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Yoshida, Hiroyuki; Hosoi, Hideaki*; Suzuki, Takayuki*; Takase, Kazuyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05
Hosoi, Hideaki*; Yoshida, Hiroyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 7 Pages, 2010/05
Tanaka, Nobuyuki; Yamaki, Tetsuya; Asano, Masaharu; Maekawa, Yasunari; Onuki, Kaoru; Hino, Ryutaro
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 5 Pages, 2010/05
For concentrating of HI in HI-I
-HO mixture by electro-electrodialysis (EED) in the thermochemical water-splitting IS process, an application of self-made polymer electrolyte membranes fabricated by radiation-induced graft polymerization and cross-linking method has been studied. In order to bring the EED technology into practical application, stability of membrane is required in the severe environments of high temperature and strongly acidic solution. The present study examined thermal, chemical and electrochemical stability of the grafted membranes in the service environments by performing the EED operation over 100 hours at 373 K, while measuring the evolution of cell voltage and the change of ion exchange capacity. The results showed that chemical cross-linking could largely improve the membrane stability.
Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Sakaba, Nariaki; Tachibana, Yukio
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05
In the present study, a practical safety evaluation method, which enables to design, construct and operate hydrogen production plants under conventional chemical plant standards, is proposed. An event identification for the HTTR-IS nuclear hydrogen production system is conducted in order to select abnormal events which would change the scenario and quantitative results of the evaluation items from the existing HTTR safety evaluation. In addition, a safety analysis is performed for the identified events. The results of safety analysis for the indentified five AOOs and three ACDs show that evaluation items such as a primary cooling system pressure, temperatures of heat transfer tubes at pressure boundary, etc., do not exceed the acceptance criteria during the scenario. In addition, the increase of peak fuel temperature is small in the most severe case, and therefore the reactor core was not damaged and cooled sufficiently.
Furukawa, Tomohiro; Kato, Shoichi; Hirakawa, Yasushi; Kondo, Hiroo; Nakamura, Hiroo
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
The EVEDA lithium test loop is constructed at the Oarai Research and Development Center, Japan Atomic Energy Agency. Since lithium is specified as a dangerous substance by a Japanese law, the countermeasure which assumed the lithium combustion incident is indispensable. In this experimental study, the fire-extinguishing behavior of four kinds of fire extinguishers - dryness sand, pearlite, Natrex-L and Natrex-M - to burning lithium was examined. In addition, the effect of depth of lithium pool on the fire-extinguishing performance of the candidate fire extinguisher was investigated to determine the amount of the fire extinguisher placed at the EVEDA lithium test loop.
Enoeda, Mikio; Hirose, Takanori; Tanigawa, Hisashi; Tsuru, Daigo; Yoshikawa, Akira; Seki, Yohji; Nishi, Hiroshi; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), p.645 - 649, 2010/05
This paper overviews the research and development activity of Water Cooled Ceramic Breeder (WCCB) Blanket in Japan. Japan is performing development of WCCB Blanket as the primary candidate of the breeding blanket for the fusion DEMO reactor. Regarding the development of blanket module fabrication technology, a real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS) F82H. Using fabricated FW mockup, thermo-hydraulic performance and high heat flux tests were successfully performed with the heat flux equivalent to ITER TBM condition, 0.5 MW/m
, 80 cycles with the coolant condition as DEMO, 15 MPa 300 
C. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure have been successfully fabricated. Furthermore, assembling of the real scale FW plate mockup and SW plate mockup was successfully performed. Development of major key technologies for the WCCB TBM structure fabrication has been almost completed.
Naito, Morimasa; Kishi, Hirokazu; Fukuoka, Naomi; Yamada, Tsutomu*; Ishida, Hideaki*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 7 Pages, 2010/05
As an alternative grouting material for the geological repository of long-lived radioactive waste, the "Superfine Spherical silica Grout" (SFSG) material is developed using a fine spherical silica and a fine calcium hydroxide. The developed SFSG material takes an advantage of its smaller particle size distribution (max. 
1 micron or less) than those of the cementitious materials, and also provides a low alkaline environment so as to reduce unfavorable effects on the long-term performance of geological disposal system. The SFSG is a mixture of the super fine silica powder, the superfine calcium hydroxide and additives such as superplasticizer. Some preliminary laboratory experiments were carried out to characterize its fundamental properties from the viewpoint of practical use for geological disposal, which is required to be equivalent with the conventional cementitious materials in terms of penetrability, strength, pH performance and workability.
Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Toda, Mikio*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
JAEA is now performing a FaCT project. The first milestone is in 2010; decisions about whether or not to adopt innovative technologies in the JSFR design will be made in the year. Preliminary assessment is underway to produce recommendations for final discussion. This paper describes some important progress in the preliminary assessment. As for the reactor system design, structural integrity against both thermal stress and seismic force was investigated. Then, the specification of the reactor system was established. Also, investigation of design options to extend a design margin against seismic force has been suggested. Regarding thermal hydraulics issues, design measures have been introduced to restrain cover gas entrainment and vortex cavitations. Further investigation is now in progress for design optimization or improvement of preventive effect. Concerning the piping design of primary cooling circuit, the creep strength reduction by Type-IV damage was taken into account.
Ebara, Shinji*; Aoya, Yuta*; Sato, Tsukasa*; Hashizume, Hidetoshi*; Yuki, Kazuhisa*; Aizawa, Kosuke; Yamano, Hidemasa
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
In this study, pressure measurement test is conducted to find out the FIV characteristic features related to the elbow turbulent flow, using 1/7 scale experimental loop simulating the JSFR cold leg piping. As the first step of multielbow piping, pressure measurement for single elbow was performed. The same measurement procedure was taken as a 1/15 scale experiment to assess the influence originated from the different scale flow at the same Reynolds number.
Terada, Atsuhiko; Yan, X.; Hino, Ryutaro; Sato, Hiroyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 5 Pages, 2010/05
When a primary pipe of the HTGR ruptures, helium coolant gas in the reactor blows out into the reactor confinement structure and the reactor primary system depressurizes. Consequently, the core graphite structures may be oxidized by the air and the complicated natural convection of multi component gas mixtures with chemical reactions would take place inside the reactor. Hence, JAEA studies showed the air ingress phenomena in the depressurized reactor and proposed a new passive mechanism of sustained counter air diffusion (SCAD) that has been shown effective in preventing major air ingress through natural circulation in the reactor. In the present plan, JAEA will construct an experimental reactor mockup including reactor core, the SCAD system, pressure vessel, coaxial pipe and so on. The core is made of graphite or ceramics and heated by electric heaters to allow for test operation up to 1200C. Present status of these activities will be presented.
Takegami, Hiroaki; Terada, Atsuhiko; Onuki, Kaoru; Hino, Ryutaro
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
The Japan Atomic Energy Agency has been conducting R&D on thermochemical water-splitting Iodine-Sulfur (IS) process for hydrogen production. A concept of sulfuric acid decomposer was developed featuring a heat exchanger block made of SiC. Although knowing the strength of the SiC block is important for the reliability assessment, it is difficult to evaluate a large-scale ceramics structure without destructive test. In this study, a novel approach for strength estimation of SiC structure was proposed. Optimum value of the Weibull modulus was determined for evaluating the lowest strength estimation. The strength estimation line was developed by using the determined modulus. The validity of the line was verified by destructive test of SiC block model, which is small-scale model of the SiC block. The fracture strength of small-scale model satisfied the predicted strength of the model.
Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05
This paper describes experimental analyses using the SIMMER-III computer code, which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code. Two topics of key phenomena in core disruptive accidents were presented in this paper: debris-bed coolability and metallic fuel freezing behavior. To analyze the debris-bed coolability, the ACRR-D10 in-pile experiments were selected. SIMMER-III well simulated the heat transfer mechanisms including conduction, boiling and channeling observed in the experiment. Metallic fuel may freeze onto the stainless steel (cladding or wrapper tube) together with eutectic formation during core disruption in a metallic-fueled reactor. The CAFE-UT2 experiment carried out using pure Uranium melt to investigate such phenomena was selected for the experimental analysis. In spite of no eutectic formation model in the SIMMER-III code, the calculated fuel penetration behavior was in good agreement with the experimental data.
Morita, Koji*; Zhang, S.*; Arima, Tatsumi*; Koshizuka, Seiichi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Inoue, Fusao*; Yugo, Hiroaki*; et al.
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of specific phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The specific phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several specific phenomena are summarized.
Narumiya, Yoshiyuki*; Hirano, Mitsumasa*; Hirano, Masashi
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 7 Pages, 2010/05
Over the last thirty years, there have been many accomplishments throughout the world with regard to severe accident studies and the development and application of the PSA techniques. Based on the results and experience gained from these efforts, it is necessary to shift the emphasis toward risk-informed decision-making (RIDM) in Japan. In this context, the Atomic Energy Society of Japan (AESJ) has developed an implementation standard for RIDM. In this report, the content and background of the standard are summarized.
