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Kikuchi, Shin; Seino, Hiroshi; Kurihara, Akikazu; Ohshima, Hiroyuki
Journal of Power and Energy Systems (Internet), 7(2), p.79 - 93, 2013/06
For the purpose of elucidating the mechanism of the sodium-water surface reaction in a steam generator of sodium-cooled fast reactors, kinetic study of the sodium (Na)-sodium hydroxide (NaOH) reaction has been carried out by using Differential Thermal Analysis (DTA) technique. The parameters, including melting points of Na and NaOH, phase transition temperature of NaOH, Na-NaOH reaction temperature, and decomposition temperature of sodiumhydride (NaH) have been identified from DTA curves. Na, NaOH and Na
O as major chemicalspecies were identified from the X-ray diffraction (XRD) analysis of the residues after the DTA experiment. It was inferred that NaO could be generated as a reaction product. Based on the measured reaction temperature, the rate constant of sodium monoxide (NaO) generation was obtained by the application of the laws of chemical kinetics. From the estimated rate constant, it was confirmed that NaO generation should be considered during the sodium-water reaction.
Uchibori, Akihiro; Ohshima, Hiroyuki
Journal of Power and Energy Systems (Internet), 7(2), p.106 - 121, 2013/06
In order to establish a safety evaluation method of a steam generator of sodium-cooled fast reactors, a computer program called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction under tube failure accident has been developed. In this study, applicability of the numerical models for the compressible multiphase flow, chemical reaction and liquid droplet entrainment from an interface of the gaseous jet and its transport was investigated through the analyses of the related experiments. In the analysis of vertical discharging of water vapor in the liquid sodium pool under the actual condition of the steam generator, numerical result reproduced appearance of the underexpanded vapor jet. The calculated peak temperature agreed with the experimental results well. Also, appearance of the dispersed phase of droplets in the reacting jet and its acceleration by the supersonic gaseous flow were represented successfully.
Sotsu, Masutake; Kurisaka, Kenichi
Journal of Power and Energy Systems (Internet), 6(3), p.462 - 471, 2012/12
The limiting conditions of operation defined in the safety regulations for MONJU given the allowed outage time were evaluated by a probabilistic safety assessment technique in our previous study. This paper describes a method to assess the validity of the 24 h allowable time in view of core damage risk, it is necessary to analyze the conditions to be changed when a stuck rod is discovered. The results showed that the timeframe defined in the present safety regulations was concluded to be appropriate.
Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki; Hashimoto, Kenro*
Journal of Power and Energy Systems (Internet), 6(2), p.76 - 86, 2012/06
In a sodium-cooled fast reactor (SFR), liquid sodium is used as heat transfer fluid to carry the energy from the reactor core to the steam generation (SG) system for power generation. If the heat transfer tube in the SG is failed, high pressurized water vapor blows into the liquid sodium and the sodium-water reaction (SWR) takes place. The extremely high-temperature reaction jet formed by the SWR, causes damage to the surface of the neighboring heat transfer tubes by thermal and chemical effects. Therefore, it is important to clearly understand the SWR for safety assessment of SG. In this study, we investigated the surface reaction mechanism in the SWR by ab initio method. The potential energy profiles of the dissociations of HO and OH were obtained. The estimated rate constant of the former was much larger than the latter.
Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo
Journal of Power and Energy Systems (Internet), 6(2), p.87 - 98, 2012/06
Ishigaki, Masahiro; Watanabe, Tadashi; Nakamura, Hideo
Journal of Power and Energy Systems (Internet), 6(2), p.264 - 274, 2012/06
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.
Maeda, Shigetaka; Ito, Chikara; Sekine, Takashi; Aoyama, Takafumi
Journal of Power and Energy Systems (Internet), 6(2), p.184 - 196, 2012/06
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 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, Masaaki; Ohshima, Hiroyuki
Journal of Power and Energy Systems (Internet), 6(2), p.210 - 228, 2012/06
A study on flow induced vibration of the primary cooling system of JSFR consisting of a large diameter pipe and a pipe elbow with short curvature radius corresponding to its diameter has been conducted. In this paper, numerical simulations for several pipe elbows with different pipe diameters and curvature radii were conducted at high Reynolds number conditions. Numerical results in each condition were compared with the experimental results in literature. Unsteady flow characteristics and pressure fluctuation generation mechanism in the short-elbow were clarified in relation to the large-scale eddy motion.
Ohno, Shuji; Ohshima, Hiroyuki; Tajima, Yuji*; Oki, Hiroshi*
Journal of Power and Energy Systems (Internet), 6(2), p.241 - 254, 2012/06
The authors are initiating systematic verification and validation activity to demonstrate the reliability of numerical simulation tool for sodium fire behavior postulated in a fast reactor plant. The activity is in progress with the main focuses on already developed sodium fire analysis codes SPHINCS and AQUA-SF. The events to be evaluated are sodium spray, pool, or combined fire accidents followed by thermodynamic behaviors. The present paper describes that the "Phenomena Identification and Ranking Table" is developed at first to clarify the important validation points in the codes, and that an "Assessment Matrix" is proposed which summarizes the tests for validating the computational models for important phenomena. Furthermore, the paper shows a practical validation with a separate effect test in which the spray droplet combustion model of both codes predicts the burned amount of a sodium droplet with the error mostly less than 30%.
Ito, Kei; Ohshima, Hiroyuki; Nakamine, Yoshiaki*; Imai, Yasutomo*
Journal of Power and Energy Systems (Internet), 6(2), p.151 - 164, 2012/06
Suppression of gas entrainment (GE) phenomena caused by free surface vortices are very important to establish an economically superior design of the sodium-cooled fast reactor in Japan (JSFR). Therefore, the authors are developing a CFD-based evaluation method in which the non-linearity and locality of the GE phenomena can be considered. In this study, the authors develop a turbulent vortex model to evaluate the GE phenomena more accurately. Then, the improved GE evaluation method with the turbulent viscosity model is validated by analyzing the GC lengths observed in a simple experiment. The evaluation results show that the GC lengths analyzed by the improved method are shorter in comparison to the original method, and give better agreement with the experimental data.
