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Journal Articles

Non-proliferation features in partitioning and transmutation cycle using accelerator-driven system; Evaluation of ${it Material Attractiveness}$ of fuel assembly in early period of burnup cycle

Oizumi, Akito; Sugawara, Takanori; Sagara, Hiroshi*

Dai-41-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2020/11

Research and development of partitioning and transmutation cycle with accelerator drive systems (ADSs) transmuting minor actinides (MAs) separated from the commercial cycles has been continuously conducted to reduce the high-level radioactive waste (HLW) contained in spent fuel discharged from nuclear power plants. Since the chemical form and composition of the fuels are different from those of the current commercial cycles, it is necessary to examine the accuracy of the safeguards (SGs) and the level of physical protections (PPs) which are required for the partitioning and transmutation cycle. In this study, ${it Material Attractiveness}$ of the first cycle fuel assemblies (fresh and spent fuels) in the fuel storage pool in the ADS facility was evaluated and it was compared with that of the MOX fuel assemblies (fresh and spent fuels) for a general boiling water reactor (BWR). As a result, it made clear that the fuel storage pool in the ADS facility storing the first cycle fuel assemblies were required the SG detection accuracy and PP level equal to or lower than the MOX fuel assembly of the BWR since the ADS fuel assembly in the first cycle was less attractive than the MOX fuel assembly for the BWR.

Journal Articles

Development of a user-friendly interface IRONS for atmospheric dispersion database for nuclear emergency preparedness based on the Fukushima database

El-Asaad, H.*; Nagai, Haruyasu; Sagara, Hiroshi*; Han, C. Y.*

Annals of Nuclear Energy, 141, p.107292_1 - 107292_9, 2020/06

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Atmospheric dispersion simulations can provide crucial information to assess radioactive plumes in the environment for nuclear emergency preparedness. However, it is a difficult and time-consuming task to make simulations assuming many possible scenarios and to derive data from a vast number of results. Therefore, an interface was developed to assist users in conveying characteristics of plumes from simulation results. The interface uses a large database that contains WSPEEDI-II simulations for the first 20-days of radioactive release from the Fukushima Daiichi Nuclear Power Plant, and it displays essential quantitative data to the user from the database. The user may conduct sensitivity analysis with the help of the interface by changing release condition to generate many different case scenarios.

Journal Articles

Proliferation resistance evaluation of an HTGR transuranic fuel cycle using PRAETOR code

Aoki, Takeshi; Chirayath, S. S.*; Sagara, Hiroshi*

Annals of Nuclear Energy, 141, p.107325_1 - 107325_7, 2020/06

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The proliferation resistance (PR) of an inert matrix fuel (IMF) in the transuranic nuclear fuel cycle (NFC) of a high temperature gas cooled reactor is evaluated relative to the uranium and plutonium mixed-oxide (MOX) NFC of a light water reactor using PRAETOR code and sixty-eight input attributes. The objective is to determine the impacts of chemical stability of IMF and fuel irradiation on the PR. Specific material properties of the IMF, such as lower plutonium content, carbide ceramics coating, and absence of $$^{235}$$U, contribute to enhance its relative PR compared to MOX fuel. The overall PR value of the fresh IMF (an unirradiated direct use material with a one-month diversion detection timeliness goal) is nearly equal to that of the spent MOX fuel (an irradiated direct use nuclear material with a three-month diversion detection timeliness goal). Final results suggest a reduced safeguards inspection frequency to manage the IMF.

Journal Articles

User interface of atmospheric dispersion simulations for nuclear emergency countermeasures

Hamuza, E.-A.; Nagai, Haruyasu; Sagara, Hiroshi*

Energy Procedia, 131, p.279 - 284, 2017/12

 Times Cited Count:1 Percentile:27.23

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.

Journal Articles

A Simple method to create gamma-ray-source spectrum for passive gamma technique

Shiba, Tomooki; Maeda, Shigetaka; Sagara, Hiroshi*; Ishimi, Akihiro; Tomikawa, Hirofumi

Energy Procedia, 131, p.250 - 257, 2017/12

 Times Cited Count:0 Percentile:100

In the present paper, the $$gamma$$ 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 $$gamma$$ 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 $$gamma$$ 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 $$gamma$$ 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 $$gamma$$ ray spectrum shape agreed well with the shape of $$gamma$$ ray spectrum obtained by the experiment.

