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
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.
Sono, Hiroki; Ono, Akio*; Kojima, Takuji; Takahashi, Fumiaki; Yamane, Yoshihiro*
Journal of Nuclear Science and Technology, 43(3), p.276 - 284, 2006/03
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.
Sono, 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.
Sono, Hiroki; Yanagisawa, Hiroshi*; Ono, Akio*; Kojima, Takuji; Soramasu, Noboru*
Journal of Nuclear Science and Technology, 42(8), p.678 - 687, 2005/08
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
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.
Kokusen, Junya; Seki, Masakazu; Abe, Masayuki; Nakazaki, Masato; Kida, Takashi; Umeda, Miki; Kihara, Takehiro; Sugikawa, Susumu
JAERI-Tech 2005-004, 53 Pages, 2005/03
This report presents operating records of dissolution of uranium dioxide and concentration of uranyl nitrate solution and acid removal, which have been performed from 1994 through 2003, for the purpose of feeding 10% and 6% enriched uranyl nitrate solution fuel to Static Experimental Critical Facility(STACY) and Transient Experimental Critical Facility(TRACY) in Nuclear Fuel Safety Engineering Facility(NUCEF).
Ad-hoc Committee for Investigation on Malfunction of Safety Rod in
JAERI-Review 2005-006, 60 Pages, 2005/03
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.
Sono, Hiroki; Tsukamoto, Michio; Aizawa, Eiju; Takeuchi, Masaki; Fukaya, Yuji; Iseda, Hirokatsu*; Ogawa, Kazuhiko; Sakuraba, Koichi; Tonoike, Kotaro
UTNL-R-0446, p.3_1 - 3_10, 2005/03
This report describes an investigation and countermeasures of the unexpected reactor shutdown of the TRACY facility at the Japan Atomic Energy Research Institute on June 17, 2004, caused by a malfunction of a safety rod of the facility. As a result of the investigation, the principal cause of the malfunction was ascribed to a drop in attraction of a binding magnet for the safety rod by a small piece of a foreign material. The following countermeasures were taken against its recurrence: prevention of incidence of foreign materials, reduction of the chance of intrusion of foreign materials into devices, and certain detection of foreign materials. This lesson learned through the experience is expected to be shared with the staffs concerned in reactor operation and maintenance.
Shimizu, Kaori; Gunji, Kazuhiko*; Haga, Takahisa*; Fukaya, Hiroyuki; Sonoda, Takashi; Sakazume, Yoshinori; Sakai, Yutaka*; Akutsu, Hideyuki; Niitsuma, Yasushi*; Inoue, Takeshi; et al.
JAERI-Tech 2004-078, 27 Pages, 2005/02
Analysis of the uranyl nitrate solution fuel are carried out at the analytical laboratory, NUCEF (the Nuclear Fuel Cycle Engineering Research Facility), which provide essential data for the operations of STACY (the Static Experiment Critical Facility), TRACY (the Transient Experiment Critical Facility) and the fuel treatment system.In the FY 2003, analysis of the uranyl nitrate solution fuel from STACY/TRACY on its pre- and post-operations, analysis of the uranyl nitrate solution under preparation stage for the fuel and analysis for nuclear material accountancy purpose, have been conducted. In addition, analysis on the third U/Pu extraction/separation tests among the preliminary tests to confirm adjustment condition of plutonium solution fuel for its further use at STACY from 2000, and analysis on the experiments to treat extraction waste, were conducted. A total number of analytical samples in the FY 2003 were 156.This report summarizes works related to the analysis and management of the analytical laboratory in the FY 2003.
Nakamura, Takemi*; Tonoike, Kotaro; Miyoshi, Yoshinori
Radiation Protection Dosimetry, 110(1-4), p.483 - 486, 2004/09
Dose measurement and evaluation technique in criticality accident conditions with a thermo luminescence dosimeter (TLD) was studied at the Transient Experiment Critical Facility (TRACY) of Japan Atomic Energy Research Institute (JAERI). The tissue absorbed dose can be derived from the ambient dose equivalent given by measurement with a TLD using the conversion factor given by computation. Using this technique, the neutron dose around the SILENE reactor of the Institute for Radioprotection and Nuclear Safety (IRSN) of France was measured in the Accident Dosimetry Intercomparison Exercise (June 10-21, 2002) organized by OECD/NEA and IRSN. In this exercise, the dose was also measured with a TLD. In this report, measurements and evaluation results at TRACY and SILENE are presented.
Sakai, Yutaka; Gunji, Kazuhiko; Haga, Takahisa*; Fukaya, Hiroyuki; Sonoda, Takashi; Sakazume, Yoshinori; Akutsu, Hideyuki; Niitsuma, Yasushi; Shirahashi, Koichi; Sato, Takeshi
JAERI-Tech 2004-006, 25 Pages, 2004/02
Analyses of the uranyl nitrate solution fuel are carried out at the analytical laboratory, NUCEF (the Nuclear Fuel Cycle Engineering Research Facility), which provide essential data for the operations of STACY (the Static Experiment Critical Facility), TRACY (the Transient Experiment Critical Facility) and the fuel treatment system. In the FY 2002, analyses of the uranyl nitrate solution fuel from STACY/TRACY on its pre- and post-operations, analyses of the uranyl nitrate solution under preparation stage for the fuel and analyses for nuclear material accountancy purpose, have been conducted. In addition, analyses on the preliminary tests to confirm adjustment condition of plutonium solution fuel for its further use at STACY, and analyses on the americium extraction/separation tests to provide americium for the research on high temperature chemistry of TRU, were conducted. A total number of analytical samples in the FY 2002 were 275. This report summarizes works related to the analyses and management of the analytical laboratory in the FY 2002.
