Izumo, Sari; Hayashi, Hirokazu; Nakata, Hisakazu; Amazawa, Hiroya; Motoyama, Mitsushi*; Sakai, Akihiro
JAEA-Technology 2018-018, 39 Pages, 2019/03
JAEA has planed the near surface disposal of LLW generated from research, industrial, and medical facilities. Maximum radioactivity concentration of each waste and total radioactivity of disposed wastes are needed to be less than the permitted values in the license of disposal facility. Thus, it is important not to evaluate the radioactivity of each waste in unduly conservative ways so as to dispose of the total amount of the waste that is originally planned. Accordingly, the detection limit is required to be as low as the clearance level for the very low level radioactive waste planned to be disposed of trench-type. In this report, the feasibility of the non-destructive assay method is studied by model calculations for gamma emitters. It is confirmed that the detection limit less than the clearance level can be achieved as regards the box type metal container that is difficult to measure. This report summarizes the requirements for the non-destructive measuring equipment.
Hayashi, Hirokazu; Chiba, Rikiya*
Progress in Nuclear Science and Technology (Internet), 5, p.196 - 199, 2018/11
Uranium-free nitride fuel has been chosen as the first candidate for transmutation of long-lived minor actinides (MA: Np, Am, Cm) using sub-critical accelerator-driven system (ADS) under the double strata fuel cycle concept by Japan Atomic Energy Agency (JAEA). Dissolution behavior of ZrN-based nitrides in nitric acid is examined using lanthanides as surrogate materials of TRU elements. Chemical analysis of the ZrN-based lanthanide nitrides dissolved in nitric acid is also carried out.
Hayashi, Hirokazu; Izumo, Sari; Nakata, Hisakazu; Amazawa, Hiroya; Sakai, Akihiro
JAEA-Technology 2018-001, 66 Pages, 2018/06
It is necessary to establish evaluation methodology of radioactivity concentrations of each radionuclide in waste packages for operation of the Near-surface Trench disposal and Sub-surface Pit disposal facility in near future, which has been preparing for low-level radioactive wastes generated from research facilities in JAEA. The radionuclides containing in waste packages generated from both JRR-2 and JRR-3, which are H-3, C-14, Cl-36, Co-60, Ni-63, Sr-90, Nb-94, Tc-99, Ag-108m, I-129, Cs-137, Eu-152, Eu-154, U-234, U-238, Pu-239+240, Pu-238+Am-241, Cm-243+244, were evaluated their density based on radiochemical analysis data, and the Evaluation Methodology of the Radioactivity Concentration such as Scaling Factor method and mean activity concentration method was studied in this report.
Nakata, Hisakazu; Hayashi, Hirokazu; Amazawa, Hiroya; Sakai, Akihiro
JAEA-Technology 2017-031, 41 Pages, 2018/01
JAEA plans to install disposal facilities for radioactive waste arising from research institutes. It must meet the technical standards specified in the relevant rule. One technical standard is that the disposal facilities shall be performance so as not to be left with the voids after the backfilling with soil. Additionally, the rule also requires this radioactive waste be enclosed in a container in which no harmful voids remain. In order to contribute to the development of a method that adapts the disposal facilities to these technical standards, JAEA adopts a waste conditioning artifice that aims for reducing a quantity of voidage in each waste container by a vibration filling method using sandy soil, providing with average void ratios inside the disposal facilities being adequately controlled. In this reports, filling property tests are conducted in the light of filling sand characteristics, types of metal waste and vibration conditions.
Hayashi, Hirokazu; Sato, Takumi; Shibata, Hiroki; Tsubata, Yasuhiro
NEA/NSC/R(2017)3, p.427 - 432, 2017/11
Transmutation of long-lived radioactive nuclides including minor actinides (MA: Np, Am, Cm) has been studied in Japan Atomic Energy Agency (JAEA). Pb-Bi cooled sub-critical accelerator-driven system (ADS) is regarded as one of the powerful tools for transmutation of MA under the double strata fuel cycle concept. Uranium-free MA-Pu nitride fuel was chosen as the first candidate for MA transmutation. Reprocessing of spent ADS fuel and reusing MA recovered from the spent ADS fuels is necessary to improve the transmutation ratio. A pyrochemical process has been proposed as the first candidate for reprocessing of the spent nitride fuel for MA transmutation, because this technique has some advantages over aqueous process, such as the resistance to radiation damage, which is an important issue for the fuels containing large amounts of highly radioactive MA, and feasibility for recovering expensive N-15 in the spent fuels to be reused. This paper overviews the current status of the technology development, including our recent study. Development of the anode suitable for electro-refining of nitride fuels and that of the apparatus for renitridation of the metals recovered in Cd cathode for 100g-Cd scale cold tests are main topics. Evaluation of the batch sizes of each process, which is necessary for estimating the scale of the engineering-apparatus, with considering the decay heat of MA and FP, will also be introduced.
