language：Japanese  

Country：Japan  

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Event date： 2012.9

Language：Japanese  

Country：Japan  

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Event date： 2012.9

Language：Japanese  

Country：Japan  

Use of Rock-like oxide (ROX) fuel is proposed for the purpose of reduction of plutonium amount in the case of gradually stopping nuclear power generation without introducing fast breeder reactors. In this study, the mass flow and repository of three scenarios that is once-through, Pu-thermal and ROX, was analysed. As the result, amount and attractiveness as nuclear weapon of Pu was decreased in the ROX scenario in comparison with the once-through scenario, although the repository size was not reduced.

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Event date： 2012.9

Language：Japanese  

Country：Japan  

The treatment of radioactive water using zeolite adsorbents at Fukushima Dai-ichi Nuclear Power Station produces the waste zeolite. The waste zeolite contains residual water and radioactive elements. The radiolysis of water in the waste zeolite produces molecular hydrogen. The temperature dependence of the hydrogen production was studied by $\gamma$ radiolysis experiment to understand the influence of the temperature rise by decay heat. The result shows that the hydrogen production from the mixture of seawater and zeolite has little temperature dependence. On the other hand the hydrogen production from the mixture of pure water and zeolite increased with increasing temperature. However, the hydrogen production from the pure water mixture was less than that from the seawater mixture. Therefore it is not necessary to take into account the temperature dependence when the hydrogen production is conservatively evaluated using the results obtained from the seawater mixture.

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Event date： 2012.8

Language：English  

Venue：Charlotte  

After the Fukushima accident in 2011, many issues about the nuclear power have been discussed all over the world and some counties have decided the denuclearization for the energy sources. In reality, there are many issues to achieve the denuclearization for the energy sources; for instance, the securement of the electricity by new energy sources and the decommissioning of the existing reactors. It is also supposed that one of the most difficult issues would be the treatment of plutonium discharged from the existing reactors. In such scenario, the denuclearization for the energy sources, a fast reactor would not be assumed. Therefore, the countries which have decided the denuclearization for the energy sources would have a large amount of surplus plutonium. This study aims to investigate the possibility of an accelerator-driven system (ADS) to reduce the surplus plutonium.

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Event date： 2012.7

Language：Japanese  

Country：Japan  

The hydrogen production from the mixture of zeolite and seawater was studied using $\gamma$-radiolysis technique in order to pointed out how the zeolite adsorbents and salts in seawater affect the radiolytic production of hydrogen in the waste zeolite generated in the water treatment at Fukushima Dai-ichi NPS. Two important effects were observed. One is inhibition of the reaction of hydroxyl radical with hydrogen. The measured hydrogen production indicates that hydrogen produced in the mixture totally transferred to the gas phase. The result suggests that the reaction of hydroxyl radical with hydrogen scarcely proceeded. The other is an additional production of hydrogen. The measured amount of hydrogen suggests that the absorbed energy of the zeolite partially transferred to seawater and additionally produced hydrogen. However, the additional production of hydrogen due to the energy transfer is less efficient than the radiolysis of water.

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Event date： 2012.4

Language：Japanese  

Country：Japan  

The accumulated water in Fukushima Dai-ichi Nuclear Power Station contains salts from seawater and is treated by zeolite adsorbents. Thus, radiolysis of seawater and the hydrogen production in the presence of zeolites were studied. Seawater was irradiated by $^{60}$Co $\gamma$-rays. The produced amount of hydrogen and the concentration of hydrogen peroxide were measured. The concentrations of anions were also analyzed. The results show that the oxidation of hydrogen in aqueous phase was inhibited and the hydrogen by radiolysis totally transferred to the gas phase. Then, the hydrogen production in the presence of mordenite-type zeolite and zeolites supplied by KURION was investigated. The results show that the hydrogen production was suppressed by decreasing the amount of seawater. The difference among the zeolites was not significant. Therefore it was suggested that the hydrogen from the waste zeolite can be suppressed by maintaining the water content at low level.

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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： 2012.3

Language：Japanese  

Country：Japan  

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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： 2012.3

Language：Japanese  

Country：Japan  

Analysis method for the pulsed neutron method is newly developed to obtain neutron lifetime with subcriticality based on the maximum likelihoods method. The method was validated by experiments in the KUCA with several subcriticalities injected by a DT neutron source with several repetitions.