Sono, Hiroki; Yanagisawa, Hiroshi*; Miyoshi, Yoshinori
JAERI-Tech 2003-096, 84 Pages, 2004/01
Prior to the supercritical experiments using a water-reflected core of the TRACY Facility, neutronic characteristics regarding criticality and reactivity of the core system were evaluated. In the analyses, a continuous energy Monte Carlo code, MVP, and a two-dimensional transport code, TWOTRAN, were used together with a nuclear data library, JENDL-3.3. By comparison to the characteristics in the former-used bare core system of TRACY, the water reflector was estimated not to change the kinetic parameter and to reduce the critical solution level by 20 %, the temperature coefficient of reactivity by 610 %, and the void coefficient of reactivity by 18 %, respectively. According to the Nordheim-Fuchs model, the first peak power during a power excursion was evaluated to be 15 % smaller than that in the bare system under the same conditions of fuel and reactivity insertion. The influence of the void feedback effect of reactivity, which is left out of consideration in the model, on the power characteristics will be evaluated from the results of the experiments.
Watanabe, Shoichi; Yamane, Yuichi; Miyoshi, Yoshinori
JAERI-Tech 2003-045, 73 Pages, 2003/03
Since exact information is not always acquired in the criticality accident of fuel-solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation which can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a three-dimensional nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating the thermal-hydraulics analysis code PHOENICS as a subroutine. The computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change.
Nakamura, Takemi*; Tonoike, Kotaro; Miyoshi, Yoshinori
JAERI-Tech 2003-031, 38 Pages, 2003/03
Institute for Radioprotection and Nuclear Safety (IRSN) and the OECD Nuclear Energy Agency (NEA) jointly organized SILENE Accident Dosimetry Intercomparison Exercise to intercompare the dose measurement systems of participating countries. Each participating country carried out dose measurements in the same irradiation field, and the measurement results were mutually compared. The authors participated in the exercise to measure the doses of gamma rays and neutron from SILENE by using thermoluminescence dosimeters (TLD's) and an alanine dosimeter. In this examination, the authors derived evaluation formulae for obtaining a tissue-absorbed dose from measured value (ambient dose equivalent) of TLD for neutron. We reported the tissue-absorbed dose computed using this evaluation formula to OECD/NEA. TLD's for neutron were irradiated in the TRACY facility to verify the evaluation formulae. The results of TLD's were compared with the calculations of MCNP and measurements with alanine dose meter. We found that the ratio of the dose by the evaluation formula to the measured value by the alanine dosimeter was 0.94 and the formula agreed within 6%. From examination of this TRACY, we can conclude that the value reported to OECD/NEA has equivalent accuracy.
Nakajima, Ken; Yanagisawa, Hiroshi; Miyoshi, Yoshinori
JAERI-Tech 2003-028, 31 Pages, 2003/03
We have tried to measure the power profile in the TRACY supercritical experiment with high accuracy by detecting epithermal neutrons. In order to measure the epithermal neutrons, a cadmium covered U fission chamber was used, and polyethylene, a neutron moderator, was set inside the cadmium to enhance the neutron detection efficiency. In addition, a lead shield was used to reduce the noise current due to gamma-rays. The measured results were compared with the ones using a thermal neutron detector, and it was found that the time delay effect in the thermal neutron detection, which was caused by the flight time of neutrons to reach the detector, distorted the power profile and reduced its peak value. The reduction ratio of peak power was about 4% for the relatively slow power change with the inserted reactivity of 1.5$, but it became over than 40% for the rapid power change with the reactivity of about 3$.
Nakajima, Ken; Yamamoto, Toshihiro; Miyoshi, Yoshinori
Journal of Nuclear Science and Technology, 39(11), p.1162 - 1168, 2002/11
A modified quasi-steady-state method has been developed to evaluate the mean power during a nuclear excursion in fissile solution. The original method used the critical equation based on the one-group theory to calculate the reactivity. However, the one-group approximation reduces the calculation accuracy and the geometrical buckling used in the critical equation is not applicable to complex geometries. Then, we have modified the method to use the feedback coefficients of reactivity. Although the modified method requires an external calculation to obtain the feedback coefficients, it is applicable to complex geometries and gives more accurate results than the one-group approximation. Moreover, a new method to calculate the boiling power has been developed. using the experimental data of a supercritical experiment facility, SILENE. Experimental analyses were conducted to validate the new method for supercritical xperiments using CRAC and TRACY. The results showed good agreements with the experiments.
Tsubata, Yasuhiro; Tashiro, Shinsuke; Koike, Tadao; Abe, Hitoshi*; Watanabe, Koji; Uchiyama, Gunzo
Transactions of the American Nuclear Society, 87, p.60 - 61, 2002/11
no abstracts in English
Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 8 Pages, 2002/10
The nuclear characteristics of TRACY, such as the criticality, the / ratio, the peak power, the energy of pulse, and the total energy, have been evaluated using the experimental data. TRACY is a supercritical reactor fueled with low-enriched uranyl nitrate aqueous solution to simulate criticality accidents in a fuel processing facility, such as a spent-fuel reprocessing plant. In this evaluation, the availability of criticality calculation and the models to evaluate the power and energy have been studied.