Soejima, Goro; Iwai, Hiroki; Nakamura, Yasuyuki; Hayashi, Hirokazu; Kadowaki, Haruhiko; Mizui, Hiroyuki; Sano, Kazuya
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 5 Pages, 2017/07
no abstracts in English
Hayashi, Hirokazu; Okada, Shota; Izumo, Sari; Hoshino, Yuzuru; Tsuji, Tomoyuki; Nakata, Hisakazu; Sakai, Akihiro; Amazawa, Hiroya; Sakamoto, Yoshiaki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
A near surface disposal for low-level radioactive waste (LLW) generated from commercial nuclear power plants (NPP) is operating in Japan. However, the disposal of LLW from other nuclear facilities and radioisotope utilization facilities has not yet been implemented. Japan Atomic Energy Agency (JAEA) plans to implement the near surface disposal. In order to be disposed of these wastes, it must be confirmed by the regulator that each waste package (radioactive waste solidified with filling materials, such as cement, in a container by a regulated method is termed a waste package) conforms to technical standards that aim for safe disposal. JAEA has studied reasonable confirmation methods to demonstrate the conformity of the waste package to the technical standard as NPP operators have studied it. This report describes the outline of our activities focused on development of the confirmation method applicable to radioactive wastes from research facilities.
Totsuka, Masayoshi; Kurosawa, Ryohei*; Sakai, Akihiro; Nakata, Hisakazu; Hayashi, Hirokazu; Amazawa, Hiroya
JAEA-Technology 2017-001, 40 Pages, 2017/03
Japan Atomic Energy Agency is planning for the near surface disposal of low level radioactive wastes generated from research, industrial and medical facilities industry in Japan. This document provides the values of radioactivity concentrations equivalent to dose criterion for trench-type disposal. These values are derived based on the safety assessment for ground water scenarios by using a model which describes the release of radionuclides from wastes to a cover soil caused by elution. These concentrations are compared with the one calculated by a model that describes the nuclide release mechanisms as solid-liquid partitioning equilibrium. Additionally, the change in the concentrations is evaluated when the amount of water percolating into a disposal facility varies.
Tsujimoto, Kazufumi; Sasa, Toshinobu; Maekawa, Fujio; Matsumura, Tatsuro; Hayashi, Hirokazu; Kurata, Masaki; Morita, Yasuji; Oigawa, Hiroyuki
Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.657 - 663, 2015/09
To continue the utilization of the nuclear fission energy, the management of the high-level radioactive waste is one of the most important issues to be solved. Partitioning and Transmutation technology of HLW is expected to be effective to mitigate the burden of the HLW disposal by reducing the radiological toxicity and heat generation. The Japan Atomic Energy Agency (JAEA) has been conducting the research and development on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. This paper overviews the recent progress and future R&D plan of the study on the ADS and related fuel cycle technology in JAEA.
Hayashi, Hirokazu; Nishi, Tsuyoshi*; Sato, Takumi; Kurata, Masaki
Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1811 - 1817, 2015/09
Transmutation of long-lived radioactive nuclides including minor actinides (MA: Np, Am, Cm) has been studied in Japan Atomic Energy Agency (JAEA). Accelerator-driven system (ADS) is regarded as one of the powerful tools for transmutation of MA under the double strata fuel cycle concept. Uranium-free nitride fuel was chosen as the first candidate fuel for MA transmutation using ADS. To improve the transmutation ratio of MA, reprocessing of spent fuel and reusing MA recovered from the spent fuels is necessary. Our target is to transmute 99% of MA arisen from commercial power reactor fuel cycle, with which the period until the radiotoxicity drops below that of natural uranium can be shorten from about 5000 years to about 300 years. A pyrochemical process has been proposed as the first candidate for reprocessing of the spent nitride fuel. This paper overviews the current status of the nitride fuel cycle technology. Our recent study on fuel fabrication, fuel property measurements, reprocessing of spent fuel, development of the property database of MA nitride fuel, and fuel behavior simulation code are introduced. Our research and development (R&D) plan based on the roadmap of the development is also introduced.
