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Journal Articles

Characterization study of four candidate technologies for nuclear material quantification in fuel debris at Fukushima Daiichi Nuclear Power Station, 4; Numerical simulations for active neutron technique

Komeda, Masao; Maeda, Makoto; Ozu, Akira; Kureta, Masatoshi; Toh, Yosuke

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

We have developed a special Nuclear Material Accountancy (NMA) technique using the Fast Neutron Direct Interrogation (FNDI) method which is one of active neutron techniques. A measurement system, for fuel debris at Fukushima Daiichi Nuclear Power Station, implemented in the simulation was designed as follows. This system has a neutron generator, which can produce neutron yield of 1$$times$$10$$^{6}$$ per pulse in 1 kHz repetition rate. The length of the system is 140 cm, and the outer diameter is 80cm. Sixteen He-3 detectors, 100 cm in length and 2.5 cm in diameter, are installed. Simulations were carried out using the Monte Carlo code MVP developed at JAEA. This work provides simulation results and the applicable range of the FNDI method for fuel debris, using various debris model parameters for example, burn-up composition and heterogeneous materials.

Journal Articles

Reduction and resource recycling of high-level radioactive wastes through nuclear transmutation; Isolation techniques of Pd, Zr, Se and Cs in simulated high level radioactive waste using solvent extraction

Sasaki, Yuji; Morita, Keisuke; Ito, Keisuke; Suzuki, Shinichi; Shiwaku, Hideaki; Takahashi, Yuya*; Kaneko, Masaaki*; Omori, Takashi*; Asano, Kazuhito*

Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 4 Pages, 2017/09

no abstracts in English

Journal Articles

Actinides recovery from irradiated fuel for SmART cycle

Sano, Yuichi; Watanabe, So; Nakahara, Masaumi; Aihara, Haruka; Takeuchi, Masayuki

Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 4 Pages, 2017/09

JAEA has been promoting MA recycle project using a FR fuel cycle named as SmART cycle concept. The SmART cycle contains the recovery of all actinides, in which total amount of MA is estimated to around 1-2g, at CPF from the FR Joyo spent fuel, the fabrication of MA bearing MOX fuel pellets and pins at AGF with recovered actinides, and the irradiation test of the fabricated fuels at the Joyo. In this paper, recent activities on actinides recovery in CPF, which will make a significant contribution to the SmART cycle, were summarized.

Journal Articles

Research of process to treat the radioactive liquid waste containing chloride ion generated by pyroprocessing plant in operating

Tada, Kohei; Kitawaki, Shinichi; Watanabe, So; Aihara, Haruka; Shibata, Atsuhiro; Nomura, Kazunori

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

Radioactive liquid waste containing chloride ion (Cl) is generated by chemical analysis for process control of pyroprocessing. To realize discharging this liquid waste to the sea, it's necessary to carry out the process in order to separate Cl and recover U, Pu. This study carried out a combination of the AgCl precipitation method and extraction chromatography method to separate Cl and recover U, Pu. The result of precipitation test showed that U and Pu didn't occur the co-precipitation after the test. The result of solid phase extraction test showed that 95% of Pu was successfully recovered from the liquid waste. It was difficult to analyze $$alpha$$ radioactivity about U because the concentration of U is not enough. These results showed that these process has the feasibility of the discharging the liquid waste to the sea.

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.

Oral presentation

Reduction and resource recycling of high-level radioactive wastes through nuclear transmutation; Impact of nuclear transmutation on disposal of high-level radioactive waste

Nishihara, Kenji; Makino, Hitoshi; Koo, Shigeru*

no journal, , 

In the ImPACT project, transmutation technology of long-lived fission products (LLFPs) is being developed in addition to minor actinides (MAs) transmutation studied in other programs. If MAs, LLFPs, and heat-generating fission products (FPs), which are Sr-90 and Cs-137, are removed from high-level waste (HLW), drastic benefit is anticipated on the disposal of HLW. In the present study, we tried to estimate two impacts through transport analysis of radionuclide: (1) disposal of the waste after transmutation in the deep underground repository designed for the conventional HLW, and (2) disposal of the waste after transmutation in the intermediate depth disposal designed for the low level radioactive wastes such as hull and end-pieces of the spent fuel assembly. As the result, the reduction of public dose in analysis (1) was observed, and the dose was enough small in analysis (2).

