Study on release behavior of radioiodine from fuel solution under criticality accident condition

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.

Measurement of neutron and -ray absorbed doses inside human body in criticality accident situations using phantom and tissue-equivalent dosimeters

Sono, Hiroki; Kojima, Takuji; Soramasu, Noboru*; Takahashi, Fumiaki

JAERI-Conf 2005-007, p.315 - 320, 2005/08

https://jopss.jaea.go.jp/pdfdata/JAERI-Conf-2005-007.pdf

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.

Evaluation of -ray dose components in criticality accident situations

Sono, Hiroki; Yanagisawa, Hiroshi*; Ono, Akio*; Kojima, Takuji; Soramasu, Noboru*

Journal of Nuclear Science and Technology, 42(8), p.678 - 687, 2005/08

https://doi.org/10.3327/jnst.42.678

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.

Report on the fuel treatment facility operation

Kokusen, Junya; Seki, Masakazu; Abe, Masayuki; Nakazaki, Masato; Kida, Takashi; Umeda, Miki; Kihara, Takehiro; Sugikawa, Susumu

JAERI-Tech 2005-004, 53 Pages, 2005/03

JAERI-Tech-2005-004.pdf:5.92MB

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).

Report on investigation of unexpected reactor shutdown of the TRACY facility; Cause and countermeasures

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.

Dose evaluation in criticality accident conditions using transient critical facilities fueled with a fissile solution

Nakamura, Takemi*; Tonoike, Kotaro; Miyoshi, Yoshinori

Radiation Protection Dosimetry, 110(1-4), p.483 - 486, 2004/09

https://doi.org/10.1093/rpd/nch348

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.

Evaluation of neutronic characteristics of TRACY water-reflected core system

Sono, Hiroki; Yanagisawa, Hiroshi*; Miyoshi, Yoshinori

JAERI-Tech 2003-096, 84 Pages, 2004/01

JAERI-Tech-2003-096.pdf:3.6MB

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.

Analytical evaluation of dose measurement of critical accident at SILENE (Contract research)

Nakamura, Takemi*; Tonoike, Kotaro; Miyoshi, Yoshinori

JAERI-Tech 2003-031, 38 Pages, 2003/03

JAERI-Tech-2003-031.pdf:3.97MB

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.

Modified quasi-steady-state method to evaluate the mean power profiles of nuclear excursions in fissile solution

Nakajima, Ken; Yamamoto, Toshihiro; Miyoshi, Yoshinori

Journal of Nuclear Science and Technology, 39(11), p.1162 - 1168, 2002/11

https://doi.org/10.3327/jnst.39.1162

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.

Nuclear characteristics evaluation for a supercritical experiment facility using low-enriched uranium solution fuel, TRACY

Nakajima, Ken

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.