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Journal Articles

Chemical form consideration of released fission products from irradiated fast reactor fuels during overheating

Sato, Isamu; Tanaka, Kosuke; Koyama, Shinichi; Matsushima, Kenichi*; Matsunaga, Junji*; Hirai, Mutsumi*; Endo, Hiroshi*; Haga, Kazuo*

Energy Procedia, 82, p.86 - 91, 2015/07

 Times Cited Count:2 Percentile:22.25(Nuclear Science & Technology)

Experiments simulating overheating conditions of fast reactor severe accidents have been previously carried out with irradiated fuels. For the present study, the chemical forms of the fission products (FPs) included in the irradiated fuels were evaluated by thermochemical equilibrium calculations. At temperatures of 2773 K and 2973 K, the most stable forms of Cs, I, Te, Sb, Pd and Ag are gaseous compounds. Cs and Sb detected in the thermal gradient tube (TGT) in the experiments can take gaseous chemical forms of elemental Cs, CsI, Cs$$_{2}$$MoO$$_{4}$$, CsO and elemental Sb, SbO, SbTe, respectively. By comparing experimental results and the estimations, it is seen CsI thermochemically behaves in a manner that traps it in the TGT, while elemental Cs trends to move as fine particles. The moving behavior of the gaseous FPs will obey not only thermochemical principles, but also those of particle dynamics.

Journal Articles

Azimuthal flux distribution measurements around fuel rods in reduced-moderation LWR lattices

Yoshioka, Kenichi*; Kitada, Takanori*; Nagaya, Yasunobu

Progress in Nuclear Energy, 82, p.7 - 15, 2015/07

 Times Cited Count:1 Percentile:12.03(Nuclear Science & Technology)

A reduced-moderation LWR has been developed for the reduction of spent fuel and for the efficient utilization of uranium resources. The streaming channel concept to improve the negative void reactivity coefficient is one of the features of the reactor. This concept makes the fuel assembly more heterogeneous. The geometrical heterogeneity makes azimuthal neutron flux distribution of fuel rods steep. To validate azimuthal neutron flux distribution calculation, we measured the distribution around fuel rods in reduced moderation LWR lattices. These measurements were conducted in NCA with the foil activation method. The core consisted of the central triangular tight lattice zone and the outer driver zone of a square lattice. A pile of polystyrene plates for simulating void fraction was installed into the triangular tight lattice. Azimuthal neutron flux distributions were deduced from the activity of these small foils measured with plastic scintillators. Measurements were compared to calculations by the MVP code with JENDL-3.3. It was found that calculations agreed well with measurements.

Journal Articles

An Analysis of $$beta$$-delayed neutron emission of even-even neutron-rich nuclei with proton-neutron QRPA

Minato, Futoshi; Iwamoto, Osamu

Progress in Nuclear Energy, 82, p.112 - 117, 2015/07

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

$$beta$$-decay and $$beta$$-delayed neutron emission of neutron-rich spherical nuclei are investigated. Our formalism adopts a self-consistent QRPA approach for $$beta$$-decay and Hauser-Feshbach statistical model for particle evaporation from highly excited state of daughter (precursor) nucleus. In this work, we particularly pay attention to the effects of two contributions. One is tensor force, which is not taken into account in conventional self-consistent QRPA but is important for reproducing half-lives of closed-shell nuclei. And another is isospin $$T=0$$ finite range pairing. They play a significant role to reduce energy of excited state of precursor nuclei. We found that these effects reduce the number of decay branches above neutron threshold of precursor nuclei and consequently a predicted $$beta$$-delayed neutron yields become smaller than that without them. This work is planned to apply to nuclear data evaluation of $$beta$$-delayed neutron yield of fission fragments in future.

