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Nuclear Science Research Institute, Sector of Nuclear Science Research
JAEA-Review 2023-009, 165 Pages, 2023/06
JAEA-Review-2023-009.pdf:5.76MB
Nuclear Science Research Institute (NSRI) is composed of Planning and Management Department and six departments, namely Department of Operational Safety Administration, Department of Radiation Protection, Engineering Services Department, Department of Research Reactor and Tandem Accelerator, Department of Criticality and Hot Examination Technology and Department of Decommissioning and Waste Management, and each department manages facilities and develops related technologies to achieve the "Medium- to Long-term Plan" successfully and effectively. And, four research centers which are Advanced Science Research Center, Nuclear Science and Engineering Center, Nuclear Engineering Research Collaboration Center and Materials Sciences Research Center, belong to NSRI. In order to contribute the future research and development and to promote management business, this annual report summarizes information on the activities of NSRI of JFY 2020 as well as the activity on research and development carried out by Collaborative Laboratories for Advanced Decommissioning Science, Nuclear Safety Research Center and activities of Nuclear Human Resource Development Center, using facilities of NSRI.
Yoshida, Ryoichiro; Yamane, Yuichi; Abe, Hitoshi
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.408 - 414, 2019/09
In a criticality accident, it is known that some kinds of radiolysis gases are generated mainly due to kinetic energy of fission fragments. Hydrogen gas (H
) is one of them, which is able to initiate explosion. The rate of H generation and its total amount can be estimated from the number of fission per second if its G value is known. In this study, it was tried to estimate G value of hydrogen gas (G(H)) by using the H concentration measured as time-series data in Transient Experiment Critical Facility (TRACY) which was carried out by Japan Atomic Energy Agency. There was time lag in the measured H concentration from its generation. To overcome those problems, measured profile of H concentration was reproduced based on a hypothetical model and its total amount was evaluated. Based on the model, the obtained G(H) was 1.2.
Tsujimura, Norio; Yoshida, Tadayoshi; Sanada, Yukihisa
JPS Conference Proceedings (Internet), 24, p.011013_1 - 011013_6, 2019/01
Tashiro, Shinsuke; Abe, Hitoshi
JAEA-Technology 2015-044, 20 Pages, 2016/03
JAEA-Technology-2015-044.pdf:1.26MB
In order to estimate public dose under a criticality accident in fuel solution of a fuel reprocessing plant, release behavior of radioiodine from the fuel solution to atmosphere is very important. In this report, time evolution of 
I concentration in gas phase of TRACY core tank was measured until the concentration in the solution decreased. Furthermore, cumulative release ratio (CRR) and release rate (RR) from the solution to the atmosphere of radioiodine were evaluated by applying previously-reported evaluation model. As a result, for the case of short transient criticality, RR of I became maximum at 1 hour later from the ending and almost constant after 8 hour later. Furthermore, relationship of each elapsed time between total fission number and release rate of I could be derived. On the other hand, for the case of long criticality excursion, such as JCO criticality accident, the CRR and RR of radioiodine increased monotonously with time.
Yamane, Yuichi
Journal of Nuclear Science and Technology, 52(11), p.1425 - 1435, 2015/11
Times Cited Count：2 Percentile：17.57(Nuclear Science & Technology)A simple equation for the first peak power in a criticality accident due to instantaneous reactivity insertion into nuclear fuel solution system has been developed to improve the accuracy in the estimation of the first peak power keeping the easiness of calculation. The equation is based on the assumption that temperature feedback reactivity is a second order function of an increase in fuel temperature. Peak power estimated using the equation was in a range between about a half and twice of experimental value. Its applicability to a wide range of initial reactivity and accuracy of estimation have been confirmed in the comparison to one-point kinetics numerical calculation. The expression suggests the first peak power increases with the square of small initial reactivity and three-halves power of large initial reactivity.
Yamane, Yuichi
Advances in Nuclear Fuel, p.159 - 174, 2012/02
The aim of the chapter is to introduce a concept of new method developed to evaluate the number of fission in a criticality accident, which is expected to give reasonable value, not too much overestimated, i.e. the estimated value is in the almost same order as the actual value. The 1st section introduces the phenomena of the criticality accident with uranyl nitrate solution based on the TRACY experiment, which has been conducted by Japan Atomic Energy Agency. In the 2nd section, the condition characterizing a criticality accident is considered, such as temperature, reactivity temperature coefficient, water, cooling, etc. In the 3rd section, a new simplified method to evaluate the total fission number is described. In the 4th section, the new developing method is applied to some case to see its applicability.
