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Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi
JAEA-Research 2024-007, 24 Pages, 2024/08
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. It has been observed experimentally that volatility of RuO is suppressed by HNO generated by HNO radiolysis. The analysis of chemical reactions of NO including HNO and HNO in the waste tank is essential to simulate of these phenomena. To resolve this issue, an analytical approach has been attempted to couple dynamically two computer codes SHAWED and SCHERN. The simulation of boiling behavior in the tank is conducted with SHAWED. SCHERN simulates chemical behaviors of HNO, HNO and NO in the tank. A programmatic coupling algorithm and a trial simulation of the accident are presented in this report.
Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Nuclear Technology, 210(5), p.814 - 835, 2024/05
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.
Fukuda, Kodai; Obara, Toru*; Suyama, Kenya
Nuclear Technology, 11 Pages, 2024/00
Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2023-001, 26 Pages, 2023/05
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an analytical approach has been developed using computer simulation programs to assess the radioactive source term from those facilities. The proposed approach consists analyses with three computer programs. At first, the simulation of boiling behavior in the HLLW tank is conducted with SHAWED code. Next step, the thermal-hydraulic behavior in the facility building is simulated with MELCOR code based on the results at the first step simulation such as flowed out mixed steam flow rate, temperature and volatilized Ru from the tank. The final analysis step is carried out for estimating amount of released radioactive materials with SCHERN computer code which simulates chemical behaviors of nitric acid, nitrogen oxide and Ru based on the condition also simulated MELCOR. Series of sample simulations of the accident at a hypothetical typical facility are presented with the data transfer between those codes in this report.
Hamdani, A.; Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke
Progress in Nuclear Energy, 153, p.104415_1 - 104415_16, 2022/11
Times Cited Count:3 Percentile:63.91(Nuclear Science & Technology)Nakamura, Hideo; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
Ono, Hirokazu; Ishii, Eiichi
Geomechanics for Energy and the Environment, 31, p.100317_1 - 100317_9, 2022/09
Times Cited Count:5 Percentile:63.89(Energy & Fuels)Yamamura, Sota*; Fujiwara, Kota*; Honda, Kota*; Yoshida, Hiroyuki; Horiguchi, Naoki; Kaneko, Akiko*; Abe, Yutaka*
Physics of Fluids, 34(8), p.082110_1 - 082110_13, 2022/08
Times Cited Count:2 Percentile:35.53(Mechanics)Liquid spreading and atomization due to jet impingement in liquid-liquid systems are considered to be crucial for understanding the cooling behavior of high-temperature molten material in a shallow water pool. This phenomenon takes place when a liquid jet enters a pool filled with other immiscible liquid. The jet spreads radially after impinging on the floor while forming a thin liquid film and atomizing droplets. In this paper, we explain the result to quantify the unsteady three-dimensional behavior of the spreading jet by the employment of 3D-LIF measurements and 3-dimensional reconstruction. Under high flow velocity conditions, the phenomena of hydraulic jump and atomization of the liquid film occurred along with the spreading. To evaluate the spreading behavior, a comparison of the jump radius position of the liquid-liquid system as the representative value was made with the one calculated by the existing theory of a gas-liquid system. As the result, the spreading of the liquid film in the liquid-liquid system was suppressed compared with that in the gas-liquid system. Furthermore, the PTV method was successfully used to measure the velocity boundary layer and velocity profile in the liquid film, which are important factors that affect the spreading mechanism of the liquid film. These results revealed that in liquid-liquid systems, shear stress at the liquid-liquid interface causes a decrease in the flow velocity and suppressed the development of the velocity boundary layer. Also, to evaluate the atomization behavior, the number and diameter distribution of the droplets were measured from the acquired 3-dimensional shape data of the jet. As the result, the number of droplets increased with the flow velocity. Based on these results, we concluded that the spreading of the liquid film is affected by such atomization behavior.
Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2021-013, 20 Pages, 2022/01
An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. An idea has been proposed to implement a steam condenser as an accident countermeasure. This measure is expected to prevent nitric acid steam diffusing in facility building and to increase gaseous Ru trapping ratio into condensed water. A simulation study has been carried out with a hypothetical typical facility building to analyze the efficiency of steam condenser. In this study, SCHERN computer code simulates chemical behaviors of Ru in nitrogen oxide, nitric acid and water mixed vapor based on the conditions obtained from simulation with thermal-hydraulic computer code MELCOR. The effectiveness of steam condenser has been analyzed quantitively in preventing mixed vapor diffusion and gaseous Ru trapping effect. Some issues to be solved in analytical model has been also clarified in this study.