Sato, Takumi; Shibata, Hiroki; Hayashi, Hirokazu; Takano, Masahide; Kurata, Masaki
Journal of Nuclear Science and Technology, 52(10), p.1253 - 1258, 2015/08
In order to explore the applicability of the chlorination by MoCl as a potential pretreatment technique for waste treatment of fuel debris by pyrochemical methods, chlorination experiments of UO and (UZr)O simulated fuel debris were carried out in two steps: the first one is a chlorination reaction by homogeneous heating, the second one is a volatilization of molybdenum by-product by heating under temperature gradient condition. Most of UO and (UZr)O powder were converted to UCl or UCl and ZrCl mixture at 573 K, respectively. In the case of (UZr)Osintered particle, most of sample was converted to the chlorides because the products evaporated and be separated from sample surface at 773 K, while only the surface of the sample disk was converted to the chlorides at 573 and 673 K. Most of molybdenum by-product and ZrCl were separated from UCl by volatilization at 573 K.
Shibata, Hiroki; Hayashi, Hirokazu; Koyama, Tadafumi*
Denki Kagaku Oyobi Kogyo Butsuri Kagaku, 83(7), p.532 - 536, 2015/07
The electrochemical properties of curium in a LiCl-KCl eutectic melt were studied in the temperature range of 718-823 K. A small electrochemical cell used in this study was designed for the electrochemical measurement with a small amount (1-20 mg) of the highly radioactive minor actinides contained in molten salts achieved in a hot cell. Our data of apparent standard potentials of a Cm/Cm couple are reasonably in agreement with Osipenko's data (2011) and are lower than Martinot's data (1975). The validity of our data and the reported apparent standard potentials were discussed.
Hayashi, Hirokazu; Nishi, Tsuyoshi; Takano, Masahide; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki
NEA/NSC/R(2015)2 (Internet), p.360 - 367, 2015/06
Uranium-free nitride fuel was chosen as the first candidate for transmutation of long-lived minor actinides (MA) using accelerator-driven system (ADS) under the double strata fuel cycle concept by Japan Atomic Energy Agency (JAEA). The advantages of nitride fuel are good thermal properties and large mutual solubility among actinide elements. A pyrochemical process is proposed as the first candidate for the reprocessing of the spent nitride fuel, because this technique has some advantages over aqueous process, such as the resistance to radiation damage, which is an important issue for the fuels containing large amounts of highly radioactive MA. This paper overviews the recent progress and future R&D plan of the study on the nitride fuel cycle technology in JAEA.
Hayashi, Hirokazu; Soejima, Goro; Mizui, Hiroyuki; Sano, Kazuya
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
In the Fugen Nuclear Power Plant, we are going to conduct appropriate classification of the waste according to the contamination level of the material of the plant, to reduce the amount of radioactive waste and to promote dismantling work rationally and efficiently. For this reason, we are going to apply the clearance system to the dismantled material generated from dismantling work of the turbine system, and to reduce the radioactive waste amount as much as possible. In order to operate the clearance system properly, the target nuclides need to be selected accurately, and the evaluation method of them should be established. The assessment was conducted as follows.
Masubuchi, Takashi*; Hyuga, Hirokazu*; Ueda, Ryoshiro; Hayashi, Hidenori*; Ikenaga, Hiroshi*; Sato, Katsuya; Ono, Yutaka
JAEA-Review 2014-050, JAEA Takasaki Annual Report 2013, P. 123, 2015/03
Hayashi, Hirokazu; Akabori, Mitsuo; Minato, Kazuo
Journal of Radioanalytical and Nuclear Chemistry, 303(2), p.1331 - 1334, 2015/02
Electrochemical behavior of Am in NaCl-2CsCl melt at 823 K was investigated by transient electrochemical techniques such as cyclic voltammetry and differential pulse voltammetry. The results show that Am(III) ion is reduced to Am metal by a two-step mechanism via Am(II) ion. Formal standard potential of Am(III)/Am(II) and that of Am(II)/Am(0) redox couples have been determined to be -2.73 and -2.97 V vs Cl/Cl, respectively.
