Ibaraki, Japan
Ibaraki, Japan

Time filter

Source Type

Usuda S.,Tohoku University | Usuda S.,NAIS Co. Inc. | Yamanishi K.,Tohoku University | Mimura H.,Tohoku University | And 4 more authors.
Journal of Radioanalytical and Nuclear Chemistry | Year: 2014

Extraction chromatographic separation of trivalent minor actinides (MA: Am and Cm) was investigated by using solid adsorbents impregnating the lipophilic diamide-type ligands, TODGA (N,N,N′,N′-tetraoctyl-3-oxapentane-1,5-diamide) and DOODA(C8) (N,N,N′,N′-tetraoctyl-3,6-dioxaoctane-1,8-diamide). In nitric acid eluents, hydrophilic diamide-type ligands, DOODA(C2) (N,N,N′,N′-tetraethyl-3,6-dioxaoctanediamide) and TEDGA (N,N,N',N'-tetraethyl-diglycolamide) were included for the respective adsorbents as masking agents in order to produce a synergistic effect in mutual separation of Am and Cm. The adsorption tendency of Am and Cm for the adsorbents was opposite each other, that was similar to the case of lanthanides. © 2014 Akadémiai Kiadó, Budapest, Hungary.


Naito Y.,NAIS Co. Inc. | Yamamoto T.,Kyoto University | Misawa T.,Kyoto University | Yamane Y.,Japan Atomic Energy Agency
Journal of Nuclear Science and Technology | Year: 2013

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and MinimumCritical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticalitysafety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS. © 2013 Copyright Taylor and Francis Group, LLC.


Yamamoto T.,Japan Nuclear Energy Safety Organization | Ando Y.,Japan Nuclear Energy Safety Organization | Hayashi Y.,Toshiba Corporation | Azekura K.,NAIS Co. Inc.
Journal of Nuclear Science and Technology | Year: 2012

Critical experiments were performed in the REBUS program on a core loaded with a test bundle including 16 irradiated BWR-type MOX rods of average burnup of 61 GWd/t. The experimental data were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 or JENDL-3.3. Biases in effective multiplication factors of the critical cores were -1.0%Δk for the diffusion calculations (JENDL-3.2), -0.3%Δk for the transport calculations (JENDL-3.3), and 0.2%Δk for the Monte Carlo calculations (JENDL-3.2). The measured core fission rate and co-activation rate distributions were generally well reproduced using the three types of calculations. The burnup reactivity determined using the measured water level reactivity coefficients was -2.41±0.08%Δk/ kk', which also agreed with the results of the three type of calculations within the measurement and calculation errors. The most probable isotopic inventories in the irradiated MOX rods was tentatively obtained by using the ratios of the calculation to chemical assay data on a pellet sample, and the burnup reactivity was reanalyzed to split the calculation error into those due to the inventory and reactivity calculations. This approach showed that the inventory calculation error compensated the reactivity calculation error. © 2012 Atomic Energy Society of Japan. All rights reserved.


Yamamoto T.,Japan Nuclear Energy Safety Organization | Sakai T.,Japan Nuclear Energy Safety Organization | Ando Y.,Japan Nuclear Energy Safety Organization | Liem P.H.,NAIS Co. Inc. | Kikuchi S.,Toshiba Corporation
Journal of Nuclear Science and Technology | Year: 2012