Journal Articles

Characterization study of four candidate technologies for nuclear material quantification in fuel debris at Fukushima Daiichi Nuclear Power Station

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

 Times Cited Count:2 Percentile:11.59

Journal Articles

Characterization study of four candidate technologies for nuclear material quantification in fuel debris at Fukushima Daiichi Nuclear Power Station, 3; Numerical simulation of passive gamma technique

Shiba, Tomooki; Sagara, Hiroshi*; Tomikawa, Hirofumi

Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 3 Pages, 2017/09

Since the removal of fuel debris from the Fukushima Daiichi Nuclear Power Plant is planned to commence in 2021, measurement technologies for quantification of the nuclear material in fuel debris will be required for appropriate nuclear material management. In this paper, an outline of a passive gamma technique as one of the measurement technologies is briefly described, and the results of phase 1 and 2 of the so-called common set of simulation models for fuel debris and canisters are reported. The newly developed coupling method is applied to produce a gamma ray source for simulation. As the result of phase 1, it is revealed that the variation in the composition of fuel debris does not affect the gamma ray leakage behavior from canisters. According to the result of phase 2, the primary peak of Eu-154 at 1.27 MeV is clearly observable, although the debris is centrally located in canister. In addition, rotational scanning is effective for correcting the deviation in detection efficiency due to debris located off-center in canisters.

Journal Articles

Applicability evaluation of candidate technologies for nuclear material quantification in fuel debris at Fukushima Daiichi Nuclear Power Station; Passive gamma technique

Shiba, Tomooki; Tomikawa, Hirofumi; Sagara, Hiroshi*; Ishimi, Akihiro

57th Annual Meeting of the Institute of Nuclear Materials Management (INMM 2016), Vol.1, p.365 - 374, 2017/02

Fission products (FPs) such as Ce and Eu seem to be very low volatile even in high temperature environment in the severe accident of nuclear reactors. They are supposed to chemically coexist with nuclear fuel. In our passive gamma ray spectroscopy, gamma rays from such FPs are measured and their amount and burnup are estimated. Then we multiply the mass ratio of the FPs and nuclear materials, and we obtain the mass of nuclear material of interest. According to hypothetical fuel debris and canister models, we performed simulation of leakage gamma rays from canister of debris. gamma ray source spectra were derived from the composition of the hypothetical fuel debris and the photon transport calculation was carried out. In addition, we conducted experiments of gamma ray measurement from intact spent fuels irradiated in the Experimental Fast Reactor Joyo in order to validate the prediction performance of the simulation by the comparison of the experiments and simulation.

Journal Articles

Characterization study of four candidate technologies for nuclear material quantification in fuel debris at Fukushima Daiichi Nuclear Power Station (Interim report)

Nagatani, Taketeru; Komeda, Masao; Shiba, Tomooki; Maeda, Makoto; Nauchi, Yasushi*; Sagara, Hiroshi*; Kosuge, Yoshihiro*; Kureta, Masatoshi; Tomikawa, Hirofumi; Okumura, Keisuke; et al.

Proceedings of INMM 57th Annual Meeting (Internet), 10 Pages, 2016/07

Journal Articles

Passive gamma spectrometry of low-volatile FPs for accountancy of special nuclear material in molten core material of Fukushima Daiichi Nuclear Power Plant; Evaluation of detector response from various hypothetical fuel canister

Shiba, Tomooki; Sagara, Hiroshi*; Tomikawa, Hirofumi

56th Annual Meeting of the Institute of Nuclear Materials Management (INMM 2015), Vol.3, p.1735 - 1741, 2016/00

In response to the accident at Fukushima Daiichi Nuclear Power Station, passive gamma spectrometry is being researched and developed as one of the candidates of a mass measurement method for the special nuclear materials in molten core material. Among Fission Products (FPs) accompanied in molten core materials, some of them are very low-volatile and emit high-energy gamma rays, which enable us to derive the mass of those FPs by passive gamma spectrometry. Using the mass ratio of the FPs and nuclear materials, this technique provides the mass estimation of nuclear materials. This technique is relatively simple and was applied to the analysis of nuclear materials in the clean-up process of damaged Three Mile Island unit-2 (TMI-2) reactor. In this paper, we show the characteristics of leakage gamma rays from removed fuel canisters simulated by MCNP. Different detector responses to gamma rays from fuel debris are also evaluated supposing NaI, LaBr$$_{3}$$ and HPGe.