Hayashi, Hirokazu; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki; Iwai, Takashi; Arai, Yasuo
Science China; Chemistry, 57(11), p.1427 - 1431, 2014/11
Nitride fuels have several advantages, such as high thermal conductivity and high metal density like metallic fuels, and high melting point and isotropic crystal structure like oxide fuels. Since the late 1990s, the partitioning and transmutation of minor actinides (MA) has been studied to decrease the long term radio-toxicity of high level waste and mitigate the burden on the final disposal. Japan Atomic Energy Agency (JAEA) has been proposing dedicated transmutation cycle using the Accelerator-Driven System (ADS) with the nitride fuels containing MA. We have been developing the nitride fuel cycle including pyrochemical process. Our focus is on electrolysis of nitride fuels and refabrication of nitride fuel from the recovered actinides because other processes are similar to the technology for the metal fuel treatment and have been studied elsewhere. In this paper, we summarized our activity on developments of the pyrochemical treatment of the spent nitride fuels.
Shibata, Hiroki; Hayashi, Hirokazu; Akabori, Mitsuo; Arai, Yasuo; Kurata, Masaki
Journal of Physics and Chemistry of Solids, 75(8), p.972 - 976, 2014/08
Gibbs free energies of formation of six Ce-Cd intermetallic compounds, CeCd, CeCd, CeCd, CeCd, CeCd and CeCd, were evaluated systematically using electrochemical techniques in the temperature range from 673 to 923 K in the LiCl-KCl-CeCl-CdCl molten salt bath. The linear dependence of the Gibbs free energies of formation on temperature yields to the enthalpies and entropies of formation of these intermetallic compounds. By extrapolating the molar Gibbs free energy of Ce-Cd intermetallic compounds to the Cd distillation temperature, it was clear that the molar Gibbs free energy of Ce in Ce-Cd intermetallic compounds decreases gradually from CeCd to CeCd and attains to the minimum value at CeCd. This suggests on the Cd distillation from the U-Pu-Ce-Cd alloy that the dissolution of U or Pu into CeCd should be mostly taken into consideration.
Takano, Masahide; Hayashi, Hirokazu; Minato, Kazuo
Journal of Nuclear Materials, 448(1-3), p.66 - 71, 2014/05
A powder sample of curium nitride (CmN) containing 0.35%-PuN and 3.59%-AmN was prepared by carbothermic nitridation of the oxide. The lattice expansion induced by self-irradiation damage at room temperature was measured as a function of time. The saturated a/a value was 0.43%, which is greater than those for transuranium dioxides available in literature. The undamaged lattice parameter at 2971 K was determined to be 0.502610.00006 nm. Temperature dependence of the lattice parameter was measured by a high temperature X-ray diffractometer in the temperature range up to 1375 K. The linear thermal expansion of the lattice from 293 to 1273 K is 0.964% and the corresponding thermal expansion coefficient is 9.84 10 K. Comparing with the other actinide nitrides, it was found that CmN lies between the higher expansion nitrides (PuN and AmN) and the lower expansion nitrides (UN and NpN).
Hayashi, Hirokazu; Takano, Masahide; Kurata, Masaki; Minato, Kazuo
Journal of Nuclear Materials, 440(1-3), p.477 - 479, 2013/09
Neptunium trichloride of high purity was synthesized by the solid-state reaction of neptunium nitride, which was prepared from the oxide by the carbothermic reduction method, and cadmium chloride in a similar manner as reported for synthesis of AmCl. Lattice parameters of hexagonal NpCl were determined from the X-ray diffraction pattern to be a = 0.7421 0.0006 nm and c = 0.4268 0.0003 nm, which fairly agree with the reported values (a = 0.742 0.001 nm and c = 0.4281 0.0005 nm). Melting temperature of NpCl was measured with about 1 mg of the sample which was hermetically encapsulated in a gold crucible using a differential thermal analyzer with heating and cooling rate of 10 K/min in an argon gas flow (50 mL/min). The melting temperature of NpCl was determined to 1070 3 K, which is close to the recommended value 107530 K, which was derived from the mean value of the melting temperature for UCl(1115K) and that for PuCl (1041 K).