Absolute quantification of Cs activity in spent nuclear fuel with calculated detector response function

Sato, Shunsuke*; Nauchi, Yasushi*; Hayakawa, Takehito*; Kimura, Yasuhiko; Kashima, Takao*; Futakami, Kazuhiro*; Suyama, Kenya

Journal of Nuclear Science and Technology, 9 Pages, 2022/10

https://doi.org/10.1080/00223131.2022.2130462

A new non-destructive method for evaluating Cs activity in spent nuclear fuels was proposed and experimentally demonstrated for physical measurements in burnup credit implementation. Cs activities were quantified using gamma ray measurements and numerical detector response simulations without reference fuels, in which 137Cs activities are well known. Fuel samples were obtained from a lead use assembly (LUA) irradiated in a commercial pressurized water reactor (PWR) up to 53 GWd/t. Gamma rays emitted from the samples were measured using a bismuth germinate (BGO) scintillation detector through a collimator attached to a hot cell. The detection efficiency of gamma rays with the detector was calculated using the PHITS particle transport calculation code considering the measurement geometry. The relative activities of Cs, Cs, and Eu in the sample were measured with a high-purity germanium (HPGe) detector for more accurate simulations of the detector response for the samples. The absolute efficiency of the detector was calibrated by measuring a standard gamma ray source in another geometry. Cs activity in the fuel samples was quantified using the measured count rate and detection efficiency. The quantified Cs activities agreed well with those estimated using the MVP-BURN depletion calculation code.

Characterization study of four candidate technologies for nuclear material quantification in fuel debris at Fukushima Daiichi Nuclear Power Station (Interim report)

Nagatani, Taketeru; Komeda, Masao; Shiba, Tomooki; Maeda, Makoto; Nauchi, Yasushi*; Sagara, Hiroshi*; Kosuge, Yoshihiro*; Kureta, Masatoshi; Tomikawa, Hirofumi; Okumura, Keisuke; et al.

Proceedings of INMM 57th Annual Meeting (Internet), 10 Pages, 2016/07

Development of the 4S and related technologies, 7; Summary of the FCA XXIII experiment analyses towards evaluation of prediction accuracies for the 4S core characteristics

Ueda, Nobuyuki*; Fukushima, Masahiro; Okajima, Shigeaki; Takeda, Toshikazu*; Kitada, Takanori*; Nauchi, Yasushi*; Kinoshita, Izumi*; Matsumura, Tetsuo*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9493_1 - 9493_9, 2009/05

A series of critical experiments were carried out in the JAEA fast critical facility (FCA) named FCA XXIII cores with placing emphases on the reflector reactivity worth and the sodium void reactivity which are especially important from the view point of safety features of the 4S. The analyses of those physics mockup experiments have been carried out by the neutron transport calculation methods with continuous energy Monte Carlo code MVP and 70 energy-group discrete ordinate P0-S8 transport code DANTSYS using libraries processed from JENDL-3.3 data file. The results showed that combination of the stochastic and deterministic transport calculation methods (Monte Carlo and Sn) provided good prediction bases for criticality, reflector worth, sodium void reactivity, reaction rate ratios and absorber reactivity worth for the 4S nuclear design.

Activity ratio measurement and burnup analysis for PWR lead use assembly

Sato, Shunsuke*; Nauchi, Yasushi*; Hayakawa, Takehito; Kimura, Yasuhiko; Suyama, Kenya

no journal, , 

The ray spectrum measurement was conducted for a sample taken from a PWR lead used assembly and photon spectrum data for higher energy region than usual case was obtained. Burnup calculation adopting MVP-BURN code was carried out for the sample and the agreement between experimental data and the calculation results was confirmed.

Measurement and burnup analysis of Ru/Ce activity ratio for spent fuel cooled more than 10 years

Sato, Shunsuke*; Nauchi, Yasushi*; Hayakawa, Takehito*; Kimura, Yasuhiko; Suyama, Kenya

no journal, , 

It is difficult to measure -ray from short half-life nuclides such as Ru-106 and Ce in used fuel with more than 10 year cooling time. In this study, we make it possible to measure radiation of these nuclides by changing detection efficiency with install of polyethylene block. We obtained good agreement of activity ratio of Ru/Ce between measurement and calculation.

Characterization study of 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, , 

no abstracts in English

Characterization study of candidate technologies for nuclear material quantification in fuel debris at Fukushima Daiichi Nuclear Power Station, 6; Study on inventory evaluation method of nuclear fuel material using detector responses

Okumura, Keisuke; Sato, Wakaei*; Nagatani, Taketeru; Komeda, Masao; Shiba, Tomooki; Nauchi, Yasushi*; Takada, Akira*

no journal, , 

In order to contribute to the development of quantitative technology for the nuclear fuel materials in the fuel debris at the Fukushima Daiichi Nuclear Power Station, we examined the evaluation method and prediction accuracy of the nuclear fuel material quantity in the fuel debris based on the detector responces of the non-destructive assay simulation and on the fuel burnup calculation results.

Study on nuclear material management for fuel debris

Miyaji, Noriko; Takada, Akira*; Iwafuchi, Junichi; Tomikawa, Hirofumi; Shiba, Tomooki; Okumura, Keisuke; Nagatani, Taketeru; Nauchi, Yasushi*

no journal, , 

Nuclear materials in fuel debris need to be properly managed in order to assure to the international and national communities that no nuclear materials have been diverted. In order to provide this assurance, specific measures appropriate to the characteristics of the fuel debris are required. There is experience at managing fuel debris in similar situations, as it was the case of the partial meltdown at the TMI-2 in the US. The nuclear material in the fuel debris retrieved from the reactor was quantified in order to fulfill domestic law requirements in the US. NDA, DA, and visual observation were applied to the residual fuel in the reactor. On that specific accident, IAEA safeguards were not applied because the USA is a nuclear weapon state and therefore this power plant was not inspected by the IAEA. Nuclear material accounting (NMA)is an important measure to manage nuclear materials for states under IAEA safeguards. In this presentation a report on NMA as a measure to manage fuel debris will be discussed.

Measurement test of non-destructive assay for quantification of nuclear materials in fuel debris, 3; Determination of impurity in nuclear materials by using gamma-ray spectroscopy

Suzuki, Risa; Nomi, Takayoshi; Nagatani, Taketeru; Shiromo, Hideo; Shiba, Tomooki; Kaburagi, Masaaki; Okumura, Keisuke; Kosuge, Yoshihiro*; Takada, Akira*; Nauchi, Yasushi*

no journal, , 

no abstracts in English

Measurement test of non-destructive assay for quantification of nuclear materials in fuel debris, 1; Background and purpose

Okumura, Keisuke; Nauchi, Yasushi*; Suzuki, Risa; Nomi, Takayoshi; Shiba, Tomooki; Kaburagi, Masaaki; Takada, Akira*; Kosuge, Yoshihiro*; Nagatani, Taketeru

no journal, , 

no abstracts in English