Analysis of the radioactivity concentrations in radioactive waste generated from JRR-3, JRR-4 and JRTF facilities

Tobita, Minoru*; Konda, Miki; Omori, Takeshi*; Nabatame, Tsutomu*; Onizawa, Takashi*; Kurosawa, Katsuaki*; Haraga, Tomoko; Aono, Ryuji; Mitsukai, Akina; Tsuchida, Daiki; et al.

JAEA-Data/Code 2022-007, 40 Pages, 2022/11

JAEA-Data-Code-2022-007.pdf:1.99MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete, ash, ceramic and brick samples generated from JRR-3, JRR4 and JRTF facilities. In this report, we summarized the radioactivity concentrations of 24 radionuclides (H, C, Cl, Ca, Co, Ni, Sr, Nb, Tc, Ag, I, Cs, Ba, Eu, Eu, Ho, U, U, Pu, Pu, Pu, Am, Am, Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2020-2021.

Study on chemical form of tritium in coolant helium of high temperature gas-cooled reactor with tritium production device

Hamamoto, Shimpei; Ishitsuka, Etsuo; Nakagawa, Shigeaki; Goto, Minoru; Matsuura, Hideaki*; Katayama, Kazunari*; Otsuka, Teppei*; Tobita, Kenji*

Proceedings of 2021 International Congress on Advances in Nuclear Power Plants (ICAPP 2021) (USB Flash Drive), 5 Pages, 2021/10

Impurity concentrations of hydrogen and hydride in the coolant were investigated in detail for the HTTR, a block type high-temperature gas reactor owned by Japan. As a result, it was found that CH was 1/10 of H concentration, which was under the conventional detection limit. If the ratio of H to CH in the coolant is the same as the ratio of HT to CHT, the CHT has a larger dose conversion factor, and this compositional ratio is an important finding for the optimal dose evaluation. Further investigation of the origin of CH suggested that CH was produced as a result of a thermal equilibrium reaction rather than being released as an impurity from the core.

Analysis of the radioactivity concentrations in low-level radioactive waste generated from JRR-2, JRR-3 and hot laboratory facilities

Tobita, Minoru*; Haraga, Tomoko; Sasaki, Takayuki*; Seki, Kotaro*; Omori, Hiroyuki*; Kochiyama, Mami; Shimomura, Yusuke; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Data/Code 2019-016, 72 Pages, 2020/02

JAEA-Data-Code-2019-016.pdf:2.67MB

In the future, radioactive wastes which generated from research and testing reactors in Japan Atomic Energy Agency are planning to be buried for the near surface disposal. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes by the time it starts disposal. In order to contribute to this work, we collected and analyzed the samples generated from JRR-2, JRR-3 and Hot laboratory facilities. In this report, we summarized the radioactivity concentrations of 25 radionuclides (H, C, Cl, Co, Ni, Sr, Nb, Mo, Tc, Ag, Sn, I, Cs, Eu, Eu, U, U, U, Pu, Pu, Pu, Pu, Am, Am, Cm) which were obtained from radiochemical analysis of those samples.

Radioactivity analysis of metal samples taken from pipes of the Fugen, 4

Haraga, Tomoko; Tobita, Minoru*; Takahashi, Shigemi*; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Data/Code 2015-025, 52 Pages, 2016/03

JAEA-Data-Code-2015-025.pdf:1.92MB

Fugen Nuclear Power Station was shut down and now is under decommissioning. Many radioactivity concentration data of dismantled materials have to be accumulated to calculate the scaling factors of radioactive wastes and verify that the cleared dismantled materials conform to the clearance levels. A simple and rapid radioactivity determination method for radioactive waste samples was developed in Department of Decommissioning and Waste Management. For the demonstration, the simple and rapid radioactivity determination method was applied to metal samples, which were taken from dismantled pipes of Fugen. This report summarizes the radioactivity data obtained from the analysis of those samples.

Study on mechanism of inner duct wall failure within fuel subassembly during core disruptive accidents in an LMFBR; Results of parametric analyses for heat transfer

Toyooka, Junichi; Endo, Hiroshi*; Tobita, Yoshiharu; Takahashi, Minoru*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 13(2), p.35 - 50, 2014/05

https://doi.org/10.3327/taesj.J13.011

In the design of future sodium-cooled fast reactor, a design measure to prevent severe re-criticality events even in case of core disruptive accidents is considered. This design adopts inner duct within the fuel sub-assembly that should allow molten fuel ejection out of the core region. The effectiveness of this design is dependent on failure time of the duct and it depends significantly on heat transfer from the melting core materials to the duct. In the previous study by the authors, heat transfer from molten fuel/steel mixture to the inner duct was evaluated with a computer model simulation for an in-pile experiment performed in IGR (Impulse Graphite Reactor) focusing on demonstration of the design effectiveness. In the present study, possible uncertainties in the assumption and model parameters in the previous study were evaluated so that validity of the main conclusion of the previous study could be confirmed and re-enforced. This confirmation consisted of evaluation of necessary fuel-to-steel heat transfer area, effect of hydrodynamic fragmentation of steel droplets, steel-vapor condensation heat transfer onto the duct surface and fuel crust formation. Furthermore, possible effect of variation in fuel designs and transient scenarios to the heat transfer was evaluated changing steel volume fraction as the initial boundary conditions. It was concluded that the previous study was appropriate in representing the realistic situation and the conclusions in the previous study were enforced. An additional set of analysis showed that possible under-estimation of heat transfer from fuel/steel mixture to the duct could be enhanced with a condition where steel volume fraction is less. Future model improvement is preferable for this characteristic.

