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Nibe, Nobuaki; Shimakawa, Yoshio; ; Hayafune, Hiroki; ; ;
JNC TN9400 2000-074, 388 Pages, 2000/06
JNC-TN9400-2000-074.pdf:13.32MB
Large sized sodium-cooled fast breeder reactors of large-size are being studied and have been operated in Japan and many countries. ln this feasibility study, evaluation was made on technical feasibinty for design concepts or 1 loop type and 3 pool types, specially from the viewpoint of improvement of economical competence. The design concepts include the ideas of cost reduction measures such as large-scaled components, reduction of loop number and integration of components on the basic of utilization of sodium characteristics. From the results of the evaluation, it may be possible for all the concepts to attain the economical target of 200 thousands yen per kilowatt, though further confirmation should be made for technical feasibility of those concepts. ln addition, the following items were listed up as further cost-reduction measures. (1)Higher temperature cooling system and steam cycle efficiency (2)Shortening of construction term (3)Reduction of safety systems by using measuring instruments with high performmce (4)Adoption of SG-ACS
; Yamaguchi, Akira
JNC TN9400 2000-056, 150 Pages, 2000/05
JNC-TN9400-2000-056.pdf:6.67MB
[Purpose] The work was performed to make clear thermal-hydraulic issues resulting in the use of various coolants for fast reactors. [Method] Plant design features due to a use of working fluid other than sodium and design concepts relating a simplification of safety related systems were investigated. And based on the results, quantitative evaluation was made on the topical themal-hydraulic issues. Then both thermal stratification and striping phenomena were evaluated by the used of thermo-hydraulics computer programs. [Results] (1)Thermal-hydraulic issues Topical thermal-hydraulic issues of gaseous and heavy metal cooled reactors were extracted. (a)Gas cooled reactors : natural circulation,flow-induced vibration, depressurization accident (b)Heavy metal cooled reactors : thermal stratification, flow-induced vibration, sloshing And also the thermal-hydraulic issues relating compact reactor assembly and RVACS were extracted resulting from a simplification of safety related systems. (2)Evaluation of thermal stratification and striping phenomena. The following order of affects for the phenomena was obtained: (a) Thermal stratification: CO
Sodium Lead, (b) Thermal Striping: CO Lead Sodium
Tokuhiro, Akira; Kimura, Nobuyuki
JNC TN9400 2000-015, 26 Pages, 1999/09
JNC-TN9400-2000-015.pdf:1.43MB
The quantification of the rate-of-rise of the thermal stratification interface, a "thin" vertical zone where the temperature gradient is the steepest, is important in assessing the potential implications of thermally-induced stress problems in liquid-metal cooled reactors. Thermal stratification can likewise occur in confined volumes containing ordinary fluids (Pr
1), where there is an input of thermal convective energy. In the prominent case of liquid metal reactors, there have been many studies on quantifying the rate-of-rise of a defined stratification interface, in terms of one or more of the following dimensionless groups, mainly: Richardson (Ri), Reynolds (Re), Grashof (Gr), Rayleigh (Ra) and/or Froude (Fr) numbers. Stratification is also a transient process in the volume in question. In the present work the anthors presents a derivation based on order-of-magnitude analysis (OMA), including an sensible energy balance, that produces a new representation more consistent than p
Ohshima, Hiroyuki; Sakai, Takaaki; ; Yamaguchi, Akira; Nishi, Yoshihisa*; Ueda, Nobuyuki*; *
JNC TN9400 2000-077, 223 Pages, 1999/05
JNC-TN9400-2000-077.pdf:6.24MB
The feasibility study (Phase l) is being carried out at JNC to build up new design concepts of practical fast reactors (FRs) from the viewpoint of economy, safety, effective use of resources, reduction of environmental burden and non-proliferation. This report describes the results of the investigation, related to decay heat removal, core/fuel-assembly thermal-hydraulics and thermal-hydraulic correlations, that was performed in fiscal l999 as a part of the feasibility study. ln the study of the decay heat removal, the effects of several design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) in a middle-scale sodium-cooled FR were clarified by using a plant dynamic analysis code. The upper limit of RVACS performance was preliminarily estimated at approximately 0.5
0.6 MWe. Numerical methods for the plant dynamic analysis of gas-and heavy-metal-cooled FRs were also developed. They were applied to the preliminary calculations of the transition from scram to natural circulation and the transient characteristics in tentative plant design concepts were clarified. ln addition, a dimensionless number indicating natural circulation performance was deduced for the comparison of several plant design concepts. With respect to the core/fuel-assembly thermal-hydraulics, numerical analysis methods were improved for the pin-type fuel assembly of gas-and heavy-metal-cooled FRs, the coated-particle- type fuel assembly of helium-gas-cooled FR, and the ductless core of sodium-and heavy-metal-cooled FRs. As preliminary evaluations, thermal-hydraulics in the heavy-metal-cooled FR fuel assembly was compared with sodium-cooled one and thermal-hydraulic analyses of carbon-dioxide- and helium-gas-cooled FR fuel assemblies were performed. The analysis for the fuel assembly with inside duct of sodium-cooled FR was also carried out. The correlations of pressure loss and heat transfer coefficient were investigated for the thermal-hydraulic ...