Journal Articles

Oxidation and carburizing of FBR structural materials in carbon dioxide

Furukawa, Tomohiro; Rouillard, F.*

Progress in Nuclear Energy, 82, p.136 - 141, 2015/07

 Times Cited Count:23 Percentile:92.84(Nuclear Science & Technology)

The application of a supercritical carbon dioxide (SC-CO$$_{2}$$) turbine cycle to fast rectors has the potential to enhance reliability because the SC-CO$$_{2}$$ turbine system is capable of replacing the steam generator turbine system of conventional sodium cooled fast reactors. A key problem in the application is the corrosion of structural material by SC-CO$$_{2}$$ at high temperatures. The authors have performed corrosion test on high-chromium martensitic and austenitic stainless steels in CO$$_{2}$$ under the pressure conditions from atmospheric pressure to 25 MPa at elevated temperature, and proposed corrosion allowances of the steels for preliminary design of the SC-CO$$_{2}$$ system. This paper initially reports the results of metallurgical examination of the steels after 8010 hours in SC-CO$$_{2}$$ which is the longest immersion data in our experiments, and then describes the behavior of the oxide growth from the view point of estimation of the corrosion allowance for the design.

Journal Articles

Safety design consideration for HTGR coupling with hydrogen production plant

Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

Progress in Nuclear Energy, 82, p.46 - 52, 2015/07

 Times Cited Count:8 Percentile:65.81(Nuclear Science & Technology)

Safety requirements and design considerations for a HTGR hydrogen production system by IS process are examined. Requirements in order to construct hydrogen production plants under conventional chemical plant regulation are identified. In addition, safety requirements for the collocation of the nuclear facility and hydrogen production plant utilizing IS process are investigated. Furthermore, design considerations to comply with the requirements are suggested and the technical feasibility of the design considerations is evaluated. The evaluation results clarified that design considerations suggested for coupling IS plant to HTGR are reasonably practicable.

Journal Articles

Photonuclear reactions of calcium isotopes calculated with the nuclear shell model

Utsuno, Yutaka; Shimizu, Noritaka*; Otsuka, Takaharu*; Ebata, Shuichiro*; Homma, Michio*

Progress in Nuclear Energy, 82, p.102 - 106, 2015/07

 Times Cited Count:3 Percentile:32.14(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Removal of zirconium from spent fuel solution by alginate gel polymer

Onishi, Takashi; Koyama, Shinichi; Mimura, Hitoshi*

Progress in Nuclear Energy, 82, p.69 - 73, 2015/07

 Times Cited Count:4 Percentile:40.75(Nuclear Science & Technology)

no abstracts in English

Journal Articles

PHITS simulation of quasi-monoenergetic neutron sources from $$^7$$Li($$p$$,$$n$$) reactions

Hashimoto, Shintaro; Iwamoto, Osamu; Iwamoto, Yosuke; Sato, Tatsuhiko; Niita, Koji*

Energy Procedia, 71, p.191 - 196, 2015/05

 Times Cited Count:2 Percentile:88.64

Accelerator-based neutron sources using proton- and deuteron-induced reactions have been utilized for scientific and medical applications, such as irradiation testing of fusion reactor materials at IFMIF and BNCT. Quasi-monoenergetic neutron beams using $$^7$$Li($$p$$,$$n$$)$$^7$$Be are of special importance for calibrating a detector and measuring cross sections for neutron induced reactions. PHITS can deal with the transport of incident protons as well as secondary neutrons using various physics models, and it can estimate particle fluxes in the beam line and energy deposition in shielding materials. Therefore, PHITS is a useful code for neutron source design in accelerator facilities. However, nuclear reaction models implemented in PHITS, such as INC, were not enough to reproduce the peak structure in neutron spectra of experimental data, since these models do not consider the transition process of $$^7$$Li($$p$$,$$n$$)$$^7$$Be. We have developed a new option that adds peaks obtained by the DWBA method, which gives cross sections of the transition on the basis of quantum mechanics, to results calculated by the INC model. We had applied this option to estimate neutron spectra in the reactions at incident energies below 50 MeV. Results of the INC model using the option had been in good agreement with experimental data. In this study, we extended the applicable incident energy range up to 400 MeV for the $$^7$$Li($$p$$,$$xn$$) reactions. We will show the comparison between the calculated result and experimental data, and discuss the validity of the option for the reactions.

