Isotope dilution inductively coupled plasma mass spectrometry for determination of Sn content in spent nuclear fuel sample

Asai, Shiho; Toshimitsu, Masaaki; Hanzawa, Yukiko; Suzuki, Hideya; Shinohara, Nobuo; Inagawa, Jun; Okumura, Keisuke; Hotoku, Shinobu; Kimura, Takaumi; Suzuki, Kensuke*; et al.

Journal of Nuclear Science and Technology, 50(6), p.556 - 562, 2013/06

https://doi.org/10.1080/00223131.2013.785276

The Sn content in a spent nuclear fuel solution was determined by ICP-MS for its inventory estimation in high-level radioactive waste. An irradiated UO fuel was used as a sample to evaluate the reliability of the methodology. Prior to the measurement, Sn was separated from Te, which causes major isobaric interference in the determination of Sn content, along with highly radioactive coexisting elements using an anion-exchange column. The absence of counts attributed to Te in the Sn-containing effluent indicates that Te was completely removed. After washing, Sn retained on the column was readily eluted with 1 M HNO. The isotope ratios of Sn were successfully determined and showed good agreement with those obtained through ORIGEN2 calculations. The results reported in this paper are the first experimental values of Sn content in the spent nuclear fuel solution originating in spent nuclear fuel irradiated at a nuclear power plant in Japan.

Validation of correlations between Nd isotopes and difficult-to-measure nuclides predicted with burn-up calculation code by post irradiation examination

Asai, Shiho; Okumura, Keisuke; Hanzawa, Yukiko; Suzuki, Hideya; Toshimitsu, Masaaki; Inagawa, Jun; Kimura, Takaumi; Kaneko, Satoru*; Suzuki, Kensuke*

Proceedings of 14th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2011) (CD-ROM), p.1437 - 1442, 2011/09

https://doi.org/10.1115/ICEM2011-59086

Determination of Se and Cs in spent nuclear fuel for inventory estimation of high-level radioactive wastes

Asai, Shiho; Hanzawa, Yukiko; Okumura, Keisuke; Shinohara, Nobuo; Inagawa, Jun; Hotoku, Shinobu; Suzuki, Kensuke*; Kaneko, Satoru*

Journal of Nuclear Science and Technology, 48(5), p.851 - 854, 2011/05

https://doi.org/10.3327/jnst.48.851

Study on improvement of reliability on inventory assessment in vitrified waste for long-term safety of geological disposal

Ishikawa, Masumi*; Kaneko, Satoru*; Kitayama, Kazumi*; Ishiguro, Katsuhiko*; Ueda, Hiroyoshi*; Wakasugi, Keiichiro*; Shinohara, Nobuo; Okumura, Keisuke; Chino, Masamichi; Moriya Noriyasu*

Nippon Genshiryoku Gakkai Wabun Rombunshi, 8(4), p.304 - 312, 2009/12

https://doi.org/10.3327/taesj.J09.002

Since quality control issues for vitrified waste are defined mainly with the focus on the transport and storage of the waste rather than the long-term safety of geological disposal, they do not cover inventories of long-lived nuclides which are of most interest in the safety assessment of geological disposal. Therefore we suggest a flow chart for assessment of inventories of long-lived nuclides in the vitrified waste focusing on measured value. We started a programme to examine the applicability as well as to improve reliability of nuclide generation/decay code and nuclear data library using liquid waste from spent fuel with clear irradiation history. To solve the issue of quality control for vitrified waste, comprehensive study is needed in aspects not only of geological disposal field but also of operation of nuclear power plant, reprocessing of spent fuel and vitrification of liquid waste. This study is a pioneering study to integrate them.

Oxygen potential measurement of americium oxide by electromotive force method

Otobe, Haruyoshi; Akabori, Mitsuo; Minato, Kazuo

Journal of the American Ceramic Society, 91(6), p.1981 - 1985, 2008/06

https://doi.org/10.1111/j.1551-2916.2008.02374.x

The oxygen potentials of AmO were measured in the range of 0.01 to 0.5 and the temperature range of 1000 to 1333 K by the electromotive force (EMF) method. The oxygen potentials at 1333 K were -19.83 kJ/mol for =0.01 and -319.1 kJ/mol for =0.485, which were higher than those of CeO by approximately 200 kJ/mol for the corresponding values. From the dependence of the oxygen potentials on and temperature, a tentative phase diagram of Am-O system was proposed, which suggested the presence of the intermediate phases of AmO and AmO in the Am-O system.