Ina, Takuya*; Idomura, Yasuhiro; Imamura, Toshiyuki*; Yamashita, Susumu; Onodera, Naoyuki
Proceedings of 12th Workshop on Latest Advances in Scalable Algorithms for Large-Scale Systems ScalA21) (Internet), 8 Pages, 2021/11
Times Cited Count:1 Percentile:47.71(Computer Science, Software Engineering)A new mixed-precision preconditioner based on the iterative refinement (IR) method is developed for preconditioned conjugate gradient (P-CG) and multigrid preconditioned conjugate gradient (MGCG) solvers in a multi-phase thermal-hydraulic CFD code JUPITER. In the IR preconditioner, all data is stored in FP16 to reduce memory access, while all computation is performed in FP32. The hybrid FP16/32 implementation keeps the similar convergence property as FP32, while the computational performance is close to FP16. The developed solvers are optimized on Fugaku (A64FX), and applied to ill-conditioned matrices in JUPITER. The P-CG and MGCG solvers with the new IR preconditioner show excellent strong scaling up to 8,000 nodes, and at 8,000 nodes, they are respectively accelerated up to 4.86 and 2.39 from the conventional ones on Oakforest-PACS (KNL).
Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2021-005, 25 Pages, 2021/08
An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an empirical correlation equation of Ru mass transfer coefficient across the vapor-liquid surface, which can be useful for quantitative simulation of Ru mitigating behavior, has been obtained from data analyses of small-scale experiments conducted to clarify gaseous Ru migrating behavior under steam-condensing condition. A simulation study has been also carried out with a hypothetical typical facility building successfully to demonstrate the feasibility of quantitative estimation of amount of Ru migrating in the facility using the obtained correlation equation implemented in SCHERN computer code which simulates chemical behaviors of nitrogen oxide based on the condition also simulated thermal-hydraulic computer code.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Times Cited Count:3 Percentile:34.17(Nuclear Science & Technology)Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
Maruyama, Yu; Yoshida, Kazuo
Nihon Genshiryoku Gakkai-Shi ATOMO, 63(7), p.517 - 522, 2021/07
no abstracts in English
Wang, Z.; Duan, G.*; Koshizuka, Seiichi*; Yamaji, Akifumi*
Nuclear Power Plant Design and Analysis Codes, p.439 - 461, 2021/00
Ono, Hirokazu; Takeda, Masaki; Ishii, Eiichi
Extended abstract of International Conference on Coupled Processes in Fractured Geological Media; Observation, Modeling, and Application (CouFrac 2020) (Internet), 4 Pages, 2020/11
Ozaki, Yusuke; Ishii, Eiichi; Sugawara, Kentaro*
Extended abstract of International Conference on Coupled Processes in Fractured Geological Media; Observation, Modeling, and Application (CouFrac 2020) (Internet), 4 Pages, 2020/11
We perform the numerical simulation of the response of hydraulic head observed in HDB-6 during the excavation of the Horonobe URL to verify the existence of low effective permeable domain in the subsurface. The low permeable domain as an intact rock due to the low hydraulic fracture connectivity is estimated to exist in the deep domain while the permeability of the shallow domain is relatively high due to the hydraulic fracture connectivity there. Our simulation shows that the observed hydraulic head is affected by the Mandel-Cryer effect due to the hydrogeological structure and the effect for the duration of over years requires the low permeability as an intact rock in the deep domain. These results verify the existence of the low effective permeable domain in the deep subsurface estimated by the previous study.
Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo
Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09
Times Cited Count:4 Percentile:20.53(Nuclear Science & Technology)Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki
Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06
Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.
Hamdani, A.; Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.5463 - 5479, 2019/08
Katsuyama, Jinya; Masaki, Koichi; Lu, K.; Watanabe, Tadashi*; Li, Y.
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 7 Pages, 2019/07
For reactor pressure vessel (RPV) of pressurized water reactor, temperature of coolant water in emergency core cooling system (ECCS) may have influence on the structural integrity of RPV during pressurized thermal shock (PTS) events. Focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs in order to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. From the analysis results, it was shown that the failure probability of RPV was dramatically reduced when the coolant temperature in accumulator as well as high and low pressure injection systems (HPI/LPI) was raised, although raising the coolant temperature of HPI/LPI only did not cause reduction in the failure probability.