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Udagawa, Yutaka; Fuketa, Toyoshi*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi
Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08
Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.
Kato, Chiaki
Comprehensive Nuclear Materials, 2nd Edition, Vol.4, p.528 - 563, 2020/08
In spent fuel reprocessing plants, various nitric media are encountered throughout the PUREX process, used in the separation of fission products, uranium, and plutonium. The PUREX process is thus highly corrosive as it takes place at high temperatures under high concentrations of nitric acid solution containing oxidizing metal ions from spent fuel. In this review, the unique chemical properties of nitric acid are first described. Secondly, the process of oxidizing power generation in boiling nitric acid under heat transfer is described using the redox potential and a thermodynamic model of boiling nitric acid. Finally, the corrosion behavior and corrosion acceleration mechanism specific to the reprocessing environments are described from the perspective of solution chemistry.
Minato, Kazuo; Ogawa, Toru
Comprehensive Nuclear Materials, 2nd Edition, Vol.5, p.334 - 360, 2020/08
TRISO coated particle fuel has been developed for the high temperature gas-cooled reactors, which consists of microspherical fuel kernel and coating layers of pyrolytic carbon and silicon carbide. To improve the high temperature stability, the resistance to the chemical attack by fission products and the retention of fission products of the TRISO coated particle fuels, several types of advanced fuels were proposed and tested. Coated particle fuels for fast reactors were also proposed and tested. In this paper, fuel designs, fabrications, characterization techniques and fuel performance of these advanced coated particle fuels are systematically described. This is the updated version of the paper having the same title in Comprehensive Nuclear Materials published in 2012.
Uno, Masayoshi*; Nishi, Tsuyoshi*; Takano, Masahide
Comprehensive Nuclear Materials, 2nd Edition, Vol.7, p.202 - 231, 2020/08
On the thermodynamic and thermophysical properties of the actinide nitrides in Comprehensive Nuclear Materials published by Elsevier as the first edition in 2012, we have revised them by adding some brand-new data. The main topics added are the solid solubility of the actinide nitrides into the zirconium nitride matrix for transmutation fuel, the lattice expansion of actinide nitrides induced by self-irradiation damage, the influence of defects accumulation on thermal conductivity, and the thermal expansion in curium nitride lattice.
Uno, Masayoshi*; Nishi, Tsuyoshi; Takano, Masahide
Comprehensive Nuclear Materials, 2, p.61 - 85, 2012/03
In a review "Comprehensive Nuclear Materials", we wrote the thermophysical and mechanical properties of actinide nitrides. In Chapter 1, the introductive remark was shown. In Chapter 2, the phase diagram and crystal structure of actinide nitrides were summarized. In Chapter 3, the melting or decomposition temperature, equilibrium vapor pressure, heat capacity, Gibbs free energy of formation and thermal conductivity of actinide nitrides were described, including recent experimental data obtained in JAEA. In Chapter 4, the mechanical properties of actinide nitrides and thermal expansion data obtained in JAEA were shown. On the other hand, the characteristic of nitride fuel and its historical development were not described because there appeared in another chapter of this review.
Nagase, Fumihisa
Comprehensive Nuclear Materials, 2, p.595 - 608, 2012/03
Comprehensive Nuclear Materials will provide broad ranging, validated summaries of all of the major topics in the field of nuclear materials research for fission as well as fusion reactor system. This work will consist of an extensive collection of comprehensive review articles written by a team of leading experts. The Loss-of-Coolant Accident (LOCA) is one of the postulated accidents for LWRs. A LOCA is caused by a break in the reactor coolant pressure boundary. The temperature of the reactor core continues to rise due to the radioactive decay in the fuel as well as the reduction or loss of coolant. During the high-temperature transient, various changes and phenomena occur in the fuel rod. This chapter describes phase transition and plastic deformation of cladding during heat-up, oxidation and embrittlement of cladding, and effect of long-term irradiation on the fuel behavior under LOCA conditions.
Fuketa, Toyoshi
Comprehensive Nuclear Materials, 2, p.579 - 593, 2012/02
