JAEA-JRC collaborative development of delayed gamma-ray spectroscopy for nuclear safeguards nuclear material accountancy

Rodriguez, D.; Abbas, K.*; Bertolotti, D.*; Bonaldi, C.*; Fontana, C.*; Fujimoto, Masami*; Geerts, W.*; Koizumi, Mitsuo; Macias, M.*; Nonneman, S.*; et al.

Proceedings of INMM & ESARDA Joint Annual Meeting 2023 (Internet), 8 Pages, 2023/05

Leaching behavior of radionuclides from samples prepared from spent fuel rod comparable to core debris in the 1F NPS

Onishi, Takashi; Maeda, Koji; Katsuyama, Kozo

Journal of Nuclear Science and Technology, 58(4), p.383 - 398, 2021/04

https://doi.org/10.1080/00223131.2020.1822943

Determination of Pd in Pd purified by selective precipitation from spent nuclear fuel by laser ablation ICP-MS

Asai, Shiho; Ohata, Masaki*; Yomogida, Takumi; Saeki, Morihisa*; Oba, Hironori*; Hanzawa, Yukiko; Horita, Takuma; Kitatsuji, Yoshihiro

Analytical and Bioanalytical Chemistry, 411(5), p.973 - 983, 2019/02

https://doi.org/10.1007/s00216-018-1527-3

Determination of radiopalladium Pd is required for ensuring the radiation safety of Pd extracted from spent nuclear fuel for recycling or disposal. We employed laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) to simplify an analytical procedure of Pd. Pd was separated through selective Pd precipitation reaction from spent nuclear fuel. Laser ablation allows direct measurement of the Pd precipitates, skipping the dissolution and dilution procedure. In this study, Pd in natural Pd standard solution was used as an internal standard, taking advantage of its absence in spent nuclear fuel. The Pd precipitate was uniformly embedded on the surface of the centrifugal filter, forming a microscopically thin flat surface of Pd. The resulting homogeneous Pd layer is suitable for obtaining a stable signal ratio of Pd/Pd. The amount of Pd obtained by LA-ICP-MS corresponds to the values obtained by conventional solution nebulization measurement.

Effects of -radiation on a direct disposal system for spent nuclear fuel, 2; Review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North America

Kitamura, Akira; Takase, Hiroyasu*; Metcalfe, R.*; Penfold, J.*

Journal of Nuclear Science and Technology, 53(1), p.19 - 33, 2016/01

https://doi.org/10.1080/00223131.2015.1014874

Not only geological disposal of vitrified waste generated by spent fuel (SF) reprocessing, but also the possibility of disposing of SF itself in deep geological strata (hereinafter "direct disposal of SF") may be considered in the Japanese geological disposal program. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of -radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.

Influence of contaminants from spent fuel pools at the Fukushima Daiichi Nuclear Power Station on the reprocessing process

Aihara, Haruka; Kitawaki, Shinichi; Nomura, Kazunori; Taguchi, Katsuya

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1076 - 1083, 2015/09

On the requirement for remodelling the spent nuclear fuel transportation casks for research reactors; A Review of the drop impact analyses of JRC-80Y-20T

Review Group on the Structure of the Spent Nuclear Fuel Transportation Casks for

JAERI-Review 2005-023, 133 Pages, 2005/07

JAERI-Review-2005-023.pdf:18.88MB

The Japan Atomic Energy Research Institute (JAERI) constructed two stainless steel transportation casks, JRC-80Y-20T, for spent nuclear fuels of research reactors and had utilized them for transportation since 1981. A modification of the design was applied to the USA for transportation of silicide fuels. Additional analyses employing the impact analysis code LS-DYNA that was often used for safety analysis were submitted by the JAERI to the USA to show integrity of the packages; the casks were still not approved, because inelastic deformation was occurred on the surface of the lid touching to the body. To resolve this problem on design approval of transportation casks, a review group was formed at the end of this June. The group examined the impact analyses by reviewing the input data and performing the sensitivity analyses. As the drop impact analyses were found to be practically reasonable, it was concluded that the approval of the USA for the transportation casks could not be obtained just by revising the analyses; therefore, remodelling the casks is required.

