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Nakajima, Kunihisa; Nishioka, Shunichiro; Suzuki, Eriko; Osaka, Masahiko
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
Cesium chemisorption models were developed for estimation of amount of cesium chemisorbed onto stainless steel type 304 (SS304) during light water reactor severe accident. However, existing chemisorption models cannot accurately reproduce experimental results. In this study, a modified cesium chemisorption model was constructed based on a penetration theory for gas-liquid mass transfer with chemical reaction and was able to adequately describe effects on concentration of cesium hydroxide in gaseous phase and silicon content in SS304. It was found that the modified model can more accurately reproduce the experimental data than the existing model.
Yamano, Hidemasa; Vasile, A.*; Kang, S.-H.*; Summer, T.*; Tsige-Tamirat, H.*; Wang, J.*; Ashurko, I.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 "Methods, Models and Codes", WP-SO-2 "Experimental Programs and Operational Experience", and WP-SO-3 "Studies of Innovative Design and Safety Systems". This paper reports recent activities within the SO project.
Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Kanazawa, Toru*; Nakashima, Kazuo*; Tojo, Masayuki*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
Oxidation behaviour of Zr cladding in SFP accident condition was evaluated by using a thermobalance in this work, and the obtained data were applied to construct oxidation model for SFP accident condition. For the validation of the constructed oxidation model, oxidation tests using a long cladding tube 500mm in length were conducted in conditions simulating SFP accidents, such as flow rate of the atmosphere in spent fuel rack, temperature gradient along the axis of cladding, and heating-up history. Thickness of oxide layer formed on the surface of cladding samples was evaluated by cross sectional observation, and compared with calculation results obtained by using the oxidation model. The detail of experimental results and validation of the oxidation model will be discussed.
Tojo, Masayuki*; Kanazawa, Toru*; Nakashima, Kazuo*; Iwamoto, Tatsuya*; Kobayashi, Kensuke*; Goto, Daisuke*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05
In this study, fuel loading effects in BWR spent fuel rack accidents are widely investigated using three-dimensional analysis methods from both nuclear and thermal hydraulics viewpoints, including: (a) Decay heat of spent fuel after discharge, (b) The maximum temperature of spent fuel cladding in the spent fuel rack depending on heat transfer phenomena, and (c) Criticality of the spent fuel rack after collapsing of the fuel due to a severe accidents in the BWR spent fuel pool (SFP).
Mano, Akihiro; Katsuyama, Jinya; Li, Y.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
In Japanese nuclear power plants, non-destructive examinations (NDEs) are performed for welds in piping in accordance with the rules such as Rules on Fitness-for-Service for Nuclear Power Plant of the Japan Society of Mechanical Engineers (JSME FFS). A set of NDEs is performed in each 10-year interval, and the extent of examination in each interval is specified in the rules. Welding lines to be examined are selected considering the extent of examination based on two sampling methods. One is the fixed location sampling method that welds to be examined are selected from welds examined in the last interval. The other is the random location sampling method that welds to be examined are selected from other than welds examined in the last interval. The selection of the sampling methods is considered to be one of the important factors in in-service inspection. Probabilistic fracture mechanics (PFM) analysis is expected to be more rational method for the structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In this study, we investigated the influence of the sampling methods related to the NDE on failure probability of typical nuclear piping based on PFM analyses. Through sensitivity PFM analyses, we confirmed that failure probability value obtained from PFM analysis is useful as a quantitative numerical index for selecting the sampling method in an in-service inspection.