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Event date： 2012.3

Language：Japanese  

Country：Japan  

Zeolite adsorbent is used in the decontamination of the radioactive water in Fukushima Dai-ichi Nuclear Power Station. The waste zeolite generated in the decontamination contains radioactive Cs and water. Radiation effects on the waste zeolite are important for the safe storage. Thus, the production of hydrogen peroxide in $\gamma$-radiolysis of the mixture of Y-type synthetic zeolite and aqueous sodium nitrate solution was studied. The measured amount of hydrogen peroxide was lower than that expected from the water content and the radiation dose. Moreover adsorption of hydrogen peroxide on the zeolite was observed. Therefore, the concentration of hydrogen peroxide in the aqueous phase in the mixture was estimated to be significantly low. Because decrease in the concentration of hydrogen peroxide interferes with oxidation of hydrogen in the radiation-induced reactions, hydrogen produced in the mixture is expected not to be decomposed in the radiation-induced reactions.

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Event date： 2012.2

Language：English  

Venue：Tokyo   Country：Japan  

For decontamination of radioactive water containing salts from seawater in Fukushima Dai-ichi Nuclear Power Station, zeolite is used as adsorbent for radioactive Cs. The management of hydrogen produced by radiolysis of water in the zeolite waste is important for safe storage of the waste. Thus hydrogen production from the mixture of mordenite-type zeolite and seawater was studied. The hydrogen yield from the mixture decreased monotonically as the weight fraction of seawater was decreased. However, the measured yield was higher than the yield expected from the weight fraction of seawater, supposing that mordenite has no effect. The result suggests that hydrogen was produced in another pathway besides the direct radiolysis of seawater in the presence of mordenite but its efficiency of the hydrogen production was lower than the direct radiolysis of seawater. The radiation energy deposited to mordenite is expected to be involved in the radiolysis of the adsorbed water.

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Event date： 2011.12

Language：English  

Venue：Hengyang   Country：China  

For the purpose of examining the feasibility and applicability of the lead-bismuth eutectic (LBE) to accelerator-driven transmutation systems, JAEA possesses a static corrosion experimental equipment and four experimental lead-bismuth loops for corrosion/erosion examination, measurement device development and thermal-hydraulic experiments around beam window. In this presentation, the results of static corrosion test for Al-coated steels and present status of fundamental study for flow visualization in opaque liquid metals will be presented. Modification of JAEA Lead-Bismuth material corrosion Loop (JLBL-1) will be also reported.

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Event date： 2011.12

Language：English  

Venue：Bad Honnef   Country：Germany  

The Partitioning and Transmutation (PT) technology aims at mitigation of difficulties arising from spent nuclear fuel (SNF) management. Basic idea of PT technology is to recover long-lived minor actinide (MA) from high-level waste (HLW) produced in a reprocessing plant, and to transmute it into short-lived or stable nuclides. The transmutation of MA is achieved by fast-neutron system such as a fast breeder reactor and an accelerator driven system (ADS). Partitioning of Sr and Cs from HLW can be also effective to reduce the disposal site area for vitrified waste forms. These PT technology is expected to enhance the sustainability of nuclear power utilization.

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Event date： 2011.12

Language：English  

Venue：Hengyang   Country：China  

Frequent beam trips as experienced in existing high-power proton accelerators may cause thermal fatigue in Accelerator-Driven System (ADS) components, which may lead to degradation of their structural integrity and reduction of their lifetime. In this study, acceptable beam-trip frequencies of the ADS accelerator were evaluated and compared with the performance of the ADS accelerator, which was estimated based on the operational data on existing accelerators. Thermal transient analyses were performed to investigate the effects of beam trips on the reactor components, with the objective of determining the feasibility of engineering the ADS and the reliability of the accelerator. Our results indicated three acceptable beam-trip frequencies, depending on the beam-trip duration, $\tau_b$: $2\times10^4$ times per year for $0\leq\tau_b\leq10$s; $2\times10^3$ times per year for $10$ s $<\tau_b\leq 5$ min; and 42 times per year for $\tau_b>5$ min. In order to consider methods to reduce beam-trip frequency, we compared the acceptable beam-trip frequency with the performance of the ADS accelerator, which was estimated based on the operational data on existing accelerators. The comparison showed that for beam trips with a duration of 10 s or less, the beam-trip frequency was acceptable. On the other hand, for beam trips with durations of $10$ s $<\tau_b\leq 5$ min and $\tau_b>5$ min, it was necessary to reduce the beam-trip frequencies to about 1/6 and 1/35, respectively.

