Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Rizaal, M.; Saito, Takumi*; Okamoto, Koji*; Erkan, N.*; Nakajima, Kunihisa; Osaka, Masahiko
Mechanical Engineering Journal (Internet), 7(3), p.19-00563_1 - 19-00563_10, 2020/06
The adsorption of cesium (Cs) on calcium silicate insulation of primary piping system is postulated to contribute in high dose rate of surrounding pedestal area in Fukushima Daiichi NPP unit 2. In this study, room-temperature experiment of Cs adsorption on calcium silicate has been studied as an initial approach of Cs adsorption behavior toward higher temperature condition. As the result of analyzing of Cs adsorption kinetics, it was expected that the underlying adsorption mechanism is chemisorption. Furthermore, analysis of adsorption isotherm suggested unrestricted monolayer formation followed by multilayer formation.
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.
Shimada, Asako; Nemoto, Hiromi*; Sawaguchi, Takuma; Takeda, Seiji
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05
no abstracts in English
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
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.
Yamaguchi, Yoshihito; Katsuyama, Jinya; Kaji, Yoshiyuki; Osaka, Masahiko; Li, Y.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
After the Fukushima Daiichi (1F) Nuclear Power Plant accident, we have been developing a prediction method for rupture time and location considering creep damage mechanisms using finite element analysis for early completion of the decommissioning of nuclear power plants in 1F. We have also been obtaining material properties at high temperature near the melting point which are not provided in existing database or literature for the finite element analysis. In this study, we performed uni-axial tensile and creep tests for low alloy steel, Ni-based alloy steel and stainless steels and expanded existing database of material properties. Especially, creep data with longer rupture time at high temperature were obtained by a creep test equipment with a noncontact measurement system. To improve the accuracy of failure evaluation under severe accident conditions, we determined parameters of creep constitutive law based on the expanded database.
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.
Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
To establish a probabilistic approach for assessment of toxic gas leakage accidents in a H
plant, the present study focusses on development of an uncertainty analysis method for toxic gas concentration in a control room. The method consists of 6 steps; (1) Identification of uncertainty factors, (2) derivation of variable parameters, (3) identification of uncertainties in variable parameters, (4) identification of important factors considering the sensitivity analysis results and expert opinions, (5) uncertainty propagation analysis, (6) assessment of uncertainty analysis results. The method is then applied to representative toxic gas leakage accidents in a H plant by IS process coupled to the HTTR. The results obtained in the study leads us to the conclusion that the suggested method can successfully characterize and quantify uncertainties in the toxic gas concentration in control room.
Yamashita, Susumu; Sugawara, Takanori; Yoshida, Hiroyuki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
In order to simulate detailed flow behavior of LBE around the beam window, a numerical simulation code that can evaluate the complicated flow behavior is required. To simulate complicated and large-scale flow behavior, we apply JUPITER which originally developed in JAEA for melt relocation behavior in SAs and that can treat complicated flow behavior and has a capacity of massively parallel computing. In this paper, by using JUPITER, numerical simulations were performed for unsteady thermal-hydraulics simulation around the beam window to know tendency of LBE flow field. In addition, problems to be solved and important parameters to simulate thermal-hydraulic behavior around the beam window will be discussed.
Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 11 Pages, 2019/05
The cooling of the residual core materials after the fuel discharge from the core in the accident of SFRs can significantly affect the distribution fraction of the core materials, which is an important factor for the in-vessel retention (IVR). For the evaluation of the cooling of the residual core materials which is called "in-place cooling", behavior in a SFR core was analyzed preliminary by SIMMER-III. Based on the analysis result, method of phenomena identification and ranking table (PIRT) was applied. Fundamental experiment focusing on three thermal-hydraulic phenomena those were extracted by PIRT was considered in order to investigate them and utilize it for validation of the SIMMER-III. To achieve continuous oscillation of sodium level which can occur in the phase of in-place cooling of SFRs, analytical survey was conducted by SIMMER-III. As a result of that, the effects of experimental conditions on the oscillation amplitude and the duration time were clarified quantitatively, which are necessary to determine the specific experimental conditions.
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).
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
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.
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.