Riyana, E. S.; Okumura, Keisuke; Terashima, Kenichi
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05
Negyesi, M.; Amaya, Masaki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
Pshenichnikov, A.; Yamazaki, Saishun; Nagae, Yuji; Kurata, Masaki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
Onodera, Naoyuki; Idomura, Yasuhiro; Kawamura, Takuma; Uesawa, Shinichiro; Yamashita, Susumu; Yoshida, Hiroyuki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05
A dry method is one of practical methods for decommissioning the TEPCO's Fukushima Daiichi Nuclear Power Station. Japan Atomic Energy Agency (JAEA) has been evaluating the air cooling performance by using the JUPITER code. However, the JUPITER code requires a large computational cost to capture debris' structures. To accelerate such CFD analyses, we use the CityLBM code, which is based on the lattice Boltzmann method (LBM) and is highly optimized for GPUs. The CityLBM code is validated against free convective heat transfer experiments at JAEA, and the similar accuracy as the JUPITER code is confirmed regarding the prediction capability of heat transfer and the resulting temperature distributions. It is also shown that the elapse time of a CityLBM simulation on GPUs is reduced to 1/6 compared with that of the corresponding JUPITER simulation on CPUs with the same number of GPUs and CPUs. The results show that the LBM is promising for accelerating thermal convective simulations.
Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.
Ishida, Shinya; Kawada, Kenichi; Fukano, Yoshitaka
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 10 Pages, 2019/05
Core Disruptive Accident (CDA) has been considered as one of the important safety issues in the severe accident evaluation of Sodium-cooled Fast Reactor (SFR), and SAS4A code is developed for Initiating Phase (IP) of CDA. Phenomena Identification and Ranking Table (PIRT) approach was applied to the validation of SAS4A code in order to enhance its reliability in this study. SAS4A was validated in the following steps: (1) selection of the figure of merit (FOM) corresponding to Unprotected Loss Of Flow (ULOF) which is one of the most important and typical events in CDA, (2) identification of the phenomena involved in ULOF, (3) ranking the important phenomena, (4) development of the code validation test matrix, and (5) test analyses for validation corresponding to the test matrix. The reliability and validity of SAS4A code were significantly enhanced by this validation with PIRT approach.
Tada, Kenichi
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05
The decommissioning of Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. The criticality safety of fuel debris is imperative to prevent exposure of workers. The investigating criticality monitoring system cannot detect the criticality of fuel debris quickly. The estimation of criticality of fuel debris is required for the fuel debris retrieval. Though the expert knowledge of reactor physics is necessary to estimate the criticality of fuel debris, many people who make a plan of fuel debris retrieval may not know well about criticality analysis. We developed a handy criticality analysis tool HAND to quickly estimate the criticality of fuel debris without expert knowledge of reactor physics. Since the input data of HAND is so simple and users can intuitively understand the calculation results, this tool is expected to be the effective tool to estimate the criticality of fuel debris.
Katano, Ryota
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05
no abstracts in English
Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Maruyama, Yu; Kaneko, Akiko*; Abe, Yutaka*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05
Horiguchi, Naoki; Miyahara, Naoya; Uesawa, Shinichiro; Yoshida, Hiroyuki; Osaka, Masahiko
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
For source term evaluation from reactor buildings (RBs) in LWR severe accidents, we have launched to develop an evaluation method of FP aerosol particle deposition onto surfaces of internal structures in an RB based on computational fluid dynamics (CFD). This paper describes development of a CFD simulation tool as the base part of the evaluation method. A preliminary simulation for a representative RB under a representative flow condition was conducted to confirm the tool performance by roughly grasping the deposition behaviors of FP aerosol particle and decontamination factor (DF) in the RB. Calculation results showed that most of aerosol particles were deposited along with gas flow formed by the internal structures in the RB, demonstrating the advantageous feature of the present CFD tool. The DFs from 4 to 14 were obtained with increase of the particle diameters from 0.1 to 10 
m as expected in terms of the particle movement equation.