Uzawa, Ken; Watanabe, Tadashi*
Journal of Power and Energy Systems (Internet), 6(2), p.229 - 240, 2012/06
To prevent anomalous events of equipment and structures in a spent fuel pool from occurring, it is essential to comprehend overflow of water and fluid pressure caused by sloshing. For this purpose, it is necessary to quantitatively evaluate turbulent dissipation near water surface and inner structures instead of an empirical value which depends on size of each pool. Therefore, we test the efficacy of turbulence models by performing numerical simulation of dam break phenomenon as an elementary process of a sloshing to evaluate turbulent dissipation based on action mechanism. In this research, we derived a conservation equation of turbulent kinetic energy of RANS models and compared contributions of pressure, gravity and turbulent dissipation terms to the kinetic energy. As a result, we demonstrated that turbulent dissipation term is a predominant factor on the dynamics of dam break flow and turbulent dissipation is overestimated in eddy viscosity model.
Matsuo, Yoichiro; Miyahara, Shinya; Izumi, Yoshinobu*
Journal of Power and Energy Systems (Internet), 6(1), p.6 - 17, 2012/03
Radioactive corrosion product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE code. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE code incorporating the Particle Model.
Permana, S.; Suzuki, Mitsutoshi; Kuno, Yusuke
Journal of Power and Energy Systems (Internet), 5(3), p.343 - 359, 2011/04
Material attractiveness analysis on isotopic plutonium composition of fast breeder reactors (FBR) has been investigated based on figure of merit (FOM) formulas as key parameters as well as decay heat (DH) and spontaneous fission neutron (SFN) compositions. Increasing minor actinide (MA) doping gives the significant effect to increase 
Pu composition; however, it composition of 
Pu and 
Pu become less. DH and SFN compositions in the core regions similar to the composition of MOX-grade and its DH and SFN compositions of blanket regions are categorized as compositions of weapon grade. Material attractiveness based on FOM1 formula shows all isotopic plutonium compositions in the blanket regions as well as in the core regions are categorized as high attractive material.
Satou, Akira; Maruyama, Yu; Nakamura, Hideo
Journal of Power and Energy Systems (Internet), 5(3), p.263 - 278, 2011/04
A new model for the occurrence of the net vapor generation was developed to improve the predictive capability of best-estimate thermal hydraulic codes for transient void behavior under fast transient condition such as reactivity initiated accidents (RIA). It was clarified that the concept of vapor condensation in the model needed to be improved by analyzing the RIA simulation experiments, thus, the new model for the net vapor generation was developed by using the thickness of thermal boundary layer as a characteristic length of vapor condensation. The new model was introduced into TRAC-BF1 code and was applied to the analyses for the high pressure experiments, confirming that the predictive capability of the modified code was improved.
Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Journal of Power and Energy Systems (Internet), 5(3), p.229 - 240, 2011/04
For a steam generator with straight double-walled heat transfer tubes that used in a sodium cooled faster breeder reactor, clarification of flow instability in heat transfer tubes is one of the most important research themes. 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. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15MPa. Several two-phase flow multiplier models were checked and then, it was found that both two-phase flow multiplier models of Chisholm and homogeneous can predict the present experimental data in high accuracy.
Kurisaka, Kenichi; Nakai, Ryodai; Asayama, Tai; Takaya, Shigeru
Journal of Power and Energy Systems (Internet), 5(1), p.19 - 32, 2011/01
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.
Takaya, Shigeru; Okajima, Satoshi; Kurisaka, Kenichi; Asayama, Tai; Machida, Hideo*; Kamishima, Yoshio*
Journal of Power and Energy Systems (Internet), 5(1), p.60 - 68, 2011/01
Yamano, Hidemasa; Tobita, Yoshiharu
Journal of Power and Energy Systems (Internet), 5(1), p.2 - 18, 2011/01
This paper describes experimental analyses using the SIMMER-III computer code. Two topics of key phenomena in core disruptive accidents were presented in this paper: debris-bed coolability and metallic fuel freezing behavior. Related experimental database were reviewed to choose suitable experiments. 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 CAF
-UT2 experiment carried out using pure UO 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.
Matsuo, Yoichiro; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya
Journal of Power and Energy Systems (Internet), 5(1), p.96 - 107, 2011/01
Radioactive corrosion products (CP) are primary cause of personal radiation exposure during maintenance work at FBR plants with no breached fuel. The PSYCHE code has been developed based on the solution-precipitation model for analysis of CP transfer behavior. We predicted and analyzed the CP solution and precipitation behavior of MONJU to evaluate the applicability of the PSYCHE code to MONJU, using the parameters verified in the calculations for JOYO. From the calculation result pertaining to the MONJU system, distribution of 
Mn deposited in the primary cooling system over 20 years of operation is predicted to be approximately 7 times larger than that of 
Co. In particular, predictions show a notable tendency for Mn precipitation to be distributed in the primary pump and cold-leg. The calculated distribution of Mn and Co in the primary cooling system of MONJU agreed with tendencies of measured distribution of JOYO.
Higuchi, Hidetoshi; Koizumi, Kenji; Hirano, Hiroyasu; Tasaka, Masayuki*; Washiya, Tadahiro; Kobayashi, Tsuguyuki*
Journal of Power and Energy Systems (Internet), 4(1), p.244 - 251, 2010/04
no abstracts in English