JAEA Reports

PR&PP evaluation; ESFR full system case study final report (Tentative translation)

Sagara, Hiroshi; Kawakubo, Yoko; Inoue, Naoko

JAEA-Review 2013-011, 54 Pages, 2014/01

JAEA-Review-2013-011.pdf:3.05MB

The Generation IV (GEN IV) International Forum Proliferation Resistance and Physical Protection (PR & PP) Working Group is in charge of developing a methodology for evaluating PR & PP of potential GEN IV options. The present report, published in Oct. 2009, was used as a supporting study for development of the evaluation methodology for PR & PP, summarizing the case study of the PR & PP evaluation of Example Sodium Fast Reactor (ESFR) co located with a dry fuel storage facility and a pyrochemical spent-fuel reprocessing facility, a hypothetical nuclear energy system, consisting of nine main system elements, and it provides for designers the practical experience of applying the PR&PP evaluation methodology to a nuclear energy system. The development of the future nuclear fuel cycle system with sufficient PR & PP features is a crucial task in Japan. With the usefulness the report, it was translated and published here as a Japanese-language edition with the concurrence of the OECD-NEA.

Journal Articles

Feasibility study of passive $$gamma$$ spectrometry of molten core material from Fukushima Daiichi Nuclear Power Station unit 1, 2 and 3 cores for special nuclear material accountancy; Low-volatile FP and special nuclear material inventory analysis and fundamental characteristics of $$gamma$$-rays from fuel debris

Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

Journal of Nuclear Science and Technology, 51(1), p.1 - 23, 2014/01

 Times Cited Count:8 Percentile:36.16(Nuclear Science & Technology)

Feasibility study of $$gamma$$ spectrometry of molten core material from Fukushima Daiichi Nuclear Power Station Unit 1, 2 and 3 cores for special nuclear material accountancy has been performed, focusing on the low-volatile fission product and heavy metal inventory analysis, and fundamental characteristics of $$gamma$$-ray from fuel debris for passive measurement. The inventory ratio of low-volatile lanthanides, $$^{154}$$Eu and $$^{144}$$Ce, to special nuclear material were evaluated by whole core inventory in unit 1, 2 and 3 cores as reference value for homogenized molten fuel material, and also as a function of burnup for specific fuel debris, considering the sensitivity of enrichment, specific power, water void fraction, cooling time, calculation tool accuracy and release ratio. The same indices could be applied to unit 3, while the uncertainty of specific fuel debris of separated MOX fuel would be increased significantly. Source photon spectrum results showed the detectability of low-volatile high energy $$gamma$$-rays emitted from $$^{154}$$Eu at least in 20 years after the accident, that from Ce/$$^{144}$$Pr in 10 years, and volatile $$^{134}$$Cs and $$^{137}$$Cs at least in 20 years with 99% or more release ratio. Mass attenuation coefficients of fuel debris was evaluated to be insensitive to its compositions in high energy region. Leakage photon ratio was evaluated with variety of parameters and significant impact was confirmed with certain size of fuel debris, its correlation was summarized with respect to the photopeak ratio of source $$^{154}$$Eu. Finally, preliminary study with hypothetical canister model of fuel debris based on TMI-2 experience, and future plan were introduced.

Journal Articles

Optimization study on accelerator driven system design for effective transmutation of iodine-129

Ismailov, K.*; Nishihara, Kenji; Saito, Masaki*; Sagara, Hiroshi*

Annals of Nuclear Energy, 56, p.136 - 142, 2013/06

 Times Cited Count:5 Percentile:54.36(Nuclear Science & Technology)

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.