A part of the experimental program FUBILA was dedicated to the study of core physics characteristics of full MOX BWR cores by testing five experimental cores: a core inserted with a B 4C control blade, a core loaded with UO 2 fuel rods, the core loaded with Gd 2O 3-UO 2 fuel rods, a core loaded with 10×10 MOX assemblies and a core loaded with time-elapsed MOX fuel. The present article describes analysis results of the experimental data with deterministic analysis codes and a continuous energyMonte Carlo code coupled with major nuclear data libraries. The calculated critical keff's with the Monte Carlo calculations range from 0.999 to 1.007. Those of the transport calculations with sixteen energy groups are close to those of the Monte Carlo calculations while those of diffusion calculations with the same sixteen energy groups are systematically smaller by -0.3 to -0.5% δk than those of the Monte Carlo calculations. The RMSs of differences between the calculated and measured core radial fission rates are 2 to 3%, 1 to 2%, and 1 to 2% for the diffusion, transport and Monte Carlo calculations, respectively. For the analysis of the Gd 2O 3-UO 2 fuel rod loaded core, the C/Es of the radial fission rates were improved by adopting a detailed lattice calculation model and a newly measured thermal and resonance cross-sections of Gadolinium. © 2012 Atomic Energy Society of Japan. All rights reserved.


Sakurai T.,Japan Atomic Energy Agency | Mori T.,Japan Atomic Energy Agency | Suzaki T.,Japan Atomic Energy Agency | Okajima S.,Japan Atomic Energy Agency | And 3 more authors.
Journal of Nuclear Science and Technology | Year: 2011

The reactivity worths of 22.82 grams of 241Am oxide sample were measured and theoretically analyzed in water-moderated UO2 fuel lattices in seven cores of the Tank-Type Critical Assembly (TCA) at the Japan Atomic Energy Agency for an integral test of 241Am nuclear data. These cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The sample reactivity worth was measured with an uncertainty of 2.1% or less. The theoretical analysis was performed using the JENDL-3.3 nuclear data by a Monte Carlo calculation method. Ratios of calculation to experiment (C/Es) of the reactivity worth were between 0.91 and 0.97, and showed no apparent dependence on the neutron spectrum. In addition, sensitivity analysis based on the deterministic calculation method was carried out to obtain the impact of changing the 241Am capture cross section on the sample reactivity worth. The result of this analysis showed that the C/E could be significantly improved by almost uniformly increasing the 241Am capture cross section of JENDL-3.3 by 25-30%. © 2011 Atomic Energy Society of Japan, All Rights Reserved.


Yamamoto T.,Japan Nuclear Energy Safety Organization | Ando Y.,Japan Nuclear Energy Safety Organization | Hayashi Y.,Toshiba Corporation | Azekura K.,NAIS Co. Inc.
Journal of Nuclear Science and Technology | Year: 2011

Critical experiments of two cores each loaded with fresh 5 x 5 test PWR-type fuel rods of 235U enrichment of 3.8 wt% or irradiated 5 x 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors keff's of the critical cores were about - 1.2%δk for the diffusion calculations (JENDL-3.2), -0.5%δk for the transport calculations (JENDL-3.3), and -0.5 and 0.1%δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was -2.35 ± 0.07Ak/kk'. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel. © 2011 Atomic Energy Society of Japan, All Rights Reserved.


Namekawa M.,NAIS Co. Inc. | Naito Y.,NAIS Co. Inc.
Journal of Nuclear Science and Technology | Year: 2010

A method to judge whether the source iteration in Monte Carlo calculations is converged or not by applying a "sandwich method" is described in this report. A procedure is also introduced to determine the number of skip generations in the source iteration satisfying the convergence criterion. Moreover, the effectiveness of this method is demonstrated by calculating using the fresh fuel vault model in the OECD source convergence benchmark problemss. ©Atomic Energy Society of Japan.


Sembiring T.M.,National Nuclear Energy Agency of Indonesia | Liem P.H.,NAIS Co. Inc.
ARPN Journal of Engineering and Applied Sciences | Year: 2016