Kurisaka, Kenichi; Nakai, Ryodai; Asayama, Tai; Takaya, Shigeru
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05
The present paper describes a new method for determining the target value of structural reliability in the framework of the System Based Code by considering the safety point of view. In the new method, the reliability target is derived from the proposal to a quantitative safety goal that was published by the nuclear safety commission of Japan and the quantitative safety design requirements on the core damage frequency and the containment failure frequency that were determined in the Fast Reactor Cycle Technology Development Project by Japan Atomic Energy Agency, by utilizing analysis models of a probabilistic safety assessment (PSA). The present method was applied to determination of the reliability target of the structures and components which constitute the reactor cooling system in the Japanese sodium-cooled fast reactor. As a result, we confirmed that the present method combined with the PSA analysis model for internal initiating events is applicable to determination of the reliability target associated with a random failure of the structures and components, and that the method related to seismic initiating events can derive the target value of the occurrence frequency at which any of the important structures and components fails due to an earthquake.
Takada, Shoji; Abe, Kenji; Inagaki, Yoshiyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
The high temperature isolation valve (HTIV) is a key component to assure the safety of a HTGR connected with a hydrogen production system. The structure design of an angle type HTIV was carried out. A seat made of Hasteloy-XR is welded inside a valve box. Internal thermal insulation is employed. Inner diameter of the top of seat was 445 mm. Numerical analysis was carried out to estimate temperature and thermal stress of metallic components by 3-dimensional finite element method. Numerical results showed that the temperature of the seat was simply decreased from the top to the root, and the thermal stress locally increased at the root. Thermal stress was lowered below the allowable limit 120 MPa by optimizing the structure. The thermal stress increased at the top of the seat. A creep analysis showed that a creep damage factor was limited below allowable limit during the start-up and shut-down during normal operation, as well as during the depressurization accident.
Matsuo, Yoichiro; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05
Radioactive corrosion products are main cause of personal radiation exposure during maintenance with no breached fuel in FBR plants. CP is produced in the core region by activation of fuel cladding and sub-assembly wrappers, and they are transported to the primary circuit with sodium flow and deposited on the wall of the primary piping and components. In order to establish the techniques of radiation dose reduction for of personnel, program system for corrosion hazard evaluation code PSYCHE has been was developed. The PSYCHE code is based on the solution-precipitation model. The density of each deposited CP and dose rate of primary coolant system in Monju was estimated by using the PSYCHE and QAD-CG code.
Miyahara, Shinya; Nishimura, Masahiro; Nakagiri, Toshio
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
Equilibrium partition coefficients were experimentally measured using "Transpiration method" for volatile fission products of cesium and iodine between liquid sodium pool and the inert cover gas. The objectives of the experiments are to : (1) Obtain the equilibrium partition coefficients of cesium and iodine at high temperature between 600 and 850 C and (2) Study the dependence of the partition coefficients upon the concentration in the sodium pool. The obtained empirical equations are consistent with Castleman's theoretical equations. The partition coefficients of cesium measured at five different points of mole concentration in the pool were almost consistent with the theoretical values. On the other hand, the measured partition coefficients of iodine increased with the increase in the concentration in the pool and this tendency was incompatible with the theoretical consideration. The reason of this discrepancy might be attributed to the formation of NaI in the cover gas.
Ioka, Ikuo; Suzuki, Jun; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
An Extra High Purity austenitic stainless steel (EHP alloy) was developed with conducting the new multiple refined melting in order to suppress the harmful impurities less than 100 ppm. EHP alloy has great intergranular corrosion resistance. It is considered that intergranular corrosion becomes initiation of SCC. So, we try to apply EHP alloy to weld overlay materials to prevent from SCC. EHP alloy was melted by the new multiple refined method. The conventional weld metals were also prepared as comparisons. Specimens were machined from the welded metal of each material. Intergranular corrosion tests were performed in boiling 8 kmol/m
HNO
solutions containing 1 kg/m Cr(VI) ions. The intergranular corrosion of conventional weld metals was severer than those of EHP alloys. Crevice Beam bending tests to evaluate susceptibility of SCC were carried out in high temperature water of 561 K with saturated oxygen for 1000 h. Cracks and intergranular corrosion of conventional weld metals were much more than those of EHP alloys. It was confirmed that EHP alloy had excellent SCC resistance in comparison with conventional materials when EHP alloy was used as a weld metal.
Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
For steam generator with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important issues need researching. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube with a similar inner diameter as that in the designed SG. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15 MPa. Six models for the prediction of two-phase multiplier were evaluated. The results showed the Chisholm correlation and homogeneous model gave best predictions. Note that in the homogeneous model verification, the homogeneous model was only used in the friction loss calculation. In the calculation of void fraction, which is necessary for static head, drift flux model, instead of homogeneous model, was used.
