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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.
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
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.
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.
Umezawa, Katsuhiro; Morimoto, Yasuyuki; Nakayama, Takuya; Nakagiri, Toshio
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05
In December 2016, the Ningyo-toge Environmental Engineering Center of Japan Atomic Energy Agency (JAEA Ningyo-toge) announced new concept of "Uranium and Environmental Research Platform". As part of the concept, uranium waste engineering research are now undergoing. The objective of the research is to establish the processing technology for safely and reasonable disposal of uranium waste. In particular, estimation of the amount of uranium and harmful substances and development of technologies to reduce their concentration in the waste to the permissive level for the disposal in shallow ground disposal are needed. We are now developing the technologies to reduce the concentration of uranium and harmful substances shown below. (1) Survey on uranium inventory. Uranium waste is now stored in 10-odd thousands of 200 L drums. We are surveying amount and chemical form of uranium in the drums. (2) Development of decontamination technology of metal and concrete waste. We are investigating decontamination methods for metals and concrete contaminated with uranium. (3) Development of technologies to remove, detoxify and fix the harmful substances. We are surveying the types and amounts of harmful substances in waste. In addition, we are investigating the method to remove, detoxify, and fix harmful substances. (4) Measurement technology of uranium radioactivity. We are investigating and examining ways to improve the quantitative accuracy of measurement and shorten the measurement time. (5) Development of uranium removal technology from sludge. We are investigating new processing method to remove uranium from sludge which is applicable for several kind of sludge. The results of these technological developments and environmental research will be reflected to "small-scale field test" and "disposal demonstration test" which are planned for demonstration of the uranium waste disposal technology.
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
Matsunaga, Shoko*; Matsubara, Shinichiro*; Kato, Atsushi; Yamano, Hidemasa; D
derlein, C.*; Guillemin, E.*; Hirn, J.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
This paper presents a design of Curie Point Electro-Magnet (CPEM) which will be installed as a passive shutdown system for a French Sodium-cooled Fast Reactor (ASTRID) development program which is conducted in collaboration between France and Japan. To confirm CPEM design validity, a qualification program for CPEM is developed on the basis of past comprehensive test series of Self-Actuated Shutdown System (SASS) in Japan. The main outcome of this paper is results of holding force tests in hot gas, which satisfy design requirements. Moreover, the result of a numerical magnetic field analysis showed the same tendency as that of the holding force test.
Matsuo, Eiji*; Sasa, Kyohei*; Koyama, Kazuya*; Yamano, Hidemasa; Kubo, Shigenobu; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05
Discharged molten-fuel from the core during Core Disruptive Accident (CDA) could become solidified particle debris by fuel-coolant interaction in the lower sodium plenum, and then the debris could form a bed on a core catcher located at the bottom of the reactor vessel. Coolability evaluations for the debris bed are necessary for the design of the core catcher. The purpose of this study is to evaluate the coolability of the debris bed on the core catcher for the ASTRID design. For this purpose, as a first step, the coolability calculations of the debris beds formed both in short term and later phase have been performed by modeling only the debris bed itself. Thus, details of core catcher design and decay heat removal system are not described in this paper. In all the calculations, coolant temperature around the debris bed is a parameter. The calculation tool is the debris bed module implemented into a one-dimensional plant dynamics code, Super-COPD. The evaluations have shown that the debris beds formed both in short term and later phase are coolable by the design which secures sufficient coolant flow around the core catcher located in the cold pool.
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.
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.