Oral presentation

Characterization study of four candidate technologies for nuclear material quantification in fuel debris at Fukushima Daiichi Nuclear Power Station, 1; General outline

Nagatani, Taketeru; Komeda, Masao; Shiba, Tomooki; Nauchi, Yasushi*; Sagara, Hiroshi*; Kosuge, Yoshihiro*; Okumura, Keisuke; Maeda, Makoto; Toh, Yosuke; Hori, Keiichiro

no journal, , 

Oral presentation

Thermal degradation and vitrification properties of MA adsorbents

Kofuji, Hirohide; Watanabe, So; Goto, Ichiro; Oriuchi, Akio; Takeuchi, Masayuki; Kobayashi, Hidekazu; Sasage, Kenichi

no journal, , 

Vitrification procedure for extraction chromatography using porous silica particles covered with extractant was investigated. In this method, minor actinides (MAs) were separated from high-level radioactive liquid waste by extractant CMPO and/or HDEHP. In this study, thermal degradation behavior and vitrification properties were evaluated from the viewpoints of high-level wasteform properties using porous silica adsorbent impregnated with extractant CMPO and HDEHP. As the results of various experiments, suitable vitrification procedure for MA/RE adsorbents was clarified and selected adsorbent was generally well-vitrified and had enough chemical durability.

Oral presentation

Development of the technology for pyroprocessing of minor actinide nitride fuel

Hayashi, Hirokazu; Sato, Takumi; Tsubata, Yasuhiro

no journal, , 

Transmutation of long-lived radioactive nuclides including minor actinides (MA: Np, Am, Cm) is expected to be effective to reducing the burden of high level radioactive wastes and using repositories efficiently. 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 has been chosen as the first candidate fuel for MA transmutation using ADS in Japan Atomic Energy Agency (JAEA). 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. Pyroprocessing has been proposed as the first candidate for reprocessing of the spent MA nitride fuel. This paper overviews the current status of development of the technology for pyroprocessing of MA nitride fuel. It contains studies on detailed behavior of the elements that are thought to exist in spent MA nitride fuels, including fission product elements and inert matrix elements such as zirconium (Zr) and titanium (Ti). Progress of the development aiming at an engineering-scale apparatus for pyroprocessing of nitride fuel is also introduced.

Oral presentation

Current status and future plan of research and development on partitioning and transmutation based on double-strata concept in JAEA

Tsujimoto, Kazufumi; Hayashi, Hirokazu; Matsumura, Tatsuro; Takano, Masahide

no journal, , 

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.

Oral presentation

Japan Atomic Energy Agency's Okuma Analysis and Research Center for decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station

Ogawa, Miho; Sakazume, Yoshinori; Inoue, Toshihiko; Yoshimochi, Hiroshi; Koyama, Shinichi; Koyama, Tomozo; Nakayama, Shinichi

no journal, , 

Decommissioning of TEPCO's 1F is in progress according to the Roadmap. The Roadmap assigned the construction of a hot laboratory and analysis to the JAEA. The hot laboratory, Okuma Analysis and Research Center consists of the three buildings; Administrative building, the Laboratory-1 and Laboratory-2. The Laboratory-1 and Laboratory-2 are hot laboratories. Laboratory-1 is for radiometric analysis of low and medium level radioactive rubble and secondary wastes. The license of the Laboratory-1's implementation was approved by The Secretariat of the Nuclear Regulation Authority and the construction started in April 2017 and plans an operational start in 2020. Laboratory-2 provides concrete cells, steel cells for the analysis of the fuel debris and high level radioactive rubble. The Laboratory-2's major analysis items is reviewed by review meeting organized of cognoscente.

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