Journal Articles

Sensitivity and uncertainty analysis of the VENUS-F critical core

Iwamoto, Hiroki; Stankovskiy, A.*; Uyttenhove, W.*

Energy Procedia, 71, p.33 - 41, 2015/05

 Times Cited Count:1 Percentile:75.48

Journal Articles

Validation of burnup calculation code SWAT4 by evaluation of isotopic composition data of mixed oxide fuel irradiated in pressurized water reactor

Kashima, Takao; Suyama, Kenya; Mochizuki, Hiroki*

Energy Procedia, 71, p.159 - 167, 2015/05

 Times Cited Count:1 Percentile:75.48

The nuclear fuel cycle program of Japan would be delayed because of the impact of the Fukushima Daiichi NPP accident in 2011. Excessive plutonium, however, has to be utilized as mixed-oxide (MOX) fuel to reduce the quantity of plutonium possessed by Japan. Calculation codes and libraries adopted in the fuel cycle analyses of MOX fuel should be benchmarked based on comparison between calculation results and experimental data. From another viewpoint, nuclide inventory analyses of MOX fuel is important for evaluations of the Fukushima accident because MOX fuel has been loaded in the Unit 3 reactor. ARIANE is a PIE program which includes measurements of nuclide compositions of spent MOX fuels discharged from both of pressurized and boiling water reactors. In this study, the PIE data of MOX fuels irradiated in a pressurized water reactor were analyzed by the integrated burnup code system SWAT4 that combines the point burnup system ORIGEN2 and neutron transport calculation solvers, the continuous energy Monte Carlo code MVP or MCNP, and the deterministic neutronics calculation code SRAC. The calculation results of SWAT4 have generally same trends with the case of UO$$_{2}$$ fuel analyses. For major uranium and plutonium isotopes, deviations less than 5% were obtained. This means that SWAT4 has the same accuracy to predict isotopic compositions of irradiated MOX fuel with the case of UO$$_{2}$$ fuel. The radial distribution of isotopes in a pellet was also analyzed, whose results were compared with that measured by SIMS. SWAT4 predicted well the isotope and burnup distributions in an irradiated MOX pellet.

Journal Articles

Problem on MATXS files with multiple temperature cross section data

Konno, Chikara; Maeda, Shigetaka; Kosako, Kazuaki*

Energy Procedia, 71, p.213 - 218, 2015/05

 Times Cited Count:0 Percentile:0.11

We report a problem on multigroup cross section data MATXS files with multiple temperatures. This problem was newly found out through neutron and $$gamma$$ flux calculations in a simple model of experimental fast reactor Joyo with DORT and MATXSLIB-J40, which is a multigroup cross section data file (300, 600, 900, 1200, 1800 K) of the latest Japanese Nuclear Data Library version 4.0 (JENDL-4.0) processed with the NJOY99 code. The calculated total neutron fluxes were almost the same both in 300 K and 600 K, while the total $$gamma$$ fluxes in 600 K were by 10% higher those that in 300 K. Through our detailed investigation, it was found out that the MATXS data format processed with NJOY was not consistent to that assumed in TRANSX for $$gamma$$ production data. In order to solve this problem, we made a simple program for modifying MATXS files to ones suitable to TRANSX. MATXSLIB-J40 will be revised with this program.

Journal Articles

Development of a calculation code system for evaluation of deuteron nuclear data

Nakayama, Shinsuke*; Araki, Shohei*; Watanabe, Yukinobu*; Iwamoto, Osamu; Ye, T.*; Ogata, Kazuyuki*

Energy Procedia, 71, p.219 - 227, 2015/05

 Times Cited Count:9 Percentile:98.51

A calculation code system for evaluation of deuteron nuclear data is extended so that the stripping reaction to bound states in the residual nucleus can be taken into account properly using a conventional zero-range DWBA approach. The code system is applied to deuteron induced-reactions on $$^{27}$$Al for incident energies up to 100 MeV. It is found that the spectroscopic factors derived from the present DWBA analysis have incident energy dependence. The calculation using the extended code system reproduced experimental double-differential (d,xp) cross sections at 25.5, 56, and 100 MeV, and production cross sections of $$^{28}$$Al in the incident energy range from the threshold to 20 MeV.