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Event date： 2011.12

Language：English  

Venue：Hengyang   Country：China  

Accuracy of nuclear data, especially for minor actinides (MAs), is important in neutronics design of advanced reactors for MA transmutation such as an Accelerator-Driven System (ADS). A benchmark activity to understand current accuracy of neutronics calculation for the ADS was performed in the Coordinated Research Project (CRP) on "Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems" held by the International Atomic Energy Agency (IAEA). Results revealed that a discrepancy among k-effectives by different nuclear data libraries is as large as 3\% even for an initial criticality before burn-up. Summary of the results and improvement plan for nuclear data accuracy will be presented.

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Event date： 2011.10

Language：Japanese  

Country：Japan  

In the Fukushima Daiichi Nuclear Power Station, zeolite is used for decontamination of radioactive water as an adsorbent for radioactive Cs. During the process and the storage of the waste zeolite after the process, the management of hydrogen produced by water radiolysis is an important issue of safety. Thus hydrogen production from the mixture of zeolite and seawater was studied because the radioactive water contains salts from seawater. The hydrogen production from the mixture was decreased monotonically with decreasing weight fraction of seawater. However, the measured yield of hydrogen was higher than the yield expected from the direct radiolysis of water in the mixture. This result shows an additional production of hydrogen as an influence of zeolite. Moreover, dilution of seawater became less effective on the inhibition of hydrogen production with increasing weight fraction of zeolite, although it effectively decreased the hydrogen production in the absence of zeolite.

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Event date： 2011.9

Language：Japanese  

Country：Japan  

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Event date： 2011.9

Language：Japanese  

Country：Japan  

Radioactive water has been emitted following incidents at the Fukushima Dai-ichi Nuclear Power Station, and needed to be quickly decontaminated by the adsorption and/or the coagulation sedimentation of radioactive nuclides, where hydrogen molecules are produced by water radiolysis. The decontamination process with zeolites as absorbents of radioactive cesium was assumed, the observed yields of hydrogen production determined in mixtures of seawater and zeolite, and then the production rates estimated for the safety operation of process before the operation started.

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Event date： 2011.9

Language：Japanese  

Country：Japan  

The amount of hydrogen produced in $\gamma$-radiolysis of mixture of seawater and mordenite was studied. The measured yield was decreased with decrease in the weight fraction of seawater. However, the yield was not simply proportional to the weight fraction of seawater and the influence of the mordenite was observed as additional production of hydrogen. At a weight fraction of seawater more than 0.2, the measured amount of hydrogen could be explained by an empirical equation given by a liner combination of two hydrogen production terms representing hydrogen production due to the energy deposition to seawater and that to mordenite. Each of the terms was expressed as a product of the absorbed energy and a yield given as the proportionality coefficient, and each of the yields could be a constant independent of the mixture ratio. The result imply an instinctive explanation of hydrogen production from the mixture of solid and liquid using the two respective yields.

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Event date： 2011.8

Language：English  

Venue：Warsaw   Country：Poland  

For decontamination of radioactive water, zeolite is a suitable adsorbent for radioactive Cs. The management of hydrogen produced by water radiolysis is an important issue of safety. Thus hydrogen production from the mixture of mordenite-type zeolite and water was studied. In particular, we focused on the influence of the incorporation of seawater into the mixture. The hydrogen yield from the mixture decreased monotonically as the weight fraction of seawater was decreased. However, the measured yield was higher than the yield expected from the direct radiolysis of water in the mixture. Moreover the hydrogen yield was not so effectively decreased by the dilution of seawater in the mixture as in seawater alone. Therefore the additional formation of hydrogen should be considered in the evaluation of hydrogen production from radiolysis of the mixture of mordnite and water.

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Event date： 2011.6

Language：English  

Venue：Hollywood  

The International Atomic Energy Agency (IAEA) is conducting a Coordinated Research Project (CRP) on "Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems". The paper summarizes results obtained for two of the benchmark exercises in the CRP, viz. the transmutation performance analyses of an Accelerator Driven System (ADS) and the analyses of long lived fission product transmutation rates ($^{99}$Tc, $^{127,129}$I).

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Event date： 2011.6

Language：English  

Venue：Hollywood  

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Event date： 2011.3

Language：Japanese  

Country：Japan  

Frequent beam trips as experienced in existing high power proton accelerators may cause thermal fatigue problems in Accelerator-Driven System (ADS) components which may lead to degradation of their structural integrity and reduction of their lifetime. Thermal transient analyses were performed to investigate the effects of beam trips on the reactor components, with the objective of formulating the feasible ADS design and determining the requirements for accelerator reliability. Our results indicated that the acceptable beam-trip frequency was classified by three criteria, according to the beam trip duration, $\tau_b$: $2\times10^4$ times per year for 0 $\leq\tau_b\leq$ 10 s, $2\times10^3$ for 10 s $<\tau_b\leq$ 5 min, and 42 for $\tau_b>$ 5 min. In order to consider measures to reduce the beam trip frequency of the high power accelerator for ADS, we compared the acceptable beam-trip frequency with the operational data of existing accelerators. The result of the comparison showed that the beam trip frequency with durations of 10 s or less was within the acceptable level. On the other hand, for the beam trip with durations of 10 s $<\tau_b\leq$ 5 min and $\tau_b>$ 5 min, it is necessary to reduce the beam trip frequency to about $\frac{1}{6}$ and $\frac{1}{35}$ to satisfy the acceptable level, respectively.

