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Journal Articles

Fukushima $$^{137}$$Cs releases dispersion modelling over the Pacific Ocean; Comparisons of models with water, sediment and biota data

Peri$'a$$~n$ez, R.*; Bezhenar, R.*; Brovchenko, I.*; Jung, K. T.*; Kamidaira, Yuki; Kim, K. O.*; Kobayashi, Takuya; Liptak, L.*; Maderich, V.*; Min, B. I.*; et al.

Journal of Environmental Radioactivity, 198, p.50 - 63, 2019/03

A number of marine radionuclide dispersion models were applied to simulate $$^{137}$$Cs releases from Fukushima Daiichi Nuclear Power Plant accident in 2011 over the northwest Pacific. Simulations extended over two years and both direct releases into the ocean and deposition of atmospheric releases on the ocean surface were considered. Dispersion models included an embedded biological uptake model (BUM). Three types of BUMs were used: equilibrium, dynamic and allometric. Model results were compared with $$^{137}$$Cs measurements in water, sediment and biota. A reasonable agreement in model/model and model/data comparisons was obtained.

Journal Articles

Status of JMTR decommissioning plan formulation, 2

Otsuka, Kaoru; Ide, Hiroshi; Nagata, Hiroshi; Oomori, Takazumi; Seki, Misaki; Hanakawa, Hiroki; Nemoto, Hiroyoshi; Watanabe, Masao; Iimura, Koichi; Tsuchiya, Kunihiko; et al.

UTNL-R-0499, p.12_1 - 12_8, 2019/03

no abstracts in English

JAEA Reports

Applicability of statistical geometry model to light water moderating systems

Mori, Takamasa; Kojima, Kensuke*; Suyama, Kenya

JAEA-Research 2018-010, 57 Pages, 2019/02

JAEA-Research-2018-010.pdf:6.25MB

In order to estimate applicability of the statistical geometry model (STGM) of MVP/GMVP, a parametric study in infinite geometry and criticality safety analyses for direct disposal of spent fuel in simple finite geometry have been carried out by using the MVP Monte Carlo code. It has been found that calculations with STGM for larger fuel spheres give larger thermal utilization factors and larger infinite multiplication factors compared with explicit random models in the range of fuel sphere packing fraction between 6.5 % and 63.3 %. Substantial differences are not observed between the results with two nearest neighbor distributions (NNDs); that given by the MCRDF code and the analytical expression based on a statistically uniform distribution. It is inferred that the overestimation by STGM is caused by the facts that STGM cannot take account of the surroundings of each neutron, whether a fuel sphere rich region or a water moderator rich one, because STGM always uses an NND averaged over such surroundings and that STGM, therefore, cannot take the effect of consecutive scatterings in the water moderator into account.

JAEA Reports

Review of research on advanced computational science in FY2017

Center for Computational Science & e-Systems

JAEA-Evaluation 2018-002, 32 Pages, 2019/02

JAEA-Evaluation-2018-002.pdf:1.09MB

Research on advanced computational science for nuclear applications, based on "Plan to Achieve Medium to Long-term Objectives of the Japan Atomic Energy Agency (Medium to Long-term Plan)", has been performed at Center for Computational Science & e-Systems (CCSE), Japan Atomic Energy Agency. CCSE established the committee consisting outside experts and authorities which does research evaluation and advices for the assistance of the research and development. This report summarizes the followings. (1) Results of the R&D performed at CCSE in FY 2017 (April 1st, 2017 - March 31st, 2018), (2) Results of the evaluation on the R&D by the committee in FY 2017

JAEA Reports

Comparison of potential radiotoxicity of actinide elements; Data for consideration of optimum recovery of actinide elements

Morita, Yasuji; Nishihara, Kenji; Tsubata, Yasuhiro

JAEA-Data/Code 2018-017, 32 Pages, 2019/02

JAEA-Data-Code-2018-017.pdf:2.35MB

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.