Murazaki, Minoru; Nobuhara, Fumiyoshi*; Iwai, Shohei*; Tonoike, Kotaro; Uchiyama, Gunzo
JAEA-Technology 2009-045, 46 Pages, 2009/09
JAEA-Technology-2009-045.pdf:12.88MB
Neutron doses under criticality accident conditions at TRACY were measured using ebnites, which are hard rubber containing sulfur. To evaluate a neutron dose, beta rays emitted from 
P induced by S(n,p) reaction are measured with Geiger-M
ller (GM) counter. A calibration factor (Gy/cpm), which is pre-determined using a 
Cf source, is applied to the count rates to obtain neutron doses. Factors to correct for the difference between responses of S(n,p) to the spectrum of Cf source and to spectra of TRACY were calculated and applied to the doses. Ebonites were exposed by TRACY with and without the water reflector. Neutron doses in TRACY without a reflector were evaluated with an uncertainty of less than about 40%. On the other hand, average of neutron doses in TRACY with the water reflector were accurate. By these measurements, it was found that ebonites can be used as a neutron dosimeter for criticality accidents.
Murazaki, Minoru; Tonoike, Kotaro; Uchiyama, Gunzo
JAEA-Technology 2009-022, 49 Pages, 2009/06
JAEA-Technology-2009-022.pdf:5.36MB
We have re-evaluated the dose measurements at SILENE with TLDs for neutrons and with TLDs for 
rays, and at TRACY with TLDs for neutrons. The measurements with TLDs for neutrons were re-evaluated by revising factors used for calculation of doses from measured data. The re-evaluated results of TLDs for neutrons at SILENE agreed with the reference value given by IRSN within 10%. The re-evaluated results of TLDs for neutrons at TRACY are consistent with dose and distance from the surface of the core tank. The re-evaluated results at TRACY were about 50% larger than results of polymer-alanine dosimeters and those of TLDs for neutrons measured by the authors. The measurements at SILENE with TLDs for rays were re-evaluated by revising the method for obtaining doses from measurement data. By the re-evaluation, it was confirmed that the methods described in the present report are valid for processing measured data of TLDs for neutrons and those for rays.
Murazaki, Minoru; Tonoike, Kotaro; Uchiyama, Gunzo
Journal of Nuclear Science and Technology, 46(2), p.193 - 203, 2009/02
Times Cited Count：1 Percentile：10.22(Nuclear Science & Technology)To develop a method for measuring neutron dose easily during criticality accidents, neutron ambient dose equivalents at the TRACY facility have been measured using neutron dose equivalent monitors with thermoluminescence dosimeters (TLD monitor). The TLD monitor is composed of two TLD badges and a cubical polyethylene case, and has a response similar to the ambient dose equivalent. In our experiments, TRACY was operated with and without water reflector to irradiate the TLD monitors. Measured ambient dose equivalents were proportional to the integrated power of TRACY, and agree well with calculation results of MCNP5. The measurement data were converted into tissue kerma using dose conversion factors calculated by MCNP5. Response correction factors to be applied to the measurement data considering the difference between responses of the TLD monitor to the Cf calibration source and to TRACY were also calculated by MCNP5. The neutron kerma ranged from 30 mGy to 15 Gy, which covers the range from 100 mGy to 10Gy specified as important in criticality accident dosimetry by the IAEA. The TLD monitor also satisfies the time limit on determination of doses required by the IAEA.
Ogawa, Kazuhiko; Kaminaga, Fumito*
Journal of Nuclear Science and Technology, 46(1), p.1 - 5, 2009/01
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In the safety research regarding a criticality accident in nuclear reprocessing plants, prediction of nuclear fission yields is essential to evaluate its influence on a radiation of neutrons/-rays to the public. For numerical calculations of the number of fissions in a criticality accident, a void reactivity coefficient due to the formation of radiolytic gas voids in the solution is required to evaluate the negative feedback reactivity. Small oscillations of both the power and the core pressure are measured after the first power pulse in a slow transient operation in TRACY. The small power oscillations are due to the reactivity feedback effect of the change of the volume of gas voids. In this paper, we propose a new estimation method of the void reactivity coefficient and estimate the coefficient using the TRACY transient data. The present estimated void reactivity coefficient is less than the calculated values in the uniform distribution of gas voids.
