How to confirm the polarity of the DTQ in the J-PARC DTL

Kosaka, Satoshi*; Arai, Sora; Nemoto, Yasuo*; Kitamura, Ryo

JAEA-Technology 2023-003, 34 Pages, 2023/05

JAEA-Technology-2023-003.pdf:3.79MB

The negative hydrogen (H) ion beam is accelerated with a peak current of 50 mA in the J-PARC linac. The linac consists of the H- ion source and four kinds of accelerating cavities; a radio-frequency quadrupole linac (RFQ), a drift-tube linac (DTL), a separated-DTL (SDTL), and an annular-ring coupled structure linac (ACS). The accelerating electric field exists between the drift-tubes in the tank of the DTL. The quadrupole magnet for the beam focusing (DTQ) is installed in the drift-tube. The DTQ at J-PARC employs electromagnets to generate the magnetic field. It is important to set correctly the polarity of the magnetic field of the DTQ for tuning the beam. In this paper, we report the method of confirming the magnetic field polarity of DTQ, that is, the current wiring polarity by measuring DC current using a clamp meter on the power supply side and visually confirming the cable wiring connection on the electromagnet side.

Chapter 4, Technology on transmutation fuels and targets, 4.1 Fuel fabrication technology, 4.1.4 Nitride fuel (Double-strata concept with ADS), 4.1.5 The other fuels and targets, 4.2 Evaluation of fuel behavior, 4.2.4 Nitride fuel (Double-strata concept with ADS), 4.2.5 The other fuels and targets

Arai, Yasuo

Bunri Henkan Gijutsu Soron, p.134 - 146, 2016/08

no abstracts in English

Pyrochemical treatment of spent nitride fuels for MA transmutation

Hayashi, Hirokazu; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki; Iwai, Takashi; Arai, Yasuo

Science China; Chemistry, 57(11), p.1427 - 1431, 2014/11

https://doi.org/10.1007/s11426-014-5205-9

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.

Evaluation of Gibbs free energies of formation of Ce-Cd intermetallic compounds using electrochemical techniques

Shibata, Hiroki; Hayashi, Hirokazu; Akabori, Mitsuo; Arai, Yasuo; Kurata, Masaki

Journal of Physics and Chemistry of Solids, 75(8), p.972 - 976, 2014/08

https://doi.org/10.1016/j.jpcs.2014.04.004

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.

Thermal conductivity of minor actinide nitride fuels

Nishi, Tsuyoshi; Takano, Masahide; Arai, Yasuo; Kurata, Masaki

Dai-34-Kai Nihon Netsu Bussei Shimpojiumu Koen Rombunshu, p.199 - 201, 2013/11

By installing the laser flash apparatus and the drop calorimeter in the glove box, the thermal diffusivity and the heat capacity measurements of nitride containing MA elements of long-lived radioactive nuclides were enabled. The sample holder and the platinum container were designed to measure the thermal diffusivity and the heat capacity of very small quantity of MA nitride samples. The thermal conductivities of MA nitride increased with temperature, unlike that of conventional oxide-type nuclear fuels. In addition, the thermal conductivities of MA nitride decreased with increasing Am contents. The thermal conductivity of ZrN-based MA nitride, which is proposed as a candidate material for the ADS fuel, was fitted to equations as functions of the temperature and ZrN concentration. The predicted values agreed well with the experimental ones, indicating that the thermal conductivity of nitride fuel for ADS can be predicted for a practical design.

