Status of R&D of high-performance monitoring system under sever accident

Tsuchiya, Kunihiko; Takeuchi, Tomoaki; Komanome, Hirohisa*; Miura, Kuniaki*; Araki, Masanori; Ishihara, Masahiro

Nihon Hozen Gakkai Dai-13-Kai Gakujutsu Koenkai Yoshishu, p.375 - 378, 2016/07

no abstracts in English

An Advanced aqueous reprocessing process for the next generation's nuclear fuel cycle

Mineo, Hideaki; Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Morita, Yasuji

Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1250 - 1255, 2003/11

An advanced aqueous reprocessing process has been proposed for the next generation fuel cycle. Key technologies applied to the process are: removal of I-129, separation of Np and FP(Tc) separation by selective reduction of Np(VI) and high acid scrubbing of Tc within a single cycle process, MA separation by extraction chromatography and Cs/Sr separation. U separation just after dissolution was supposed to be effective to reduce the required capacity of the following extraction step. Among them Np reduction rate in TBP solution was measured, which was found to be lower than that in aqueous solution. Using an improved flow sheet spent fuel test, based on the Np reduction test, was carried out and about 90% of Np was separated before U and Pu partitioning step.

A study on improvements in accuracy of nuclear data measurements using -ray spectroscopic methods

Furutaka, Kazuyoshi

JNC TN8400 2000-028, 70 Pages, 2000/10

JNC-TN8400-2000-028.pdf:1.71MB

This report describes the study done by the author as a postdoctoral research associate at Japan Nuclear Cycle Development Institute. This report is divided into two parts: improvements in accuracy in determination of thermal neutron capture cross sections, and improvements in accuracy of photo-nuclear absorption cross section measurements using the HHS. (1)In the measurements of thermal neutron capture cross sections using an activation method, accuracies of the final results attained are limited by (1) accuracy of -ray peak detection efficiencies, and (2) accuracies of -ray emission probabilities. In this study; to determine thermal neutron capture cross sections more accurately, the following researches have been done using a newly developed three-dimensional coincidence measurement system: (1)accurate determination of -ray standard sources using a - coincidence method, for precise calibration of -ray peak detection efficiency, and (2) development of a - coincidence measurement system using a plastic scintillation detector as a -ray detector, for the determination of -ray emission probabilities of short-lived nuclides, and measurement of -ray emission probabilities of Tc nuclide using the coincidence system. (2)To transform radioactive nuclides with small thermal neutron capture cross sections, use of photonuclear absorption reaction has been suggested. In order to transform these nuclides efficiently using the reaction, one has to know detailed behavior of the photo-absorption cross sections. In this study, a Monte-Carlo simulation code has been used to create a standard set of -ray response functions of the high-resolution high-energy spectrometer (HHS), to enable reliable analyses of the data obtained by the spectrometer.

Report on neutronic design calculational methods

; *; *; *

JNC TN8410 2000-011, 185 Pages, 2000/05

JNC-TN8410-2000-011.pdf:4.67MB

This report describes the neutronic design calculational methods used in Fuel Design and Evaluation Group in order to inform other related sections of FBR core analysis technology and hand down the technology. Especially we show the neutronics calculation procedures used for the conceptual design study of the advanced core with 127 pin bundle for MONJU that has been carried out in our group. The topics include effective cross section preparation calculations, two-dimensional depletion calculations, three-dimensional diffusion calculations, reactivity coefficient calculations, and control rod worth calculations. The calculational methods shown in this report are the standard neutronics calculation methods employed in our group at the moment. However, the improvement of calculation codes, the reduction of correction factors and uncertainties for design using the nuclear data obtained in the start-up test for MONJU and so on, and the update of nuclear data file will be planned in order to improve evaluation accuracies. Those may change the neutronic design calculational methods, but we decided to describe the present standard calculational methods in our group from the viewpoint of sharing information in JNC.

Modification of the evaluation model for Pu redistribution phenomena

; *;

JNC TN9400 2000-045, 64 Pages, 2000/03

JNC-TN9400-2000-045.pdf:2.47MB

During the irradiation, the Pu redistribution phenomena would occur in the FBR MOX fuel pellets. The phenomena would considerably affect on the thermal properties of the fuels, therefore, it is need to establish the evaluation method for Pu redistribution phenomena. ln JNC, the efforts for development of the evaluation model for the phenomena had been continued and the simple evaluation model was constructed in 1992. In this work, the modification of the simple model developed in JNC has been done and the following results were obtained. (1)Based on the recent data of the MOX fuel irradiation tests, the evaluation model for Pu redistribution phenomena constructed in l992 is modified. And the model is included into the fuel performance analysis code "CEDAR". (2)To calibrate the modified CEDAR code, it is confirmed that the uncertainty in the Pu concentration evaluation for the center of the fuel pellet at EOL is about 3wt.%. (3)Based on the results of the evaluations using the modified CEDAR code, it is found that, in the early stage of the irradiation, the Pu redistribution is controlled by the vapor transportation mechanism via pores, and after that, the Pu redistribution is kept in progress due to the thermal diffusion mechanism with the change of the Pu concentration due to the degradation of U and Pu by fissions. And it is also found that the O/M ratio dependence of the U-Pu inter diffusion coefficients would affect on the Pu redistribution mechanisms, in especial, in the early stage of the irradiation.

None

JNC TJ1400 99-003, 70 Pages, 1999/03

JNC-TJ1400-99-003.pdf:2.26MB

no abstracts in English

On the JHPS symposium "Sophistication of phantom and the strategy"

Yamaguchi, Yasuhiro

Hoken Butsuri, 33(3), p.253 - 254, 1998/09

https://doi.org/10.5453/jhps.33.253

no abstracts in English

A Test line newly installed in NUCEF and research program on advanced reprocessing process by utilizing it

Asakura, Toshihide; Uchiyama, Gunzo; Kihara, Takehiro; Hotoku, Shinobu; *; ; *; ; ; Fujine, Sachio; et al.

Proc. of 5th Int. Nucl. Conf. on Recycling, Conditioning and Disposal (RECOD '98), 1, p.746 - 753, 1998/00

no abstracts in English

None

PNC TJ1603 97-002, 66 Pages, 1997/03

PNC-TJ1603-97-002.pdf:1.7MB

no abstracts in English

None

PNC TN8410 96-284, 245 Pages, 1996/03

PNC-TN8410-96-284.pdf:6.22MB

None