Journal Articles

Effect of radial zoning of $$^{241}$$Am content on homogenization of denatured Pu with broad range of neutron energy based on U irradiation test in the experimental fast reactor Joyo

Shiba, Tomooki*; Sagara, Hiroshi*; Onishi, Takashi; Koyama, Shinichi; Maeda, Shigetaka; Han, C. Y.*; Saito, Masaki*

Annals of Nuclear Energy, 51, p.74 - 80, 2013/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The design consideration of DU-Am oxide fuel pin was performed for Pu denaturing in the framework of the protected plutonium production based on the irradiation analyses of the depleted U (DU) samples irradiated in the environment of broad range of neutron energy in the experimental fast reactor Joyo. From the results of irradiation analyses of DU, it was confirmed that there is a strong dependence of transmutation behavior of DU on the resonance neutron ratio even in a fast reactor. Also, it was confirmed that there is a strong effect of sample material form and shape on generated Pu nuclide inventory especially near the reflector area ($$>$$20% resonance neutron ratio), because of the intensive self-shielding of $$^{238}$$U, though less is expected for $$^{241}$$Am. Sensitivity study of hypothetical DU-Am oxide fuel pellet irradiation on neutron energy and burn-up was performed to evaluate significant gradient of radial $$^{238}$$Pu isotopic composition profile (e.g., from 12 to 18% distribution in 3% Am doping, in 30% resonance neutron ratio and in 4.0$$times$$10$$^{22}$$ [n/cm$$^{2}$$] of neutron fluence inside a large pellet with softened neutron spectrum), and vulnerability of the fuel pellet surface in terms of Pu denaturing was revealed. Design consideration of radial zoning of $$^{241}$$Am content was introduced to flatten the radial distribution of isotopic composition of Pu. The results of radial zoning of $$^{241}$$Am (4% and 3% of Am in the outer and inner zone of DU-Am oxide fuel pellet) in hypothetical irradiation neutronics analysis showed the radial profile of produced Pu is over 15 at.% of $$^{238}$$Pu isotopic composition in any zone and meets the criteria of Kimura et al. based on decay heat of Pu to impede utilization to fission explosive devices.

Journal Articles

Sensitivity analysis of low-volatile FPs and Cm-244 inventory in irradiated nuclear fuel for special nuclear material accountancy in fuel debris

Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

Transactions of the American Nuclear Society, 107(1), p.803 - 804, 2012/11

Fission products(FPs) such as Eu, Ce, Ru have low release ratio in case of core melting event in severe accident to co-exist inside fuel debris in oxide or metallic phase, based on TMI-2 experience and source term experiments. Passive $$gamma$$ spectroscopy of Ce-144/Pr-144 was utilized in quantifying special nuclear material in fuel debris of TMI-2 historically, and the same methodology might be applied to high energy $$gamma$$ emitter, Eu-154 and Ru-106. Passive neutron measurement has been also utilized for quantifying nuclear material practically. Different from conventional spent fuel, however, fuel debris would be lost in information of irradiated profile, release ratio of volatile FPs, and so on. In the present paper, sensitivity analysis of low-volatile FPs and Cm-244 inventory in irradiated nuclear fuel in light water reactor was performed numerically to clarify the uncertainty of low-volatile FP and Cm inventory regarding fuel irradiation parameters and calculation methodology, and to derive applicable index to quantify nuclear material in fuel debris.

Journal Articles

Effect of neutron moderator on protected plutonium production in fast breeder reactor blanket

Matsumoto, Koji*; Sagara, Hiroshi*; Han, C. Y.*; Onishi, Takashi; Saito, Masaki*; Yamauchi, Ippei*

Transactions of the American Nuclear Society, 107(1), p.1018 - 1019, 2012/11

Protected plutonium production (P$$^{3}$$) with a high proliferation-resistance was proposed by increasing the $$^{238}$$Pu ratio in the total plutonium through minor actinides (MAs) doping into the fresh blanket fuel. Moderator effect on P$$^{3}$$ was evaluated. As a result, the transmutation ratio is larger with the heterogeneous moderator than with the homogeneous one, and the isotopic ratio of $$^{238}$$Pu was increased.