Journal Articles

Determination of $$^{107}$$Pd in Pd purified by selective precipitation from spent nuclear fuel by laser ablation ICP-MS

Asai, Shiho; Ohata, Masaki*; Yomogida, Takumi; Saeki, Morihisa*; Oba, Hironori*; Hanzawa, Yukiko; Horita, Takuma; Kitatsuji, Yoshihiro

Analytical and Bioanalytical Chemistry, 411(5), p.973 - 983, 2019/02

Determination of radiopalladium $$^{107}$$Pd is required for ensuring the radiation safety of Pd extracted from spent nuclear fuel for recycling or disposal. We employed laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) to simplify an analytical procedure of $$^{107}$$Pd. Pd was separated through selective Pd precipitation reaction from spent nuclear fuel. Laser ablation allows direct measurement of the Pd precipitates, skipping the dissolution and dilution procedure. In this study, $$^{102}$$Pd in natural Pd standard solution was used as an internal standard, taking advantage of its absence in spent nuclear fuel. The Pd precipitate was uniformly embedded on the surface of the centrifugal filter, forming a microscopically thin flat surface of Pd. The resulting homogeneous Pd layer is suitable for obtaining a stable signal ratio of $$^{107}$$Pd/$$^{102}$$Pd. The amount of $$^{107}$$Pd obtained by LA-ICP-MS corresponds to the values obtained by conventional solution nebulization measurement.

Journal Articles

Implementation of a gyrokinetic collision operator with an implicit time integration scheme and its computational performance

Maeyama, Shinya*; Watanabe, Tomohiko*; Idomura, Yasuhiro; Nakata, Motoki*; Nunami, Masanori*

Computer Physics Communications, 235, p.9 - 15, 2019/02

 Percentile:100(Computer Science, Interdisciplinary Applications)

We have implemented the Sugama collision operator in the gyrokinetic Vlasov simulation code, GKV, with an implicit time-integration scheme. The new method is versatile and independent of the details of the linearized collision operator, by means of an operator splitting, an implicit time integrator, and an iterative Krylov subspace solver. Numerical tests demonstrate stable computation over the time step size restricted by the collision term. An efficient implementation for parallel computation on distributed memory systems is realized by using the data transpose communication, which makes the iterative solver free from inter-node communications during iteration. Consequently, the present approach achieves enhancement of computational efficiency and reduction of computational time to solution simultaneously, and significantly accelerates the total performance of the application.

Journal Articles

Abnormally enhanced diamagnetism in Al-Zn-Mg alloys

Nishimura, Katsuhiko*; Matsuda, Kenji*; Lee, S.*; Nunomura, Norio*; Shimano, Tomoki*; Bendo, A.*; Watanabe, Katsumi*; Tsuchiya, Taiki*; Namiki, Takahiro*; Toda, Hiroyuki*; et al.

Journal of Alloys and Compounds, 774, p.405 - 409, 2019/02

Journal Articles

Unified description of the fission probability for highly excited nuclei

Iwamoto, Hiroki; Meigo, Shinichiro

Journal of Nuclear Science and Technology, 56(2), p.160 - 171, 2019/02

 Percentile:100(Nuclear Science & Technology)

We present a new model to describe the fission probability of the high-energy fission model, as deduced from the intranuclear cascade calculation with the Intra-Nuclear Cascade model of Li$`{e}$ge (INCL) version 4.6 and Prokofiev's phenomenological systematics of the proton-induced fission cross sections. This model is implemented in the de-excitation model of the Generalized Evaporation Model (GEM), and applied to Monte Carlo spallation reaction simulation using the Particle and Heavy Ion Transport code System (PHITS). Comparing with experimental data for subactinide nuclei shows that this model can provide a unified prediction of the proton-, neutron-, and deuteron-induced fission cross sections with markedly improved accuracy. The calculated fission fragments tend to shift to higher mass numbers. To account for the isotopic distributions of fission fragments within the framework of a coupled INCL/GEM, modification of INCL is required, especially for description of the highly-excited states of residual nuclei.

Journal Articles

Report on the IAEA Technical Meeting "Nuclear Data Processing"

Tada, Kenichi

Kaku Deta Nyusu (Internet), (122), p.9 - 21, 2019/02

This paper reports the overview of the technical meeting of nuclear data processing in IAEA to Japanese researchers. In this technical meeting, the current status of nuclear data processing codes and verification of them are described.