Ogawa, Kazuhiko; Kaminaga, Fumito*
Journal of Nuclear Science and Technology, 45(10), p.1028 - 1035, 2008/10
Times Cited Count：3 Percentile：23.55(Nuclear Science & Technology)no abstracts in English
Sono, Hiroki
JAEA-Review 2007-025, 141 Pages, 2007/06
JAEA-Review-2007-025.pdf:2.88MB
A new technique for effect assessment of radiation incidents was required for effective implementation of emergency measures in the early phase of unforeseen circumstances such as criticality accidents. In the present study, a new dosimetry technique for neutron and gamma-ray absorbed doses in human bodies was developed using two kinds of tissue-equivalent dosimeters: an alanine dosimeter and a lithium tetra borate dosimeter. The applications of this technique to personal dosimetry under criticality accident conditions and to retrospective assessment of criticality accident situations were also studied. The experiments and analyses using the Transient Experiment Critical Facility (TRACY) demonstrated that the dosimetry technique and its applications could estimate the magnitude and radiological consequence of criticality accidents precisely enough to plan emergency measures.
Sono, Hiroki; Ono, Akio; Kojima, Takuji; Yamane, Yoshihiro*
Journal of Nuclear Science and Technology, 44(1), p.43 - 53, 2007/01
Times Cited Count：2 Percentile：18.73(Nuclear Science & Technology)A practical method is proposed for retrospective estimation of the spatial dose distribution and the number of fissions in a typical criticality accident. In this method, two kinds of low-fading and tissue-equivalent dosimeters are used as area dosimeters: an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate. The procedure of the method consists of four parts: (1) dosimetry using area dosimeters (2) search for the center of the radiation source, (3) estimation of the spatial dose distribution, and (4) estimation of the number of fissions. Dosimetry of criticality accident situations simulated at the TRACY facility demonstrates the practicability of the method. Although a post-hoc analysis in principle, this method gives useful information on the magnitude and hazard level of the accident to determine the strategy of radiation emergency medicine and other post-accident measures if the area dosimeters are retrieved immediately after the accident.
Takahashi, Fumiaki; Endo, Akira; Yamaguchi, Yasuhiro; Oda, Keiji*
Hoken Butsuri, 41(3), p.180 - 187, 2006/09
no abstracts in English
Sono, Hiroki; Ono, Akio*; Kojima, Takuji; Takahashi, Fumiaki; Yamane, Yoshihiro*
Journal of Nuclear Science and Technology, 43(3), p.276 - 284, 2006/03
Times Cited Count：1 Percentile：9.98(Nuclear Science & Technology)For a study on the applicability of a personal dosimetry method to criticality accident dosimetry, an assessment of the human body surface and internal dose estimations was performed by experimental and computational simulations. The experimental simulation was carried out in a criticality accident situation at the TRACY facility. The neutron and -ray absorbed doses in muscle tissue were separately estimated by a dosimeter set of an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate with a phantom. The computational simulation was conducted with a Monte Carlo code taking account of dose components of neutrons, prompt -rays and delayed -rays. The computational simulation was ascertained to be valid by comparison between the calculated dose distributions in the phantom and the measured ones. The assessment based on the experimental and computational simulations confirmed that the personal dosimetry using the dosimeter set provided a first estimation of the body surface and internal doses with precision.
-ray absorbed doses inside human body in criticality accident situations using phantom and tissue-equivalent dosimetersSono, Hiroki; Kojima, Takuji; Soramasu, Noboru*; Takahashi, Fumiaki
JAERI-Conf 2005-007, p.315 - 320, 2005/08
Personal dosimeters provide a fundamental evaluation of external exposures to human bodies in radiation accidents. The dose distribution inside the body, which is needed to estimate the exposures from a result of personal dosimetry, has been evaluated mostly by computational simulations, while experimental data to verify the simulations are not sufficiently supplied, in particular, in criticality accident situations. For the purpose of obtaining the experimental data on external exposures inside the body, a preliminary experiment on criticality accident dosimetry was carried out at the Transient Experiment Critical Facility (TRACY) using a human phantom and tissue-equivalent dosimeters. The neutron and -ray absorbed doses inside the phantom could be satisfactorily measured by the combined use of an alanine dosimeter and a thermoluminescent dosimeter made of enriched lithium tetra borate. The doses measured in and on the phantom were regarded as reasonable in dose level and distribution by comparison with the doses measured in the free air.