An optimum fuel composition is a very important parameter in the operation of a pebble bed high temperature gas-cooled reactor (HTGR). In the present scoping study, the optimum ranges of heavy metal (HM) loading per pebble and the uranium enrichment are investigated. The HM loading range covers 4 to 10 g per pebble, while the uranium enrichment covers 5 to 20 w/o. Two fuel loading schemes typical to pebble-bed HTGRs are also investigated, i.e. the OTTO and multi-pass schemes. All calculations are carried out using BATAN-MPASS, a general in-core fuel management code dedicated for pebble-bed type HTGRs. The reference reactor design case is adopted from the German 200 MWth HTR-Module but with core height of half of the original design. Other design parameters follow the original HTR-Module design. The results of the scoping study show that, for both once-through-then-out (OTTO) and multi-pass fueling schemes, the optimal HM loading per pebble is around 7 g HM/ball. Increasing the uranium enrichment minimizes the fissile loading however higher enrichment than 15 w/o is not effective anymore. The multi-pass fueling scheme shows lower fissile loading requirement and a significantly lower axial power peaking than the OTTO scheme. It can be concluded that the optimum range of HM loading and uranium enrichment are found to be around 7 g per pebble and 15 w/o. In addition the multi-pass fueling scheme shows superior BURNUP and safety characteristics than the OTTO fueling scheme. © 2006-2016 Asian Research Publishing Network (ARPN).


Yamamoto T.,Japan Nuclear Energy Safety Organization | Ando Y.,Japan Nuclear Energy Safety Organization | Hong L.P.,NAIS Co. Inc.
International Conference on the Physics of Reactors 2010, PHYSOR 2010 | Year: 2010

JNES has performed MOX core physics experiments FUBILA using the EOLE LWR critical facility of the CEA Cadarache center in collaboration with a French consortium (CEA and COGEMA). The experiments aim to obtain core physics data of operating conditions of full MOX BWR cores consisting of high Pu-enriched BWR MOX assemblies. One of the experimental cores is a full MOX core containing 10x10 MOX fuel assemblies which have an assembly average total Pu enrichment of 10.6 wt%. Measurement core parameters are critical mass and core fission rate distributions. Theoretical analysis of the experimental data has been carried out using deterministic codes based on diffusion and transport calculations and a continuous energy Monte Carlo calculation code with major nuclear data libraries. The critical keff's are 0.997 for the diffusion calculation and from 1.001 to 1.002 for the transport calculations with JENDL-3.3-base group constants. Those of the Monte Carlo calculations are 1.001 for JENDL-3.3, 0.999 for ENDF/B-VI.8, 1.004 for -VII and 1.001 for JEFF-3.1. The root mean squares (RMS's) of differences between the calculated and measured core radial fission rates are 2.3% for the diffusion calculation, 1.5% for the transport calculation and 1.2% for the Monte Carlo calculation.


PubMed | University of Tsukuba and NAIS Co. Inc.
Type: Journal Article | Journal: Physica medica : PM : an international journal devoted to the applications of physics to medicine and biology : official journal of the Italian Association of Biomedical Physics (AIFB) | Year: 2016

We simulated the effect of patient displacement on organ doses in boron neutron capture therapy (BNCT). In addition, we developed a faster calculation algorithm (NCT high-speed) to simulate irradiation more efficiently.We simulated dose evaluation for the standard irradiation position (reference position) using a head phantom. Cases were assumed where the patient body is shifted in lateral directions compared to the reference position, as well as in the direction away from the irradiation aperture. For three groups of neutron (thermal, epithermal, and fast), flux distribution using NCT high-speed with a voxelized homogeneous phantom was calculated. The three groups of neutron fluxes were calculated for the same conditions with Monte Carlo code. These calculated results were compared.In the evaluations of body movements, there were no significant differences even with shifting up to 9mm in the lateral directions. However, the dose decreased by about 10% with shifts of 9mm in a direction away from the irradiation aperture. When comparing both calculations in the phantom surface up to 3cm, the maximum differences between the fluxes calculated by NCT high-speed with those calculated by Monte Carlo code for thermal neutrons and epithermal neutrons were 10% and 18%, respectively. The time required for NCT high-speed code was about 1/10th compared to Monte Carlo calculation.In the evaluation, the longitudinal displacement has a considerable effect on the organ doses. We also achieved faster calculation of depth distribution of thermal neutron flux using NCT high-speed calculation code.

Loading NAIS Co. Inc. collaborators
Loading NAIS Co. Inc. collaborators