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Takamatsu, Kuniyoshi; Ueta, Shohei; Sawa, Kazuhiro
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor (HTGR). All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the Reactor Pressure Vessel (RPV) to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test is performed. From the result of analysis, it is confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR.
Kurikami, Hiroshi; Niizato, Tadafumi; Yasue, Kenichi
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Description of system evolution is an important task in safety and investigation strategies. This paper showed the method to describe system evolution based on safety functions and FEPs and its application to the Horonobe site. Based on the SDM and the Horonobe-specific FEPs, the important FEPs were put in a timeline with the main safety functions. According to the FEPs, we defined thermal and resaturation phase, steady geology phase and geologically evolution phase. In the steady geology phase, the functions of retardation and dilution in the deeper part of the Wakkanai Formation are important, therefore, advection, dispersion and sorption in the domain should be assessed based on the nuclide migration scenario. In the geologically evolution phase, the uplift and denudation are important. Thus, the uplift, denudation and the consequent THMC processes were involved in the reference evolution. Through the application and the discussion, the method was found to be applied to other sites.
Yan, X.; Sato, Hiroyuki; Tachibana, Yukio
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 4 Pages, 2011/10
This paper describes the design of a modular HTGR system cogenerating electricity and process heat and discusses its operational feasibility of electric load follow using a new control scheme. The load follow operations are performed by controlling the reactor coolant inventory while keeping the primary system thermal conditions including reactor power, reactor temperature, and turbine temperatures unchanged. This control strategy is designed to achieve high thermal efficiency in a wide range of part electric load and, by minimizing thermal transient of the reactor, to enable response to rapid load follow demand. The newly proposed control strategy is evaluated by simulation of a bounding load-follow event.
Kurisaka, Kenichi; Okamura, Shigeki*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Japan Atomic Energy Agency (JAEA) has been developing the Japan Sodium-cooled Fast Reactor (JSFR) in the Fast Reactor Cycle Technology Development (FaCT) Project. Risk targets were set out as part of the safety-related design requirement: i.e., the quantitative safety design requirements on the core damage frequency (CDF) and the containment failure frequency (CFF). This paper describes a preliminary evaluation of achievement level of JSFR to the risk targets at the FaCT project phase-I: JFY2006 to JFY2010. A Level-1 PSA has been implemented preliminarily to evaluate the CDF related to internal initiators in power operation. The calculated CDF became lower than the both requirements on CDF and CFF. For seismic events, the seismic fragility of principal structures and components was evaluated in terms of core damage prevention. This evaluation was based on the seismic response analysis, which considered the seismic isolation effect and the hardening effect of the laminated rubber bearing in the isolation devices. As a result, we confirmed that the principal structures and components of JSFR have sufficient seismic margin. Based on this, we judged the risk target could be achieved against the seismic event.
Furukawa, Tomohiro; Kato, Shoichi; Inagaki, Yoshiyuki; Aritomi, Masanori*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 5 Pages, 2011/10
A key problem in the application of a supercritical carbon dioxide (CO
) turbine cycle to a fast breeder reactor is the corrosion of structural materials brought about by supercritical CO at high temperatures. In this study, high-temperature oxidation tests on the structural materials were performed in carbon dioxide pressurized at 0.2 and 1 MPa, and in air, and the oxidation behavior were compared. Results of investigating the effect of CO pressure including the previous reports tested at 10 MPa and at 20 MPa, the effect was hardly observed for all steels. In air environment, weight gain caused by high temperature oxidation was much lower than that in CO.
Kobayashi, Jun; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki; Watanabe, Osamu*; Oyama, Kazuhiro*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Design study of an advanced loop-type sodium-cooled fast reactor, JSFR, has been carried out in a frame work of Fast Reactor Cycle Technology Development Project (FaCT) in Japan. As the temperature differences among the control rod channels, blanket assemblies and the core fuel assemblies are 100
C centigrade in the maximum, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of Upper Internal Structure (UIS). In this investigation, a water experiment was conducted using a 1/3 scale 60 sector model of the core and reactor upper plenum. Characteristics of temperature fluctuations near the cold fluid outlets were obtained and it was confirmed that several countermeasures can reduce temperature fluctuations at the bottom of UIS.
Sano, Yuichi; Katsurai, Kiyomichi; Washiya, Tadahiro; Koizumi, Tsutomu; Matsumoto, Satoshi*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 7 Pages, 2011/10
Japan Atomic Energy Agency has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developed the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver.