Electrooxidation of hydrazine hydrate using Ni-La catalyst for anion exchange membrane fuel cells

Sakamoto, Tomokazu*; Asazawa, Koichiro*; Martinez, U.*; Halevi, B.*; Suzuki, Toshiyuki*; Arai, Shigeo*; Matsumura, Daiju; Nishihata, Yasuo; Atanassov, P.*; Tanaka, Hirohisa*

Journal of Power Sources, 234, p.252 - 259, 2013/07

https://doi.org/10.1016/j.jpowsour.2013.01.181

Simple formula to evaluate helium production amount in fast reactor MA-containing MOX fuel and its accuracy

Akie, Hiroshi; Sato, Isamu; Suzuki, Motoe; Serizawa, Hiroyuki; Arai, Yasuo

Journal of Nuclear Science and Technology, 50(1), p.107 - 121, 2013/01

https://doi.org/10.1080/00223131.2013.750066

A simple formula is developed for the evaluation of the helium production amount in the fast reactor fuel. For the subroutine use in the existing fuel behavior analysis code, the formula is designed putting emphasis on simplicity and quickness rather than accuracy. The accuracy of the formula is confirmed by comparing with the detailed calculation with SWAT code, and also with the post irradiation examination (PIE) results of the fuel pin irradiated at the experimental fast reactor JOYO. As a result, the formula is found to evaluate the helium amount with the difference of less than about 10% from the detailed calculation and from the PIE results. Based on these results, the formula is installed in the fuel behavior analysis code for the simulation of helium behavior in fast reactor fuels.

Nitride fuel

Arai, Yasuo

Comprehensive Nuclear Materials, 3, p.41 - 54, 2012/03

https://doi.org/10.1016/B978-0-08-056033-5.00050-1

Development of nitride fuel was extensively reviewed in this paper, which will be published in the Comprehensive Nuclear Materials. Characteristics of nitride fuel and history of the R&D were described in the introductive chapter. In the chapter of fabrication, nitride fuel preparation by carbothermic reduction and its sintering behavior were centered on the chapter. In the chapter of irradiation behavior, results of the irradiation tests performed so far were reviewed and the in-pile behavior of nitride fuel was described. Furthermore, in the chapter of reprocessing, hydrochemical and pyrochemical reprocessing technologies of nitride fuel were briefly explained. Outlook of nitride fuel was given in the last chapter. On the other hand, physical, chemical and thermodynamic properties of nitride fuel will be reviewed in another part of the Comprehensive Nuclear Materials.

Fundamental research on actinide materials for sustainable fuel cycles in JAEA

Arai, Yasuo

Procedia Chemistry, 7, p.425 - 430, 2012/00

https://doi.org/10.1016/j.proche.2012.10.066

The fundamental research on actinide materials has been carried out in order to contribute to the development of future nuclear fuel cycle and actinide science database. Among actinide materials, the R&D has been focused on Pu and minor actinide (MA; Np, Am, Cm) bearing compounds. The chemical forms of actinide compounds concerned include oxides, nitrides, chlorides and alloys, which are prepared, characterized and subjected to property measurements. In this paper those results on Pu and MA bearing oxides obtained in recent several years are summarized. In addition, the possible challenges of actinide materials research to the subjects of post severe accident of Fukushima Dai-ichi Nuclear Power Station are briefly discussed.

U-Pu-Zr metal fuel fabrication for irradiation test at JOYO

Nakamura, Kinya*; Kato, Tetsuya*; Ogata, Takanari*; Nakajima, Kunihisa; Iwai, Takashi; Arai, Yasuo

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00

http://www-pub.iaea.org/mtcd/meetings/PDFplus/2009/cn176/cn176_BoeS.pdf

The first irradiation campaign of U-Pu-Zr metal fuel in Japan is planned in the experimental fast reactor JOYO. In the fabrication of U-Pu-Zr fuel, two methods were adopted for preparing U-Pu alloy from the oxide; one is the electrochemical reduction and the other is the electrorefining followed by reductive extraction. Injection casting for U-Pu-Zr slug was carried out after adding U and Zr metals to meet the target specifications of the irradiated fuel. Several conditions of Na-bonding process were determined from the results of tests using simulated metal fuel pins. Based on these results, six U-Pu-Zr fuel pins for the irradiation tests are now being fabricated.