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Event date： 2010.11

Language：English  

Venue：Dimitrovgrad   Country：Russian Federation  

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Event date： 2010.9

Language：Japanese  

Country：Japan  

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Event date： 2010.3

Language：Japanese  

Country：Japan  

The Japan Atomic Energy Agency proposed to construct the Transmutation Experimental Facility (TEF) under a framework J-PARC Project. The TEF consists of two facilities named as Transmutation Physics Experimental Facility (TEF-P) and Accelerator Driven System (ADS) Target Facility (TEF-T). The TEF-P is planned as an experimental facility to study basic reactor physics for the ADS and the reactor for transmutation of minor actinide. The facility is a critical assembly based on existing Fast Critical Assembly (FCA) in JAEA, in which proton beam and external neutron source are available.

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Event date： 2009.11

Language：English  

Venue：Charleston  

The reduction of the storage facility and repository were evaluated for the cases with or without partitioning of Sr-Cs and transmutation of MA in the advanced FBR fuel cycle with the cost estimation for the storage facilities, transporting and repository. As the result, the transmutation of minor actinides is inevitable for the small repository because the heat generation by $^{241}$Am is considerable in FBR. The introduction of partitioning, transmutation and long-term storage enable the very compact layout in the repository like TRU wastes such as hulls and end-pieces of the fuel assembly. The cost for the storage and disposal in this case is much smaller than that in other cases, which mitigates the cost increase by the separation process and transmutor.

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Event date： 2009.11

Language：English  

Venue：Washington, D.C.  

The Transmutation by Adiabatic Resonance Crossing (TARC) experiment was performed in order to demonstrate the efficiency of the neutron Adiabatic Resonance Crossing (ARC) moderation effect in the transmutation of long lived fission products (as $^{99}$Tc and $^{127,129}$I) in a large Lead Block. The high quality of data makes TARC experiment a good benchmark for validation of neutron physics over a broad energy. This article describes the result of the validation for the PHITS code and JENDL/HE file by the TARC experiment. The analysis revealed the neutron flux by the PHITS code is 1.15 times larger than that by the MCNPX code, while both flux are within the experimental error. Flux using JENDL/HE file agreed better to the experimental result than that using the LA-150 nuclear library.

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Event date： 2009.9

Language：Japanese  

Country：Japan  

The investigation of the beam window which is one of the most critical components for the Accelerator Driven System (ADS) were performed from the viewpoint of the proton beam profile change. In this investigation, three distributions, gaussian, parabola and flat were assumed as the proton beam profile. As the results of the neutronic calculation, the thermal hydraulics analysis and the finite element analysis, it was found that the buckling pressures were 4.11 MPa, 4.89 MPa and 4.08 MPa in the gaussian, the parabola and the flat distribution, respectively. These results indicates that the buckling pressure will be improved about 20\% by changing the proton beam profile from the gaussian distribution which is the current reference case to the parabola distribution.

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Event date： 2009.9

Language：English  

Venue：New Delhi   Country：India  

JAEA has been conducting the research and development (R\&D) on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. The ADS proposed by JAEA is a lead-bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the thermal power of 800 MWth driven by a 30 MW superconducting LINAC. The R\&D activities in JAEA are divided into three technical areas peculiar to the ADS: (1) superconducting linear accelerator, (2) LBE as spallation target and core coolant, and (3) subcritical core design and technology. In addition, the Transmutation Experimental Facility (TEF) is planned to be built as Phase-II of the J-PARC project.

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Event date： 2009.8

Language：English  

Venue：Kyoto   Country：Japan  

Japan Atomic Energy Agency (JAEA) develops the accelerator-driven system (ADS), which is a subcritical nuclear reactor dedicated to the transmutation. The design study will be presented about the MA core, beam window, effect of the accelerator trip, lead-bismuth eutectic coolant, and so on. As to the MA core design, the multi-zone design and inert-matrix adjustment at every burn-up stage are adopted to reduce the burn-up swing and power peaking. The thermo-hydraulics analysis and buckling analysis for the ellipse beam window show the soundness in the un-irradiated condition. The acceptable frequency of the beam trip was determined through the fatigue fracture estimation. The flow velocity measurement for lead-bismuth eutectic by the ultrasonic Doppler method is being attempted.

