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Yamashita, Kiyoto; Maki, Shota; Yokosuka, Kazuhiro; Fukui, Masahiro; Iemura, Keisuke
JAEA-Technology 2023-023, 97 Pages, 2024/03
JAEA-Technology-2023-023.pdf:8.21MB
The incinerator adopted to incineration room, Plutonium Waste Treatment Facility had been demonstrated since 2002 for developing technologies to reduce the volume of fire-resistant wastes such as vinyl chloride (represented by Polyvinyl chloride bags) and rubber gloves for Radio Isotope among radioactive solid wastes generated by the production of mixed oxide fuels. The incinerator, cooling tower, and processing pipes were replaced with a suspension period from 2018 to 2022, which fireproof materials on the inner wall of the incinerator was cracked and grown caused by hydrogen chloride generated when disposing of fire-resistant wastes. This facility consists of the waste feed process, the incineration process, the waste gas treatment process, and the ash removal process. We replaced the cooling tower in the waste gas treatment process from March 2020 to March 2021, and the incinerator in the incineration process from January 2021 to February 2022. In addition, samples were collected from the incinerator and the cooling tower during the removing and dismantling of the replaced devices, observed by Scanning Electron Microscope and X-ray microanalyzer, and analyzed by X-ray diffraction to investigate the corrosion and deterioration of them. This report describes the method of setting up the green house, the procedure for replacing them, and the results from analysis in corrosion and deterioration of the cooling tower and incinerator.
Takeda, Takeshi
JAEA-Data/Code 2023-007, 72 Pages, 2023/07
JAEA-Data-Code-2023-007.pdf:3.24MB
An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment IB-HL-01.
Li, C.-Y.; Wang, K.*; Uchibori, Akihiro; Okano, Yasushi; Pellegrini, M.*; Erkan, N.*; Takata, Takashi*; Okamoto, Koji*
Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07
Times Cited Count：1 Percentile：56.82(Chemistry, Multidisciplinary)Ishitsuka, Etsuo; Ho, H. Q.; Kitagawa, Kanta*; Fukuda, Takahito*; Ito, Ryo*; Nemoto, Masaya*; Kusunoki, Hayato*; Nomura, Takuro*; Nagase, Sota*; Hashimoto, Haruki*; et al.
JAEA-Technology 2023-013, 19 Pages, 2023/06
JAEA-Technology-2023-013.pdf:1.75MB
Eight people from five universities participated in the 2022 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the feasibility study for nuclear battery, the burn-up analysis of HTTR core, the feasibility study for 
Cf production, the analysis of behavior on loss of forced cooling test, and the thermal-hydraulic analysis near reactor pressure vessel. In the questionnaire after this training, there were impressions such as that it was useful as a work experience, that some students found it useful for their own research, and that discussion with other university students was a good experience. These impressions suggest that this training was generally evaluated as good.
Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
For contaminated water management in decommissioning Fukushima Daiichi Nuclear Power Stations, reduction in water injection, intermittent injection water and air cooling are considered. However, since there are uncertainties of fuel debris in the PCV, it is necessary to examine and evaluate optimal cooling methods according to the distribution state of the fuel debris and the progress of the fuel debris retrieval work in advance. We have developed a method for estimating the thermal behavior in the air cooling, including the influence of the position, heat generation and the porosity of fuel debris. Since a large-scale thermal-hydraulics analysis of natural convection is necessary for the method, JUPITER developed independently by JAEA is used. It is however difficult to perform the large-scale thermal-hydraulics analysis with JUPITER by modeling the internal structure of the debris which may consist of a porous medium. Therefore, it is possible to analyze the heat transfer of the porous medium by adding porous models to JUPITER. In this study, we report the validation of JUPITER applied the porous model and discuss which heat transfer models are most effective in porous models such as series, parallel and geometric mean models. To obtain validation data of JUPITER for the natural convective heat transfer analysis around the porous medium, we performed the heat transfer and the flow visualization experiments of the natural convection in the experimental system including the porous medium. In the comparison between the experiment and the numerical analysis with each model, the numerical result with the geometric mean model was the closest of the models to the experimental results. However, the numerical results of the temperature and the velocity were overestimated for those experimental results. In particular, the temperature near the interface between the porous medium and air was more overestimated.