Journal Articles

Feasibility study of $$gamma$$ spectroscopy of low-volatile FPs for special nuclear material accountancy in molten core material

Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

Proceedings of INMM 53rd Annual Meeting (CD-ROM), 10 Pages, 2012/07

Reviewing the technologies applied to TMI-2, feasibility study of $$gamma$$ spectroscopy of low-volatile FPs for special nuclear material accountancy in molten core material or debris has been performed numerically, and the sensitivity of low-volatile FP nuclides on fuel contents, especially $$^{239}$$Pu, was studied with parameters of typical BWR fuel such as enrichment, burnup and neutron spectrum, heat density. Comparing TMI-2, a PWR reactor, broader neutron spectrum axial profile and irradiation cycle complexity of typical BWR fuel assemblies make more variance in the accumulation of daughter nuclides by neutron capture reactions. As results, $$^{239}$$Pu quantification, by burnup dependent $$^{154}$$Eu/$$^{239}$$Pu index has accuracy of 15-18% in 1-sigma level mainly affected by burnup uncertainty, and $$^{239}$$Pu quantification by $$^{144}$$Ce/$$^{239}$$Pu has accuracy of 20% as long as $$^{144}$$Ce released photon could be observed within 10 years, within the scope of inventory survey except for measurement uncertainty. Finally systematic image of fuel quantification by passive $$gamma$$ spectroscopy and, as future study, FPs quantification by passive $$gamma$$ measurement tests by mockup debris measurement with self-attenuation correction are introduced.

Journal Articles

Protected plutonium production at fast breeder reactor blanket; Chemical analysis of uranium-238 samples irradiated in experimental fast reactor Joyo

Onishi, Takashi; Koyama, Shinichi; Shiba, Tomooki*; Sagara, Hiroshi*; Saito, Masaki*

Progress in Nuclear Energy, 57, p.125 - 129, 2012/05

 Times Cited Count:1 Percentile:87.88(Nuclear Science & Technology)

In order to develop blanket fuel with high proliferation resistance in fast breeder reactor, chemical analysis of nine $$^{238}$$U samples irradiated in experimental fast reactor Joyo and Pu contents and Pu isotopic composition of the samples were measured. As results, dependence of Pu production behavior from $$^{238}$$U on neutron spectra was revealed.

Journal Articles

Report and analysis on "PR&PP Evaluation; Example Sodium Fast Reactor Full System Case Study"

Sagara, Hiroshi; Inoue, Naoko; Kawakubo, Yoko; Watahiki, Masaru

Kaku Busshitsu Kanri Gakkai (INMM) Nippon Shibu Dai-32-Kai Nenji Taikai Rombunshu (Internet), 9 Pages, 2011/11

The Generation IV (GEN IV) Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PRPP WG) was established in December 2002 in order to develop the PR&PP evaluation methodology for GEN IV nuclear energy systems. In the final report of "PR&PP Evaluation; Example Sodium Fast Reactor (ESFR) Full System Case Study," issued in October 2009, the demonstration study of PR&PP evaluation with the qualitative approach are summarized using ESFR with four scenario threats. The present paper reviews and analyzes some results of the ESFR case study, and identifies the challenges and direction for the PR&PP evaluation methodology with quantitative approach.

Journal Articles

Design of $$gamma$$-ray and neutron area monitoring system for the IFMIF/EVEDA accelerator building

Takahashi, Hiroki; Maebara, Sunao; Kojima, Toshiyuki; Kubo, Takashi; Sakaki, Hironao; Takeuchi, Hiroshi; Shidara, Hiroyuki; Hirabayashi, Keiichi*; Hidaka, Kosuke*; Shigyo, Nobuhiro*; et al.

Fusion Engineering and Design, 86(9-11), p.2795 - 2798, 2011/10

 Times Cited Count:1 Percentile:87.98(Nuclear Science & Technology)

In the IFMIF/EVEDA accelerator, the engineering validation up to 9 MeV by employing the deuteron beam of 125 mA are planning at the BA site in Rokkasho, Aomori, Japan, the personnel protection system (PPS) is indispensable. The PPS inhibit the beam by receiving the interlock signal from the $$gamma$$-ray and neutron monitoring system. The $$gamma$$-ray and neutron detection level which is planned to be adopted are "80 keV to 1.5 MeV ($$gamma$$-ray)" and "0.025 eV to 15 MeV (neutron)". For the present shielding design, it is absolutely imperative for the safety review to validate the shielding ability which makes detection level lower than these $$gamma$$-ray and neutron detector. For this purpose, the energy reduction of neutron and photon for water and concrete is evaluated by PHITS code. From the calculating results, it is found that the photon energy range extended to 10 MeV by water and concrete shielding material only, an additional shielding to decrease the photon energy of less than 1.5 MeV is indispensable.

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