Fission gas release in BWR fuel with a burnup of 56 GWd/t during simulated reactivity initiated accident (RIA) condition

Amaya, Masaki; Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 45(5), p.423 - 431, 2008/05

https://doi.org/10.3327/jnst.45.423

Pulse irradiation simulating reactivity initiated accident (RIA) condition was conducted for the test rod prepared from the BWR fuel rod with a burnup of 56 GWd/t irradiated in a commercial reactor, and fission gas release during the pulse irradiation was investigated based on the result of rod-puncture test and electron-probe-microanalysis of fuel pellet. The local xenon concentration of pulse-irradiated pellet decreased compared with that of base-irradiated pellet in the relative radius of 0 to approximately 0.8. The decrease corresponds to a fractional fission gas release of approximately 11%, and this value was comparable with the rod-puncture test result. Considering the microstructural change in fuel pellet and the amount of retained gas in grain boundary, it is likely that the fission gas release during pulse irradiation was affected by the grain boundary separation which occurred in the mid-radius rather than the peripheral region of fuel pellet during pulse irradiation.

Thermal conductivity of AmO

Nishi, Tsuyoshi; Takano, Masahide; Ito, Akinori; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo; Numata, Masami

Journal of Nuclear Materials, 373(1-3), p.295 - 298, 2008/02

https://doi.org/10.1016/j.jnucmat.2007.06.007

The thermal diffusivity of americium oxide was determined in the temperature range from 299 to 1473 K by a laser flash method. The thermal diffusivity of AmO decreased with increasing temperature. The thermal conductivity of AmO was estimated from the measured thermal diffusivity, the specific heat capacity and the bulk density. It was found that the thermal conductivity of AmO decreased with increasing temperature over the temperature range investigated. It was also found that the decrease in O/Am ratio during the thermal diffusivity measurements under vacuum resulted in a slight decrease in thermal conductivity of AmO.

Development of the simulation technology for the pyrochemical process of spent nuclear fuels

Hayashi, Hirokazu; Akabori, Mitsuo; Minato, Kazuo; Mizuguchi, Koji*; Kawabe, Akihiro*; Fujita, Reiko*

Denki Kagaku Oyobi Kogyo Butsuri Kagaku, 75(7), p.528 - 534, 2007/07

https://doi.org/10.5796/electrochemistry.75.528

The simulation code for the pyrochemical processing of spent nuclear fuels was developed to analyze experimental data, to predict experimental results, and to propose adequate conditions and processes. The Simulation code for Pyrochemical Reprocessing (SPR) is based on calculations of chemical equilibrium and electrochemical reactions. The code also includes the calculations of the current-potential distribution between the electrodes. Some calculations were made to simulate the experimental results on the electro-codeposition process of UO and PuO. The phenomena of the redox reactions between Pu and Pu ions and those between Fe and Fe ions were theoretically analyzed; these redox reactions cause the low current efficiency in the electro-codeposition process. The calculated current-potential distribution around the cathode corresponds to the observed distribution of the oxide deposited on the cathode.

Experiments on the behavior of americium in pyrochemical process

Hayashi, Hirokazu; Akabori, Mitsuo; Minato, Kazuo

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 3 Pages, 2005/10

For a basis of the future nuclear cycle, it is very important to understand and control the behavior of TRU (Np, Pu, Am, Cm) in the nuclear fuel cycle. Experimental study of pyrochemical process of fuels containing TRU requires the facility having not only shielding for -ray and neutron but also ability to keep a high purity inert gas atmosphere; because minor actinide chlorides can easily react with oxygen or water vapor in an atmosphere. The module for TRU high temperature chemistry (TRU-HITEC) had been installed to study the basic properties of TRU in the pyrochemical processes. In the present work, the behavior of Am in pyrochemical process was investigated by electrochemical methods.

Report of Examination of the Samples from Primary Loop Recirculation Piping (K1-PLR) at Kashiwazaki-Kariwa Nuclear Power Station Unit-1 (Contract Research)

The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi

JAERI-Tech 2004-049, 44 Pages, 2004/06

JAERI-Tech-2004-049.pdf:7.21MB

At the Kashiwazaki-Kariwa Nuclear Power Station Unit-1, indications of cracks were identified in a weld joint portion of the primary loop recirculation piping. To investigate the cause of cracks, TEPCO conducted a material examination on the specimen including the cracks. The present investigation was carried out to ensure transparency of the examination by providing JAERI's own evaluation report as a third party organization. The following findings were made; (1) A crack was observed near the weld region. (2) Intergranular cracking was observed at almost whole fracture surface. (3) Transgranular cracking was observed at the crack opening region. Increases of hardness by cold work were observed and the crack was initiated near the region where hardness value showed the highest. (4) Content of Cr was very slightly depleted in the vicinity of grain boundary. Based on the above results with the presence of tensile residual stress near the crack generated by welding process and dissolved oxygen contents in cooling water, the observed cracks were concluded to be stress corrosion cracking.