Radioactivity analysis of metal samples taken from pipes of the Fugen

Kameo, Yutaka; Haraga, Tomoko; Ishimori, Kenichiro; Shimada, Asako; Tobita, Minoru*; Takahashi, Shigemi*; Takahashi, Kuniaki

JAEA-Data/Code 2010-028, 32 Pages, 2011/02

JAEA-Data-Code-2010-028.pdf:1.62MB

The Fugen Nuclear Power Station was shut down and decommissioning of the Fugen has been implemented. To calculate the scaling factor of radioactive waste or advance the clearance of dismantled materials, a large number of radioactivity concentration data of dismantled materials have to be accumulated. For these reasons, the simple and rapid radioactivity determination method was applied for metal samples, which were taken from pipes of the Fugen. The present report is summarized analytical procedures and obtained radioactivity data of the Fugen pipe samples.

Study on Mixing Phenomena in T-Pipe Junction; Temperature Measurement Test in Pipe by Liquid Crystal Sheet

Tanaka, Masaaki; Kawashima, Shigeyo*; Igarashi, Minoru; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki

JNC TN9400 2003-117, 65 Pages, 2004/03

JNC-TN9400-2003-117.pdf:3.49MB

Temperature fluctuation due to mixing of hot and cold fluids gives thermal fatigue to the structure (thermal striping phenomena).Investigation of this phenomenon is significant for the safety of a fast breeder reactor, which uses liquid metal as a coolant. In Japan Nuclear Cycle Development Institute, experiments and numerical analyses have been carried out to understand this phenomenon and also to construct the evaluation rule, which can be applied to the design. A water experiment of fluid mixing in T-pipe with long cycle fluctuation (WATLON),which notices thermal striping phenomena in the T-pipe junction, is performed to investigate the key factor of mixing phenomena by reason of long cycle fluctuation observed in a plant. By the former visualization test, it was showed that the flow pattern of branch pipe jet could be classified into (A) impinging jet, (B) deflecting jet (C) re-attachment jet and (D) wall jet according to the inflow condition. It was confirmed that the each jet pattern could be predicted by the momentum ratio of the each piping fluid. In this study, a thermo-chromic liquid crystal sheet was put on the inner wall in the main pipe, and temperature field on the wall surface was visualized. We established a new method to convert the color image data to temperature data. And measurement uncertainty of this method was evaluated + and - about 2.0 [deg-C], using by the typical picture in the temperature calibration test. From the temperature fluctuation visualization test by liquid crystal sheet, the cold spot was formed in just downstream region from the outlet of the branch pipe in the cases of the wall jet and impinging jet. Since this cold spot moved in time, high value of temperature fluctuation intensity was shown around the cold spot. And the validity of this method was shown from the comparison of the thermocouple data installed in a wall surface with the temperature conversion result.

Charge-exchange neutral particle measurement of MeV energy range on JT-60U

Kusama, Yoshinori; Nemoto, Masahiro; Sato, Minoru; Tsukahara, Yoshimitsu; Tobita, Kenji; Takeuchi, Hiroshi; S.Petrov*; Afanassiev, V. I.*; Kozlovskij, S. S.*; A.Kislyakov*; et al.

Review of Scientific Instruments, 66(1), p.339 - 341, 1995/01

https://doi.org/10.1063/1.1146405

no abstracts in English

Ra interaction with Ca/Ba mineral in bentonite

Fujiwara, Kenso; Tomura, Tsutomu*; Tobita, Minoru*; Tits, J.*; Curti, E.*

no journal, , 

Ra-226 is a safety-relevant radionuclide in performance assessments of high-level waste repositories and has thus been deserved special attention. The aim of the proposed study is to provide new key data on the interaction of Ra with calcite and witherite. In this study, the interaction of Ca with calcite and Ba with witherite was measured by using Ca-45 and Ba-133.

Diffusion behavior of humic acid in compacted bentonite, 3; Diffusion mechanism of humic acid and neodymium

Iijima, Kazuki; Kurosawa, Seiichi*; Tobita, Minoru*; Kibe, Satoshi*; Ouchi, Yuji*

no journal, , 

Diffusion behavior of humic acid in compacted bentonite was investigated by through-diffusion method. Concentration profile and molecular weight distribution of humic acid in the compacted bentonite indicate that only low molecular weight fraction of humic acid is likely to diffuses. Difference of diffusion behavior arising from difference of molecular weight is considered to be taken into account in evaluation of diffusion parameters.