Kang, Z.; Nagai, Minoru*; Nishida, Akemi; Tsubota, Haruji; Li, Y.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
Many empirical formulae have been proposed for evaluating the local damage to reinforced concrete (RC) structures caused by rigid projectile impact. The majority of these formulae aim at impact tests perpendicular to target structures, while few impact tests oblique to the target structure have been studied. The final objective of this study is to propose a new formula for evaluating the local damage to RC structures caused by oblique impact based on past experimental and simulation results. The finite element code LS-DYNA R7.1.2 is used to perform the numerical analysis by adopting Lagrangian finite elements and explicit time integration. So far, we validated the analytical method by comparison with the experimental results and conduct the simulation analysis of impact assessment on RC panel by rigid/soft projectile with flat nose shape using the validated approach. Results of reduction coefficient with respect to rigid/soft projectile and impact angle were obtained. Therefore, in this study, we focus on the impact problems caused by rigid projectile with hemispherical nose shape. The same analytical method is used to simulate the local damage to RC panels caused by oblique impact of rigid projectile with hemispherical nose shape. The results associated with penetration depth of RC structure, energy contribution ratio, etc. are presented. According to the comparison analysis of results of local damage to RC structure by rigid projectiles with flat and hemispherical nose shape, the influence of different nose shapes of rigid projectile on the local damage of RC panels caused by oblique impact is investigated.
Lu, K.; Katsuyama, Jinya; Li, Y.; Miyamoto, Yuhei*; Hirota, Takatoshi*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
Nagai, Minoru*; Kang, Z.; Nishida, Akemi; Tsubota, Haruji; Li, Y.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
In this study, the final purpose is to propose a new formula for evaluating the local damage caused by oblique impact based on past experimental results and previous research achievements. Up to now, we validated the analytical method by comparison with the experimental results and conducted simulation analysis associated with impact assessment on RC panel by soft/rigid projectile with flat nose shape using the validated approach. In the part 1 of this study, the same procedure of our previous work is followed to investigate the local damage to RC panel caused by rigid projectile with flat and hemispherical nose shape. In the part 2, we focus on the comparison analysis of simulation results of local damage to RC panel subjected to oblique impact by soft missile with flat and hemispherical nose shape. The structural damage of RC panel and projectiles, energy contribution ratio, etc. is studied for each case. The results indicate the difference of nose shape of projectile is of great importance to influence the penetration depth generated by oblique impact of soft projectile.
Nishimura, Satoshi*; Satake, Masaaki*; Nishi, Yoshihisa*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
In this study, accident progression analyses in the SFP were performed to investigate cooling effects of the SFP spray and an alternate water injection in the loss-of-pool water accident with MAAP ver. 5.05 beta. Fuel cladding oxidation model which was created by JAEA based on their experimental data was selected and applied in the present calculations. In case of an assessment of SFP spray effects, decay heat, spray fraction going into the fuel assembly, spray droplet diameter, spray start time were selected as analytical parameters. When the SFP spray of 12.5 kg/s (200 GPM) starts 4 hours after the onset of the accident against the spent fuels with 4 months cooling and if the spray fraction going into the fuel assembly is greater than 30%, the maximum cladding temperature can be maintained under 727
C (1000 K), resulting in avoiding the cladding failure.
Miwa, Shuhei; Miyahara, Naoya; Nakajima, Kunihisa; Nishioka, Shunichiro; Suzuki, Eriko; Horiguchi, Naoki; Liu, J.; Miradji, F.; Imoto, Jumpei; Afiqa, B. M.; et al.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
We constructed the first version of fission product (FP) chemistry database named ECUME for LWR severe accident. The first version of ECUME is equipped with dataset of the chemical reactions and their kinetics constants for the reactions of cesium(Cs)-iodine(I)-boron(B)-molybdenum(Mo)-oxygen(O)-hydrogen(H) system in gas phase, the elemental model for the high temperature chemical reaction of Cs with stainless steel, and thermodynamic data for CsBO
vapor species and solids of CsSi
O
and CsFeSiO. The ECUME will provide more accurate estimation of Cs distribution due to the evaluation of effects of interaction with BWR control material B and stainless steel on Cs behavior in the Fukushima Daiichi Nuclear Power Station.