Kosaka, Hiroshi; Saegusa, Hiromitsu; Yasue, Kenichi; Kusano, Tomohiro; Onoe, Hironori
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/10
The methodology for estimation of the long-term evolution of groundwater flow conditions are being developed using approaches on the basis of deductive and inductive methods in the case of Tono area. Based on the studies using the approach on the basis of deductive method, it has been confirmed that the method combining physical modeling of topographic change and groundwater flow simulations is useful for estimating of changes in groundwater flow conditions in the future due to topographic and climatic perturbations. Existing information for estimation of surface hydrological conditions, which are to be used for assignment of boundary conditions for the groundwater flow simulation, has been gathered from many sources and reviewed based on modern-analogue methods. In the studies using the approach on the basis of inductive method, paleo-hydrogeological studies have been carried out on several spatial and time scales. Through the study on the largest spatial scale, a methodology needed to understand changes of groundwater flow conditions due to long-term topographic change is proposed to efficiently identify the area to be carried out site characterization involving field investigations. And then, information to estimate the paleo-topography and paleo-climate has been obtained from literature surveys and field investigations. Through these studies, it has been confirmed that these two approaches are useful for estimation of the long-term evolution of deep groundwater flow conditions.
Maeda, Shigetaka; Ito, Chikara; Sekine, Takashi; Aoyama, Takafumi
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
The verification of calculated core characteristics of the Joyo MK-III core using the JUPITER fast reactor standard analysis method was conducted by comparing with the measured values through the core physics tests. The purpose is to upgrade the core performance to increase the driver fuel burn-up and to increase the excess reactivity necessary for conducting various irradiation tests in the core region. Most of the C/Es are within 5% of unity. Through the comparisons, the calculation accuracy of the JUPITER standard analysis method for a small size sodium cooled fast reactor with a hard neutron spectrum like Joyo was clarified. As a result of this study, more irradiation tests can be performed with appropriate safety margin and the efficient core and fuel management can be achieved to save the number of refueling.
Tanaka, Nobuyuki; Nagae, Masahiro*; Ioka, Ikuo; Iwatsuki, Jin; Kubo, Shinji; Onuki, Kaoru
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 4 Pages, 2011/10
The Thermochemical water-splitting cycle IS process constitutes severe environments to the materials of construction because of the corrosive chemicals. This paper discusses corrosion resistance of novel metallic materials in high temperature sulfuric and hydriodic acid. As for the sulfuric acid environment, corrosion resistances of MoN, Mo-Ta-N, and Mo-Cr-N alloys were examined. The experiments were performed for the duration of 5 hour in 90wt% sulfuric acid at 300C. Although no material showed satisfactory corrosion resistance (ca. 1.5 mm/y in the case of MoN), the addition of Ta or Cr was found to be effective to improve the corrosion resistance. As for the hydriodic acid environment, corrosion resistance of Nb-W alloy was examined by immersing the test pieces for up to 100 hours in the test solutions of 200C. It was observed that the corrosion rates lowered with the progress of immersion time and reached to a stable value of lower than 0.1 mm/y.
Sugiyama, Katsuteru*; Noguchi, Hiroki; Takegami, Hiroaki; Onuki, Kaoru; Kaneko, Akiko*; Abe, Yutaka*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 7 Pages, 2011/10
The Japan Atomic Energy Agency has been conducting R&D on thermo-chemical IS process, which is one of most attractive water-splitting hydrogen production methods using nuclear heat of a high temperature gas-cooled reactor. The present study concerns with development of IS process equipment utilizing direct contact heat exchanger (DCHX). The application of DCHX to the sulfuric acid decomposition step of IS process has been proposed such that the decomposed gas contacts with the sulfuric acid solution supplied from the Bunsen reaction step. The concept is very attractive in terms of the development of compact and efficient sulfuric acid concentrator. However, little is known on the behavior of sulfuric acid in the DCHX, which is required for the equipment design. Therefore, we considered an experimental acquisition of essential design parameter of the DCHX, the gas-phase mass transfer coefficient.
Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Iuchi, Hiroshi; Ida, Mizuho; Watanabe, Kazuyoshi; Kanemura, Takuji; Horiike, Hiroshi*; Yamaoka, Nobuo*; Matsushita, Izuru*; et al.
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 7 Pages, 2011/10
Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) were started from July 2007 under an international agreement called ITER Broader Approach. As a major Japanese activity, EVEDA Li test loop (ELTL) to simulate hydraulic and impurity conditions of IFMIF has already designed and is under construction, in which feasibility of hydraulic stability of the liquid Li target, the purification systems of hot traps are major key issues to be validated in this loop. This paper presents the engineering design of the main electro-magnetic pump of the ELTL including the pressure drop calculation and evaluation of the cavitation inception.