Li, F.; Narukawa, Takafumi; Udagawa, Yutaka
Journal of Nuclear Science and Technology, 12 Pages, 2023/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
Sato, Ikken
Nuclear Engineering and Design, 383, p.111426_1 - 111426_19, 2021/11
Times Cited Count：5 Percentile：64.12(Nuclear Science & Technology)Matsumoto, Toshinori; Iwasawa, Yuzuru; Sugiyama, Tomoyuki
Proceedings of Reactor core and Containment Cooling Systems, Long-term management and reliability (RCCS 2021) (Internet), 8 Pages, 2021/10
A methodological framework is being developed in JAEA for evaluating debris coolability at ex-vessel during the severe accident (SA) of BWR under the wet cavity strategy. The probability of ex-vessel debris coolability under the wet cavity strategy is analyzed to demonstrate the evaluation approach. Probabilistic distribution of the melt conditions ejected from the RPV was obtained as the result of the iterative analyses with MELCOR code. Five uncertainty parameters relating with the core degradation and transfer process were chosen. Parameter sets were generated by Latin hypercube sampling (LHS). JASMINE code plays the physical model to predict the mass fraction of agglomerated debris and melt pool spreading on the floor. Fifty-nine input parameter set for JASMINE code were generated by LHS again using the probabilistic distribution of melt condition determined from the results of MELCOR analyses. The depth of the water pool was set as 0.5, 1.0 and 2.0 m. The accumulated debris height was compared with the criterion to judge the debris coolability. As the result, the success probability of debris cooling was obtained through the sequence of calculations.
Sato, Ikken; Arai, Yuta*; Yoshikawa, Shinji
Journal of Nuclear Science and Technology, 58(4), p.434 - 460, 2021/04
Times Cited Count：6 Percentile：70.8(Nuclear Science & Technology)Takeda, Takeshi
JAEA-Data/Code 2020-019, 58 Pages, 2021/01
JAEA-Data-Code-2020-019.pdf:3.85MB
An experiment denoted as SB-SL-01 was conducted on March 27, 1990 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment SB-SL-01 simulated a main steam line break (MSLB) accident in a pressurized water reactor (PWR). The test assumptions were made such as auxiliary feedwater (AFW) injection into secondary-side of both steam generators (SGs) and coolant injection from high pressure injection (HPI) system of emergency core cooling system into cold legs in both loops. The MSLB led to a fast depressurization of broken SG, which caused a decrease in the broken SG secondary-side wide-range liquid level. The broken SG secondary-side wide-range liquid level recovered because of the AFW injection into the broken SG secondary-side. The primary pressure temporarily decreased a little just after the MSLB, and increased up to 16.1 MPa following the closure of the SG main steam isolation valves. Coolant was manually injected from the HPI system into cold legs in both loops a few minutes after the primary pressure reduced to below 10 MPa. The primary pressure raised due to the HPI coolant injection, but was kept at less than 16.2 MPa by fully opening a power-operated relief valve of pressurizer. The core was filled with subcooled liquid through the experiment. Thermal stratification was seen in intact loop cold leg during the HPI coolant injection owing to the flow stagnation. On the other hand, significant natural circulation prevailed in broken loop. When the continuous core cooling was ensured by the successive coolant injection from the HPI system, the experiment was terminated. The experimental data obtained would be useful to consider recovery actions and procedures in the multiple fault accident with the MSLB of PWR. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-SL-01.
Oto, Tsutomu; Asano, Norikazu; Kawamata, Takanori; Yanai, Tomohiro; Nishimura, Arashi; Araki, Daisuke; Otsuka, Kaoru; Takabe, Yugo; Otsuka, Noriaki; Kojima, Keidai; et al.
JAEA-Review 2020-018, 66 Pages, 2020/11
JAEA-Review-2020-018.pdf:8.87MB
A collapse event of the cooling tower of secondary cooling system in the JMTR (Japan Materials Testing Reactor) was caused by the strong wind of Typhoon No.15 on September 9, 2019. The cause of the collapse of the cooling tower was investigated and analyzed. As the result, it was identified that four causes occurred in combination. Thus, the soundness of the cooling tower of Utility Cooling Loop (UCL cooling tower), which is a wooden cooling tower installed at the same period as the cooling tower of secondary cooling system, was investigated. The items of soundness survey are to grasp the operation conditions of the UCL cooling tower, to confirm the degradation of structural materials, the inspection items and inspection status of the UCL cooling tower, and to investigate the past meteorological data. As the results of soundness survey of the UCL cooling tower, the improvement of inspection items of the UCL cooling tower was carried out and the replacement and repair of the structural materials of the UCL cooling tower were planned for safe maintenance and management of this facility. And the renewal plan of new cooling tower was created to replace the existing UCL cooling tower. This report is summarized the soundness survey of the UCL cooling tower.