ARTIST process; A Novel chemical process for treatment of spent nuclear fuel

Tachimori, Shoichi

JAERI-Research 2001-048, 23 Pages, 2001/10

JAERI-Research-2001-048.pdf:1.88MB

A new chemical process, ARTIST process, is proposed for the treatment of spent nuclear fuel. The main concept of the ARTIST process is to recover and stock all actinides (Ans) in two groups, uranium (U) and a mixture of transuranics (TRU), to preserve their resource value and to dispose solely fission products (FPs). The process composed of two main steps, an U exclusive isolation and a total recovery of TRU; which copes with the nuclear non-proliferation measures, and additionally Pu separation process and soft N-donor process if requested, and optionally processes for separation of long-lived FPs. These An products: U-product and TRU-product, are to be solidified by calcination and allowed to the interim stockpile for future utilization. These separations are achieved by use of amidic extractants in accord with the CHON principle. The technical feasibility of the ARTIST process was explained by the performance of both the branched-alkyl monoamides the diglycolic amide (TODGA) in thorough extraction of all TRU by tridentate fashon.

Effect of axial temperature gradient on oxidation of Zircaloy-2 cladding under high temperature steam

Ioka, Ikuo; Kato, Hitoshi; Ogawa, Hiroaki

no journal, , 

When the function of cooling system for a spent fuel pool is lost, the spent fuel pin is exposed to steam and air environment. In addition, oxidation behavior of the cladding may be changed due to axial temperature gradient and induced stress gradient during the process of dry out of the spent fuel pool. The oxidation behavior of the Zircaloy-2 cladding under axial temperature gradient was investigated in this study. The axial temperature gradients was about 100 C/cm. The oxidation test was carried out at 600 C in the saturated steam flow with Ar of 0.5 l/min as a career gas. Little difference was seen in the configuration of the surface cracks and the oxide thickness of specimens oxidized with different temperature gradients. Consequently, the high-temperature oxidation of Zircaloy-2 cladding was hardly changed by the steep axial temperature gradient of about 100 C/cm in this study.

Anticipations of NRF-based NDA of nuclear material using monochromatic -ray beams

Seya, Michio; Hajima, Ryoichi*; Hayakawa, Takehito*; Koizumi, Mitsuo

no journal, , 

The NRF-base NDA using monochromatic -rays would be used in nuclear security for secure detection of NM in thick shield and also for precise checking of interior structures of detected / suspicious objects. This NDA also would be used in nuclear safeguards for precise quantification of U/Pu isotopes in high radiation background, such as spent fuel assemblies / melted fuel debris in a canister. It is also useful for precise measurement of quantities of minor actinide isotopes in ADS (Accelerator Driven Sub-critical System; for transmutation of long-half-life minor actinides (MA)) fuels before and after irradiation in the ADS reactor core. In this presentation, we show actual NRF-NDA methods for these objects.

Oxidative uranium dissolution from UO in the presence of adsorbed phthalic acid

Kumagai, Yuta; Jonsson, M.*

no journal, , 

Contact of water with spent nuclear fuel is anticipated in scenarios of failure of the repository system for the direct disposal of spent fuel. Upon the direct contact of water, the UO matrix of the fuel is expected to gradually dissolve due to oxidation of uranium by the action of ionizing radiation. In this study, we examined effects of organic acid on the UO dissolution by using phthalic acid as a model compound. We investigated oxidation of UO by exposure to HO in aqueous solution containing phthalic acid. Significant adsorption of phthalic acid on UO was observed. The coverage of the surface was estimated to reach 80 %. The HO-exposure experiments revealed that adsorbed phthalic acid has no significant effect on the redox reaction by HO on the UO surface, despite the high surface density. HO oxidation of UO with adsorbed phthalic acid resulted in U dissolution to similar extents with the U dissolution measured in aqueous bicarbonate solutions.

Challenging and development on radiation hardend laser remote analysis for in-situ screening of nuclear fuel debris in decommissioning of Fukushima Daiichi Nuclear Power Station

Wakaida, Ikuo; Oba, Hironori; Akaoka, Katsuaki; Karino, Takahiro; Nakanishi, Ryuzo*; Sakamoto, Kan*; Ikeda, Yuji*; Taira, Takunori*

no journal, ,