Abe, Hitoshi; Tashiro, Shinsuke; Morita, Yasuji
JAERI-Conf 2005-007, p.199 - 204, 2005/08
Source term data for estimating release behavior of radioactive nuclides is necessary to evaluate synthetic safety of nuclear fuel cycle facility under accident conditions, such as fire and criticality. In JAERI, the data has been obtained by performing some demonstration tests. In this paper, the data for the criticality accident in fuel solution obtained from the TRACY experiment, will be mainly reviewed. At 4.5 h after the transient criticality, the release ratio of the iodine were about 0.2% for re-insertion of transient rod at just after transient criticality and about 0.9% for not re-insertion. Similarly the release coefficient and release ratio for Xe were estimated. It was proved that the release ratio of Xe-141 from the solution was over 90% in case that the inverse period was over about 100 (1/s). Furthermore, outline of the study on the fire accident as future plan will be also mentioned.
-ray dose components in criticality accident situationsSono, Hiroki; Yanagisawa, Hiroshi*; Ono, Akio*; Kojima, Takuji; Soramasu, Noboru*
Journal of Nuclear Science and Technology, 42(8), p.678 - 687, 2005/08
Times Cited Count：4 Percentile：30.51(Nuclear Science & Technology)Component analysis of -ray doses in criticality accident situations is indispensable for further understanding on emission behavior of -rays and accurate evaluation of external exposure to human bodies. Such dose components were evaluated, categorizing -rays into four components: prompt, delayed, pseudo components in the period of criticality, and a residual component in the period after the termination of criticality. This evaluation was performed by the combination of dosimetry experiments at the TRACY facility using a thermoluminescent dosimeter (TLD) made of lithium tetra borate and computational analyses using a Monte Carlo code. The evaluation confirmed that the dose proportions of the above components varied with the distance from the TRACY core tank. This variation was due to the difference in attenuation of the individual components with the distance from the core tank. The evaluated dose proportions quantitatively clarified the contribution of the pseudo and the residual components to be excluded for accurate evaluation of -ray exposure.
Takahashi, Fumiaki; Endo, Akira; Yamaguchi, Yasuhiro
Journal of Nuclear Science and Technology, 42(4), p.378 - 383, 2005/04
Times Cited Count：3 Percentile：24.22(Nuclear Science & Technology)Experiments were made to verify a dose assessment method from activated sodium in body in criticality accidents. A phantom containing sodium chloride solution was irradiated in the Transient Experiment Critical Facility to simulate activation of sodium. Monte Carlo calculations were performed to obtain quantitative relation between the activity of induced Na-24 and neutron dose in the phantom. In the previous work, conversion coefficients from specific activity of induced Na-24 to neutron dose had been analyzed with the MCNP-4B code concerning neutron spectra at some hypothesized configurations. One of the prepared coefficients was applied to evaluate neutron dose from the measured activity. The estimated dose agreed with the dose analyzed by the Monte Carlo calculation in the present study within an acceptable uncertainty, which is indicated by the IAEA. In addition, the dose calculated with the prepared coefficient was close to the result measured with dosimeters. These results suggest that the prepared coefficients can be applied to dose assessments from induced Na-24 in body.
Ad-hoc Committee for Investigation on Malfunction of Safety Rod in
JAERI-Review 2005-006, 60 Pages, 2005/03
JAERI-Review-2005-006.pdf:5.09MB
A trouble, malfunction of a safety rod, occurred during transient operation of the Transient Experiment Critical Facility (TRACY) in Japan Atomic Energy Research Institute (JAERI) on June 17, 2004. JAERI organized its specialists into the ad-hoc committee for investigation on malfunction of a safety rod of TRACY on June 23, 2004, to understand the cause of the trouble and take countermeasures to prevent the issue. The ad-hoc committee held 11 meetings and had discussions on the trouble inquiring the situations of the TRACY operation and the results of examination from the division relevant to the TRACY. As the result of investigation the cause of the trouble was attributed to the following reason: The holding capability of the safety rod was temporarily depressed by a small piece of polyethylene sheet put between the electromagnet and armature of the safety rod, the polyethylene sheet which had been used in overhaul activities of the rod. In the present report, the detailed results of investigation and the countermeasures to prevent the trouble are described.