Matsumoto, Toshinori; Iwasawa, Yuzuru; Ajima, Kohei*; Sugiyama, Tomoyuki
Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 10 Pages, 2020/11
The probability of ex-vessel debris coolability under the wet cavity strategy is analyzed. The first step is the uncertainty analyses by severe accident analysis code MELCOR to obtain the melt condition. Five uncertain parameters which are relating with the core degradation and transfer process were chosen. Input parameter sets were generated by LHS. The analyses were conducted and the conditions of the melt were obtained. The second step is the analyses for the behavior of melt under the water by JASMINE code. The probabilistic distribution of parameters are determined from the results of MELCOR analyses. Fifty-nine parameter sets were generated by LHS. The depth of water pool is set to be 0.5, 1.0 and 2.0 m. Debris height were compared with the criterion to judge the debris coolability. As the result, the success probability of debris cooling was obtained through the sequence of calculations. The technical difficulties of this evaluation method are also discussed.
Narukawa, Takafumi; Amaya, Masaki
Journal of Nuclear Science and Technology, 57(7), p.782 - 791, 2020/07
Times Cited Count：6 Percentile：59.94(Nuclear Science & Technology)Hotta, Akitoshi*; Akiba, Miyuki*; Morita, Akinobu*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Sk
ld, P.*; et al.
Journal of Nuclear Science and Technology, 57(4), p.353 - 369, 2020/04
Times Cited Count：14 Percentile：71.27(Nuclear Science & Technology)Ishitsuka, Etsuo; Matsunaka, Kazuaki*; Ishida, Hiroki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Kondo, Atsushi*; et al.
JAEA-Technology 2019-008, 12 Pages, 2019/07
JAEA-Technology-2019-008.pdf:2.37MB
As a summer holiday practical training 2018, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out. As a result, it is become clear that the continuous operations for about 30 years at 2 MW, about 25 years at 3 MW, about 18 years at 4 MW, about 15 years at 5 MW are possible. As an image of thermal design, the image of the nuclear battery consisting a cooling system with natural convection and a power generation system with no moving equipment is proposed. Further feasibility study to confirm the feasibility of nuclear battery will be carried out in training of next fiscal year.
Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.
Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Tsujimoto, Kazufumi
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
In order to investigate the basic neutronics characteristics of the accelerator-driven subcritical system (ADS), JAEA has a plan to construct a new critical assembly in the J-PARC project, Transmutation Physics Experimental Facility (TEF-P). This study aims to evaluate the natural cooling characteristics of TEF-P core which has large decay heat by minor actinide (MA) fuel, and to achieve a design that does not damage the core and the fuels during the failure of the core cooling system. In the evaluation of the TEF-P core temperature, empty rectangular lattice tube outer of the core has a significant effect on the heat transfer characteristics. The experiments by using the mockup device were performed to validate the heat transfer coefficient and experimental results were obtained. By using the obtained experimental results, the three-dimensional heat transfer analysis of TEF-P core were performed, and the maximum core temperature was obtained, 294
C. This result shows TEF-P core temperature would be less than 327C that the design criterion of temperature.
Takeda, Takeshi
JAEA-Data/Code 2018-004, 64 Pages, 2018/03
JAEA-Data-Code-2018-004.pdf:3.33MB
Experiment SB-SG-10 was conducted on November 17, 1992 using LSTF. Experiment simulated recovery actions from multiple steam generator (SG) tube rupture accident in PWR. Primary pressure was kept higher than broken SG secondary-side pressure due to coolant injection from high pressure injection (HPI) system into cold and hot legs even after start of full opening of intact SG relief valve (RV). Full opening of power-operated relief valve (PORV) in pressurizer (PZR) resulted in pressure equalization between primary and broken SG systems as well as PZR liquid level recovery. Broken SG RV opened once after start of intact SG RV full opening. Core was filled with saturated or subcooled liquid through experiment. Significant natural circulation prevailed in intact loop after start of intact SG RV full opening. Significant thermal stratification appeared in hot legs especially during time period of HPI coolant injection into hot legs.
Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Takada, Shoji
Journal of Nuclear Engineering and Radiation Science, 3(4), p.041013_1 - 041013_8, 2017/10
In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.