Journal Articles

Summary report on Consultants' Meeting of INDEN on the Evaluated Nuclear Data of the Structural Materials

Iwamoto, Nobuyuki

Kaku Deta Nyusu (Internet), (122), p.5 - 8, 2019/02

This is a summary report on Consultants' Meeting of INDEN on the Evaluated Nuclear Data of the Structural Materials hosted by IAEA.

Journal Articles

3rd RCM on Updating the Photonuclear Data Library and Generating a Reference Database for Photon Strength Functions

Utsunomiya, Hiroaki*; Iwamoto, Nobuyuki; Kawano, Toshihiko*

Kaku Deta Nyusu (Internet), (122), p.26 - 32, 2019/02

no abstracts in English

Journal Articles

Development of next generation nuclear data processing code FRENDY

Tada, Kenichi

Robutsuri No Kenkyu (Internet), (71), 13 Pages, 2019/02

The nuclear data processing is very important to connect between the evaluated nuclear data library and the particle transport calculation code. However, many nuclear engineers do not know well about the nuclear data processing. This paper describes the overview of nuclear data processing and our nuclear data processing code FRENDY. This paper also lists references about the nuclear data processing.

Journal Articles

Optimization of mechanical properties in aluminum alloys $$via$$ hydrogen partitioning control

Toda, Hiroyuki*; Yamaguchi, Masatake; Matsuda, Kenji*; Shimizu, Kazuyuki*; Hirayama, Kyosuke*; Su, H.*; Fujiwara, Hiro*; Ebihara, Kenichi; Itakura, Mitsuhiro; Tsuru, Tomohito; et al.

Tetsu To Hagane, 105(2), p.240 - 253, 2019/02

no abstracts in English

JAEA Reports

Nuclear data processing code FRENDY version 1

Tada, Kenichi; Kunieda, Satoshi; Nagaya, Yasunobu

JAEA-Data/Code 2018-014, 106 Pages, 2019/01

JAEA-Data-Code-2018-014.pdf:1.76MB
JAEA-Data-Code-2018-014-appendix(DVD-ROM).zip:6.99MB

A new nuclear data processing code FRENDY has been developed in order to process the evaluated nuclear data library JENDL. Development of FRENDY helps to disseminate JENDL and various nuclear calculation codes. FRENDY is developed not only to process the evaluated nuclear data file but also to implement the FRENDY functions to other calculation codes. Users can easily use many functions e.g., read, write, and process the evaluated nuclear data file, in their own codes when they implement the classes of FRENDY to their codes. FRENDY is coded with considering maintainability, modularity, portability and flexibility. The processing method of FRENDY is similar to that of NJOY. The current version of FRENDY treats the ENDF-6 format and generates the ACE file which is used for the continuous energy Monte Carlo codes such as PHITS and MCNP. This report describes the nuclear data processing methods and input instructions for FRENDY.

Journal Articles

Count-loss effect in determination of prompt neutron decay constant by neutron correlation methods that employ two sets of neutron counting systems

Kitamura, Yasunori*; Fukushima, Masahiro; Kitamura, Yasunori*

Annals of Nuclear Energy, 125, p.328 - 341, 2019/01

It has been taken for granted that the neutron correlation methods that employ two sets of neutron counting systems, e.g., the covariance-to-mean and the cross-correlation methods, are free from the count-loss effect for determination of the neutron decay constant. It was however found in the present study that these methods overestimate the neutron decay constant under high counting rate conditions. New formulae of these methods were hence obtained on the basis of a rigorous theoretical approach for treating the count-loss process. It is expected that the present formulae work better than conventional ones for determination of the neutron decay constant.