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Event date： 2008.12

Language：English  

Venue：Xi'an   Country：China  

The investigation of the beam window which is a key component in the conceptual design of Accelerator Driven System (ADS) has been performed. In the past study, it was found that the buckling failure due to hydrostatic pressure in the liquid lead bismuth was the most severe failure mode for the beam window from the simplified assessment by "design by analysis" approach for nuclear power plant. Detail structural analyses were explored to find a solution for avoiding instantaneous buckling. The results showed that the ellipse shape concepts were acceptable under the current ADS design parameters. These investigations, however, did not consider the irradiation effect by neutrons and protons. In this study, investigations based on the latest knowledge for the irradiation effect such as the data obtained in spallation target irradiation program (STIP) are presented.

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Event date： 2008.10

Language：English  

Venue：Kobe   Country：Japan  

The managements of recycled actinides and radiological wastes are important issues for sustainable energy production by nuclear power. To solve these problems, the impact and effectiveness of partitioning and transmutation technologies becomes more obvious through the latest studies at JAEA. By the introduction of dedicated minor actinide burnup cycle to power reactor fuel cycle, enhancement of waste disposal capacity and improvement of power reactor economics can be achieved concurrently. At the presentation, current status for waste disposal scenario study using accelerator-driven systems, future plan of J-PARC Transmutation Experimental Facility, and activities at Asian ADS Network initiative will be described.

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Event date： 2008.3

Language：Japanese  

Country：Japan  

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Event date： 2008.2

Language：English  

Venue：Tsuruga   Country：Japan  

A concept of accelerator-driven system using liquid molten-salt fuel is proposed for the effective transmutation of americium (Am) and curium (Cm). In the concept, Am and Cm separated from the spent fuel are converted to chloride molten-salt fuel. Adopting liquid-type fuel, defects on Am and Cm such as high radioactivity and chemical incompatibility can be overcome. In addition, there are many advantages such as omission of fuel disassembling and fabrication. The system is an accelerator-driven system consists of a 10 MW-class cyclotron, tungsten disk spallation target and subcritical core. Molten-salt fuel is reprocessed by dry reprocessing process to separate fission products and adjusting the concentration of Am and Cm. Energy balance and subcriticality of the system is designed to recover overall electricity to drive system itself during the lifetime of the reactor.

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Event date： 2007.9

Language：English  

Venue：Boise  

In order to reveal the impact of the partitioning and transmutation (P\&T) technology on the geological disposal, we investigated the production and disposal of the radioactive wastes from the P\&T facilities including the dry reprocessing for the spent fuel from accelerator-driven system. After classifying the P\&T wastes according to the heat generations, the emplacement configurations in the repository were assumed for each group based on the several disposal concepts proposed for the conventional glass waste form. Then, the sizes of the repositories represented by the vault length, emplacement area and excavation volume were estimated. The repository sizes were reduced by P\&T technology for all disposal concepts.

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Event date： 2007.3

Language：Japanese  

Country：Japan  

Heating nuclides, platinum group metals and actinides are removed from high level waste (HLW) by Partitioning-Transmutation (P-T) technology. The remaining HLW are vitrified into glass waste form as conventional HLW. Since heat of this glass waste are largely reduced comparing with conventional glass waste, we investigate the possibility to emplace compactly in a large tunnel as supposed for TRU wastes.

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Event date： 2007.3

Language：Japanese  

Country：Japan  

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Event date： 2006.10

Language：English  

Venue：Taos  

Irradiation damage to the beam window in the concept of 800 MWth accelerator-driven system is evaluated. Transport of proton and neutron up to 1.5 GeV, which is incident proton energy, is calculated by PHITS code that is the Monte Carlo code for particle and heavy ion and TWODANT code that is 2-dimentional deterministic transport code. The beam window is irradiated at center of the accelerator-driven system by 20 MW proton beam and fast neutron from the core during 300 effective full power days. A displacement, production rate of hydrogen and helium isotopes, and neutron and proton fields is estimated, assuming the realistic beam profiles.

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Event date： 2006.9

Language：Japanese  

Country：Japan  

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Event date： 2006.3

Language：Japanese  

Country：Japan  

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Event date： 2006.3

Language：Japanese  

Country：Japan  

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Event date： 2006.3

Language：Japanese  

Country：Japan  

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