Kadowaki, Masanao; Nagai, Haruyasu; Terada, Hiroaki; Katata, Genki*; Akari, Shusaku*
Energy Procedia, 131, p.208 - 215, 2017/12
When radioactive materials are released into the atmosphere due to nuclear accidents, numerical simulations that can reproduce temporal and spatial distribution of radioactive materials are useful to provide the information for emergency responses and radiological dose assessment. In this study, we attempt to improve the atmospheric dispersion simulation using an advanced meteorological data assimilation method and reconstruct the spatiotemporal distribution of radioactive materials released due to the Fukushima Daiichi Nuclear Power Station (FDNPS) accident. The atmospheric dispersion simulations were carried out by the Lagrangian particle dispersion model GEARN developed by Japan Atomic Energy Agency. To obtain meteorological fields for GEARN calculation, we used the Weather Research and Forecasting model WRF with meteorological data assimilation using four-dimensional variational method (4D-Var). GEARN calculations of the surface deposition and air concentration of radionuclides were compared with measurements. In the area close to FDNPS, the spatial distribution of the deposition of Cs-137 and I-131 simulated by GEARN agreed with the measured one. The accuracy of modeled deposition in northwest and south directions from FDNPS was particularly improved. This results were mainly attributed to the better reproducibility of wind field by using the meteorological data assimilation with 4D-Var. The improvement of the accuracy of modeled deposition distribution of Cs-137 in the East Japan area was also apparent under the meteorological fields modified by 4D-Var. The information of atmospheric dispersion processes reconstructed in this study is used for updating the existing assessment of radiological dose resulting from the FDNPS accident based on atmospheric simulations by our previous studies. It can also provide useful suggestions to make emergency response plans for nuclear facilities in Japan.
Hamuza, E.-A.; Nagai, Haruyasu; Sagara, Hiroshi*
Energy Procedia, 131, p.279 - 284, 2017/12
In this study we would like to propose a method to use atmospheric dispersion simulations by WSPEEDI for consideration of crisis management on radionuclide dispersion from a nuclear power plant. WSPEEDI can simulate and output crucial information regarding environmental distribution of radionuclides and weather pattern for nuclear emergency countermeasures, thus this study will make use of its output to display the effective information for evacuation planning from a radionuclide dispersion. We will be assembling database of atmospheric dispersion outputs for one year by using WSPEEDI for a nuclear facility, then the database will be analysed to make the summary that has useful information for nuclear emergency managements. WSPEEDI outputs are converted into numeric information showing dispersion characteristics so that users can understand WSPEEDI predictions easily.
Sekine, Megumi; Tomikawa, Hirofumi
Energy Procedia, 131, p.274 - 278, 2017/12
The IAEA has proposed in its long-term R&D plan, development of real-time measurement technology for monitoring and verifying nuclear material movement continuously. At a PULEX reprocessing facility, HNO solution with dissolved spent fuel, such as FPs and nuclear materials, flows in pipes and stores in tanks. In order to detect and deter nuclear material being stolen from the process, measuring the 10.8 MeV rays emitted by N(n,)N reaction activated by spontaneous neutrons might be useful for continuous monitoring of liquid flow. In general, since high dose rays emitted from FPs are dominant below 3 MeV, it is expected that the 10.8 MeV peak would not be affected by the FP peaks. As the first step, some kinds of detectors and measurement configuration were analyzed through MCNP based on 10.8 MeV rays activated by neutrons from a Cf source.
Shiba, Tomooki; Maeda, Shigetaka; Sagara, Hiroshi*; Ishimi, Akihiro; Tomikawa, Hirofumi
Energy Procedia, 131, p.250 - 257, 2017/12
In the present paper, the ray source data was developed for the debris composition based on "best estimates", and the subsequent photon transportation calculation was performed to evaluate the leakage ray spectra according to the fuel debris. Since the creation of the line spectrum source requires a great deal, we have developed the relatively simple but accurate enough method to build up ray source, coupling of baseline spectra evaluated by ORIGEN2 code and several line spectra of interest. One of the advantages of the method is taking bremsstrahlung X rays into consideration by utilizing the bremsstrahlung libraries of ORIGEN2. The new ray source was used to calculate the detector response of HPGe detector and the results was compared as a benchmark with experimental measurement results of irradiated fuel pins. As the result, the simulated ray spectrum shape agreed well with the shape of ray spectrum obtained by the experiment.
Onishi, Takashi; Sekioka, Ken*; Suto, Mitsuo*; Tanaka, Kosuke; Koyama, Shinichi; Inaba, Yusuke*; Takahashi, Hideharu*; Harigai, Miki*; Takeshita, Kenji*
Energy Procedia, 131, p.151 - 156, 2017/12
no abstracts in English
Nagatani, Taketeru; Komeda, Masao; Shiba, Tomooki; Nauchi, Yasushi*; Maeda, Makoto; Sagara, Hiroshi*; Kosuge, Yoshihiro*; Kureta, Masatoshi; Tomikawa, Hirofumi; Okumura, Keisuke; et al.
Energy Procedia, 131, p.258 - 263, 2017/12
Noguchi, Hiroki; Takegami, Hiroaki; Kasahara, Seiji; Tanaka, Nobuyuki; Kamiji, Yu; Iwatsuki, Jin; Aita, Hideki; Kubo, Shinji
Energy Procedia, 131, p.113 - 118, 2017/12
The IS process is the most deeply investigated thermochemical water-splitting hydrogen production cycle. It is in a process engineering stage in JAEA to use industrial materials for components. Important engineering tasks are verification of integrity of the total process and stability of hydrogen production in harsh environment. A test facility using corrosion-resistant materials was constructed. The hydrogen production ability was 100 L/h. Operation tests of each section were conducted to confirm basic functions of reactors and separators, etc. Then, a trial operation for integration of the sections was successfully conducted to produce hydrogen of about 10 L/h for 8 hours.