Journal Articles

First-principles calculation of multiple hydrogen segregation along aluminum grain boundaries

Yamaguchi, Masatake; Ebihara, Kenichi; Itakura, Mitsuhiro; Tsuru, Tomohito; Matsuda, Kenji*; Toda, Hiroyuki*

Computational Materials Science, 156, p.368 - 375, 2019/01

 Times Cited Count:1 Percentile:100(Materials Science, Multidisciplinary)

The segregation of multiple hydrogen atoms along aluminum (Al) grain boundaries (GBs) and fracture surfaces (FSs) was investigated through first-principles calculations considering the characteristics of GBs. The results indicate that hydrogen segregation is difficult along low-energy GBs. The segregation energy of multiple hydrogen atoms along GBs and FSs and the cohesive energy was obtained for three types of high-energy Al GBs. With increasing hydrogen segregation along the GBs, the cohesive energy of the GB decreases and approaches zero with no decrease in GB segregation energy. The GB cohesive energy decreases in parallel with the volume expansion of the region of low electron density along the GB.

Journal Articles

Transient ionization of the mesosphere during auroral breakup; Arase satellite and ground-based conjugate observations at Syowa Station

Kataoka, Ryuho*; Nishiyama, Takanori*; Tanaka, Yoshimasa*; Kadokura, Akira*; Uchida, Herbert Akihito*; Ebihara, Yusuke*; Ejiri, Mitsumu*; Tomikawa, Yoshihiro*; Tsutsumi, Masaki*; Sato, Kaoru*; et al.

Earth, Planets and Space (Internet), 71, p.9_1 - 9_10, 2019/01

Transient ionization of the mesosphere was detected at around 65 km altitude during the isolated auroral expansion occurred at 2221-2226 UT on June 30, 2017. A general-purpose Monte Carlo particle transport code PHITS suggested that significant ionization is possible in the middle atmosphere due to auroral X-rays from the auroral electrons of $$<$$10 keV.

Journal Articles

Development of a stochastic biokinetic method and its application to internal dose estimation for insoluble cesium-bearing particles

Manabe, Kentaro; Matsumoto, Masaki*

Journal of Nuclear Science and Technology, 56(1), p.78 - 86, 2019/01

 Percentile:100(Nuclear Science & Technology)

If an insoluble cesium-bearing particle is incorporated into the human body, the radioactivity will move as a single particle. In this case, it is impossible to estimate the number of disintegrations by considering the average behavior of countless nuclei. Then, a method was developed to simulate the behavior of the particle stochastically; and a biokinetic model was constructed to consider the characteristics of insoluble particles. Combination of the method and the model enables to estimate the number of disintegrations, and consequently the internal doses considering the stochastic behavior of the single cesium particle. We evaluated a probability density function of committed equivalent and effective doses and its 99th percentile value and arithmetic mean by repeating the above described procedure, and compared them to the reference values based on the existing models. As a result, the 99th percentile value of committed effective doses was 70 times the reference value when the number of incorporated particles was one, and consequently the dose level was quite low. When the exposure level is 1 mSv in committed effective dose, the uncertainty originating in the insolubility of cesium particles was negligibly small.

Journal Articles

Measurements of gamma-ray emission probabilities in the decay of americium-244g

Nakamura, Shoji; Terada, Kazushi*; Kimura, Atsushi; Nakao, Taro*; Iwamoto, Osamu; Harada, Hideo; Uehara, Akihiro*; Takamiya, Koichi*; Fujii, Toshiyuki*

Journal of Nuclear Science and Technology, 56(1), p.123 - 129, 2019/01

Accurate data of $$gamma$$-ray emission probabilities are frequently needed when one quantitatively determines the amount of isotope by $$gamma$$-ray measurements or obtains neutron capture cross-sections using them. Americium-243, one of the most important minor actinides, produces $$^{244}$$Am after neutron capture. The 744-keV $$gamma$$-ray decaying from the ground state of $$^{244}$$Am has a relatively large $$gamma$$-ray emission probability c.a. 66%, however, its uncertainty is as large as 29%. The uncertainty of the $$gamma$$-ray emission probability leads to a major factor of the systematic uncertainty on determining an amount of isotope, and therefore the $$gamma$$-ray emission probability was measured by using an activation method and an examined level structure of $$^{244}$$Cm. In this study, the emission probability of 744-keV $$gamma$$ ray was derived as 66.5$$pm$$1.1%, and its uncertainty was improved from 29% to 2%.

9157 (Records 1-20 displayed on this page)