O-I solution in the thermochemical hydrogen production iodine-sulfur (IS) processKasahara, Seiji; Guo, H.*; Tanaka, Nobuyuki; Imai, Yoshiyuki; Iwatsuki, Jin; Kubo, Shinji; Onuki, Kaoru
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/10
Flowsheet investigation of the subsection of HI separation from HI-HO-I solution in the thermochemical hydrogen production iodine-sulfur (IS) process was performed. Concentration of HI by electro-electrodialysis (EED) and distillation of HI were applied. Experimental data of the EED cell applying Nafion membrane was used to establish heat/mass balance equations for the cell. Heat/mass balance of HI distillation column was calculated using ESP, a process simulation software. HI molality at the cathode outlet of the cell, pressure in the HI distillation column, and flow rate ratio of the feed to the subsection to distillate of the column were focused as variable parameters for minimum heat demand. Parameters of the EED membrane, electric resistance and upper limit of HI molality between outlet streams, had a great effect on the heat demand; improvement of the membrane parameters is important to reduce the heat demand.
Kamiji, Yu; Terada, Atsuhiko; Sugiyama, Hitoshi*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 7 Pages, 2011/10
Japan Atomic Energy Agency (JAEA) has been conducting R&D on heat utilizing system of High Temperature Gas-cooled Reactor (HTGR). Toward the lower energy consumption and cost reduction of system, we think it is effective to develop waste heat recovery technology. For hydrogen production by thermochemical iodine sulfur process (IS process) which is considered as one of the heat utilization, temperature control unit is an important factors for stable process system driving. Authors focused gas-mixing temperature control and proceeded to design high performance compact unit for process recuperators. From experimental results, in the case of door opening inclining toward flow, there was quite high turbulent kinetic energy up to 3.0U
because of effects of turbulence from separation at tip of the door. Additionally in that door opening, heated flow partially goes to door, so that mixture starts in upstream of mixing region and temperature diffusion gets quite active.
Iwamoto, Yukiharu*; Kondo, Manabu*; Yasuda, Kazunori*; Sogo, Motosuke*; Tanaka, Masaaki; Yamano, Hidemasa
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Tani, Akihiro*; Shimakawa, Yoshio*; Kubo, Shigenobu*; Fujimura, Ken; Yamano, Hidemasa
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Yoshida, Atsuro*; Higashi, Yuma*; Narabayashi, Tadashi*; Khoo Chong Weng, W.*; Arae, Kunihiko*; Tsuji, Masashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/10
One of the design basis accidents in sodium-cooled fast reactor is sodium-water reaction at steam generator (SG). In case of a defect occurred on a heat transfer tube, the high-pressure water/vapor will spout into the low-pressure sodium surrounding outside the tube. As sodium is ordinarily quite reactive with water, this will initiate sodium-water reactions accompanied by high chemical heat generation. The liquid droplet in the reaction steam outflow would impinge on neighboring tubes to cause erosion, while the chemical reaction will cause corrosion, eventually may lead to secondary tube failure. Focusing on the erosion part, this study is to evaluate the liquid droplet impingement erosion (LDIE) rate on neighboring tubes caused by SG heat transfer tube rupture. In this paper, as a basic study, the pressure and temperature distribution of high -pressure two-phase free jet into the air is measured.
Ishigaki, Masahiro; Watanabe, Tadashi; Nakamura, Hideo
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 6 Pages, 2011/10
Two-phase critical flow in the nozzle tube is analyzed numerically by the best estimate code "TRACE" and the CFD code "FLUENT", and the performance of the mass flow rate estimation by the numerical codes is discussed. For the best estimate analysis by the TRACE code, the critical flow option is turned on. The mixture model is used with the cavitation model and the evaporation-condensation model for the numerical simulation by the FLUENT code. Two test cases of the two-phase critical flow are analyzed. One case is the critical flashing flow in a convergent-divergent nozzle (Super Moby Dick experiment), and the other case is the break nozzle flow for a steam generator tube rupture experiment of pressurized water reactors at Large Scale Test Facility of Japan Atomic Energy Agency. The calculation results of the mass flow rates by the numerical simulations show good agreements with the experimental results.
Niizato, Tadafumi; Imai, Hisashi*; Maekawa, Keisuke; Yasue, Kenichi; Kurikami, Hiroshi; Shiozaki, Isao*; Yamashita, Ryo*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
A critical issue for building confidence in the long-term safety of geological disposal is to demonstrate the stability of the geosphere, taking into account its likely future evolution. This study aims to establish comprehensive techniques for characterising the overall evolution of coastal sites through studying the palaeohydrogeological evolution in the coastal system around the Horonobe area, Hokkaido, northern Japan. Information on natural events and processes related to the palaeohydrogelogical evolution of the area have been integrated into the conceptual models that indicates the temporal and spatial sequences of the events and processes, such as climatic and sea-level changes, palaeogeography, and geomorphological and geological evolution in the area. The groundwater flow simulation, which is based on the conceputualisation of the long-term geosphere evolution, shows the sensitivities of natural events and processes on groundwater flow properties in a coastal setting.
Obayashi, Hironari; Sugawara, Takanori; Nishihara, Kenji
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/10