Kondo, Masatoshi*; Okubo, Nariaki; Irisawa, Eriko; Komatsu, Atsushi; Ishikawa, Norito; Tanaka, Teruya*
Energy Procedia, 131, p.386 - 394, 2017/12
The chemical behaviors of lead (Pb) based coolants in the air ingress accident of fast reactors were investigated by means of the thermodynamic considerations and the static oxidation experiments for Pb alloys at various chemical compositions. The results of the static oxidation tests for lead-bismuth (Pb-Bi) alloys indicated that Pb was depleted from the alloy due to the preferential formation of PbO in air at 773K. Pb-Bi oxide and BiO were formed after the enrichment of Bi in the alloys due to the Pb depletion. The oxidation rates of the alloys were much larger than that of the steels, and became larger with higher Pb concentration in the alloys. The compatibility of Pb-Bi alloys with stainless steel was worse when the Pb concentration in the alloys became low, since the dissolution type corrosion was promoted by the Bi composition in the alloy. The Pb-Li alloys were oxidized as they formed LiPbO and LiCO. Then, Li was depleted from the alloy.
Adhi, P. M.*; Okubo, Nariaki; Komatsu, Atsushi; Kondo, Masatoshi*; Takahashi, Minoru*
Energy Procedia, 131, p.420 - 427, 2017/12
The ionic conductivity of solid electrolyte may insufficient, and the sensor output signal will deviate from the theoretical one in low temperature. The performance of oxygen sensor with Ag/air reference electrode (RE) and liquid Bi/BiO RE was tested in low-temperature LBE at 300450C and the charge transfer reactions impedance at the electrode-electrolyte interface was analyzed by electrochemical impedance analysis (EIS). After steady state condition, both of the sensors performed well and can be used at 300450C. Bi/Bi/BiO RE has lower impedance than Ag/air RE. Therefore, the response time of the oxygen sensor with Bi/Bi/BiO RE is faster than the oxygen sensor with Ag/air RE in the low-temperature region.
Di Lemma, F. G.; Yamashita, Shinichiro; Miwa, Shuhei; Nakajima, Kunihisa; Osaka, Masahiko
Energy Procedia, 127, p.29 - 34, 2017/09
Chemical effects of molybdenum (Mo) and boron (B), which were considered to form compounds with Cs, on the Cs chemisorption were predicted using a chemical equilibrium calculation. It is seen that CsMoO were formed in the chemisorbed compounds. On the other hand, little effects were observed for B. The results suggest that the effects of Mo should be considered for further experimental investigation.
Sato, Isamu; Tanaka, Kosuke; Koyama, Shinichi; Matsushima, Kenichi*; Matsunaga, Junji*; Hirai, Mutsumi*; Endo, Hiroshi*; Haga, Kazuo*
Energy Procedia, 82, p.86 - 91, 2015/07
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, CsMoO, 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.
Hashimoto, Shintaro; Iwamoto, Osamu; Iwamoto, Yosuke; Sato, Tatsuhiko; Niita, Koji*
Energy Procedia, 71, p.191 - 196, 2015/05
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 Li(,)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 Li(,)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 Li(,) reactions. We will show the comparison between the calculated result and experimental data, and discuss the validity of the option for the reactions.
Iwamoto, Hiroki; Stankovskiy, A.*; Uyttenhove, W.*
Energy Procedia, 71, p.33 - 41, 2015/05
Kashima, Takao; Suyama, Kenya; Mochizuki, Hiroki*
Energy Procedia, 71, p.159 - 167, 2015/05
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 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 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.
Konno, Chikara; Maeda, Shigetaka; Kosako, Kazuaki*
Energy Procedia, 71, p.213 - 218, 2015/05
We report a problem on multigroup cross section data MATXS files with multiple temperatures. This problem was newly found out through neutron and 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 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 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.
Nakayama, Shinsuke*; Araki, Shohei*; Watanabe, Yukinobu*; Iwamoto, Osamu; Ye, T.*; Ogata, Kazuyuki*
Energy Procedia, 71, p.219 - 227, 2015/05
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 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 Al in the incident energy range from the threshold to 20 MeV.
Miwa, Shuhei; Yamashita, Shinichiro; Ishimi, Akihiro; Osaka, Masahiko; Amaya, Masaki; Tanaka, Kosuke; Nagase, Fumihisa
Energy Procedia, 71, p.168 - 181, 2015/05
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.
Tanaka, Kosuke; Yoshimochi, Hiroshi; Obayashi, Hiroshi; Koyama, Shinichi
Energy Procedia, 71, p.282 - 292, 2015/05
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.
Koura, Hiroyuki; Chiba, Satoshi*
Energy Procedia, 71, p.228 - 236, 2015/05
Theoretical calculation of decay for delayed neutron is studied. 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 -delayed neutron emission probabilities multiplied by cumulative fission yield. Under the consideration of the gross they of the decay, current status for reproduction of -decay rate and delayed neutron probability is studied.
Mukaida, Kyoko; Shiotani, Hiroki; Ono, Kiyoshi; Namba, Takashi
Energy Procedia, 39, p.43 - 51, 2013/09
In this study, the energy economic model was upgraded to improve the assessment method of the effect assessment for FR cycle deployment. Moreover, based on the assessment results for FR deployment in the world by the energy economic model, the effects of FR exports were assessed. This results show the FR exports have certain impact to domestic economy.