Journal Articles

Research program for the evaluation of fission product and actinide release behaviour, focusing on their chemical forms

Miwa, Shuhei; Yamashita, Shinichiro; Ishimi, Akihiro; Osaka, Masahiko; Amaya, Masaki; Tanaka, Kosuke; Nagase, Fumihisa

Energy Procedia, 71, p.168 - 181, 2015/05


 Times Cited Count:15 Percentile:99.79

A basic study towards enhanced safety management of irradiated fuels and materials from a severe accident is underway utilizing JAEA's hot laboratory complex in Oarai. The present study that consists of three basic research programs is aimed at contributing to building enhanced safety management measures (including radioactive decontamination, evaluation measurements, safekeeping, treatment and disposal) of irradiated fuels and materials from the severe accident. In this paper, not only the overview of activities of individual research programs but also the several preliminary results were shown together with future plans.

Journal Articles

Oxidation behavior of Am-containing MOX fuel pellets in air

Tanaka, Kosuke; Yoshimochi, Hiroshi; Obayashi, Hiroshi; Koyama, Shinichi

Energy Procedia, 71, p.282 - 292, 2015/05

 Times Cited Count:2 Percentile:88.64

Americium-containing MOX (Am-MOX) fuels were subjected to heating tests using thermogravimetric and differential thermal analysis (TG-DTA) measurements in a flowing gas atmosphere of dry air to investigate the effect of Am addition on oxidation behavior of MOX fuel.

Journal Articles

Theoretical study of beta decay for delayed neutron

Koura, Hiroyuki; Chiba, Satoshi*

Energy Procedia, 71, p.228 - 236, 2015/05

 Times Cited Count:0 Percentile:0.11

Theoretical calculation of $$beta$$ decay for delayed neutron is studied. $$beta$$ decay is one of the nuclear decay processes, but only occurs with the weak interaction. The number of averaged delayed neutron emission is estimated from the idea of the summation method, in which the averaged delayed neutron emission is sum of the $$beta$$-delayed neutron emission probabilities multiplied by cumulative fission yield. Under the consideration of the gross they of the $$beta$$ decay, current status for reproduction of $$beta$$-decay rate and delayed neutron probability is studied.

Oral presentation

Development of photon-induced pion production in the PHITS Code

Noda, Shusaku; Hashimoto, Shintaro; Sato, Tatsuhiko; Niita, Koji*

no journal, , 

It is of great importance to estimate the dose of secondarily generated protons and neutrons by photonuclear reaction. The Particle and Heavy Ion Transport code System (PHITS) used to be able to treat photonuclear reactions only up to 150 MeV because giant-dipole resonance and quasideuteron reaction were considered. When a high energy photon above 150 MeV induces photonuclear reaction with nucleus, nucleon resonance is produced followed by pion production. In this research, pion-production total cross section for every nucleus is determined using the systematics considering nuclear medium effect and shadowing effect of each nucleus. The process of pion production such as resonance and its decay is simulated, and subsequently emitted pion(s) and proton/neutron are transported in the JQMD model of the PHITS code. As a benchmark test the PHITS calculation results are compared with the experimental data of photon-incident proton/neutron-emission cross sections.

Oral presentation

Test-fabrication of hydrogen iodide decomposer for IS process

Iwatsuki, Jin; Noguchi, Hiroki; Kubo, Shinji; Kasahara, Seiji; Tanaka, Nobuyuki; Takegami, Hiroaki; Onuki, Kaoru; Inagaki, Yoshiyuki

no journal, , 

The Japan Atomic Energy Agency has been conducting R&D on IS process for the future hydrogen energy system. The basic technique for the closed-cycle hydrogen production has been developed and a membrane process is under study for the efficient hydrogen production. Presently, main efforts are devoted to examine the integrity of components made of industrial materials screened by corrosion tests in the IS process environments. So far, fabrication of HI decomposer as key component of the HI decomposition section was completed. Test-fabricated HI decomposer has a performance to produce hydrogen for 150 NL/h, and maximum pressure is 0.95 MPa, maximum temperature is 500$$^{circ}$$C. The components are consisted of triple structure, pressure vessel, outer tube, and inner tube. In order to secure corrosion-resistance, alloy C-276 as corrosion-resistant material was used for component.

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