Morinaga, Masahiko*; *; *
PNC TJ9603 98-002, 48 Pages, 1998/03
[PURPOSE]Both the Nb-based and Mo-based alloys have been designed and developed in order to establish the frontier technique for super-heat-resisting materials used in the liquid alkali metal environment at high temperatures. In this study, mechanical properties of the designed Nb-1Hf alloy were experimentally evaluated. In addition, the brittleness of Nb-based alloys observed at 1073K were discussed. Moreover, characteristics of both the designed Nb-based and the Mo-based alloys were summarized in a consistent way. [EXPERIMENTAL METHODS] (1)Tensile test : The tensile test was performed at room temperature and 1473K in an Ar gas atmosphere for the designed Nb-1Hf alloy and also for commercial Nb-1Zr alloy. (2)High temperature creep test:The creep test of the designed Nb-1Hf alloy was carried out at 1473K in an Ar gas atmosphere under several applied stress levels. (3)TEM observation : The TEM observation was performed with the creep specimens tested at both 1073K and 1273K in order to get information for the 1073K brittleness of the Nb-1Zr alloy. [RESULTS AND DISCUSSIONS] (1)Tensile test : The tensile stress and the proof stress of the designed Nb-1Hf alloy were slightly lower than those of commercial Nb-1Zr alloy at room tempetarure. But the alloy was superior in the elongation to the Nb-1Zr alloy. High temperature tensile properties were not able to be evaluated properly because of the large grain size of the specimens. (2)High temperature creep test : The Nb-1Hf alloy was superior in the ereep resistance to other solid solution hardened Nb-based alloys. (3)TEM observation : A modulated structure with about 1nm preiod was observed in the specimen which was brittle at 1073K. This was supposed to cause the 1073K brittleness of the Nb-1Zr alloy. [CONCLUSION] The tensile strength of the designed Nb-1Hf alloy was slightly lower at room temperature than that of the commercial Nb-1Zr alloy. But, the designed alloy was superior in high temperature creep properties to any
; Haga, Kazuo
PNC TN9410 92-050, 71 Pages, 1992/02
Objective : A conceptual design of a fast reactor system was studied for deep sea research submersibles diving to the maximum depthes of 10,924m and 8,000m. Method : A space reactor concept was used for the system. Primary coolant of the system was NaK, whose temperatures was set as 680 and 550 
C at the exit of a reactor vessel. Secondary system was a closed brayton xcycle using He(60)-Xe(40) gas as working fluid. Electric power output was 20kWe. Thermal efficiency, transported thermal energy, and flow rates and temperatures of the gas and NaK were calculated at closed Brayton cycle analyses. Results : The conceptual design was drawn, based on the size of an each component fixed with the calculated results of these values. The system could be set in a pressure hull comprising of a 2.3m
shere and a 1.1m pipe. A simple figure was drawn of the research submersible loading the system. The whole length of the submersible was about 14m. Its weights were about 100ton and 70ton for the maximum depthes of 10,924m and 8,000m respectively. It could be carried by a nother ship. Conclusion : The submersible had the following features compared with the one loading electric cells on account of affluent electric power generation by the fast reactor system. A continuous stay longer than a week and movement at a high speed were made possible over a deep sea bottom. An illuminated region was very wide during sea bottom survey. Observation by watching was possible over a wide region. Therefore the submersible could be considered to be used for detail observation over crackes in the Japan trench and etc.
; Haga, Kazuo
PNC TN9410 91-305, 20 Pages, 1991/09
Additional calculation was performed concerning "Study of Weight of Fast Reactor Power System for Deep Sea Research" (PNC ZN9410 91-176) published in 1991. In the above report, weight was calculated of power systems of 10kWe for an unmanned base located at the depth of water of 8,020m and of 20kWe for a research submersible diving to that of 10km. The power systems consisted of fast reactor systems and pressure hulls. The main points of the additional calculation were as follows. (1)The depth of water was not fixed, but treated as a parameter. (2)When the power systems will be used in future, some buoyancy material will be attached to the systems in order to make the weight of the systems zero in the sea water. The weight of the buoyancy material was also calculated in this report. (3)In the case of 10kWe power source, electric power was generated by eight sets of closed Brayton cycles of 1.3kWe, but two sets of 5kWe were used in this report. The additional calculation lead to the result that the total weight of the power systems at the depth of water containing the buoyancy material were 13.6t and 13.9t for 10kWe and 20kWe respectively.
Nomura, Norio; Haga, Kazuo;
PNC TN9410 91-298, 74 Pages, 1991/08
High-temperature transpotable liquid-metal-cooled fast reactors are also expected as an energy source in sever environments such as lunar base. Only electricity has been considered as the energy supply system to the lunar base, however, both use of electricity and heat power (co-generation system) may offer better effective use of energy than only the electrification system is considered. In this report, the advantage of co-generation system based on a transpotable reactor examined to a usage in a lunar base. Two energy system diagrams on heat and electricity flows have been compared from the view point of total weight. One is the case of all electrification. The other (co-generation system) is the case in which the exhaust heat from the nuclear reactor is actively utilized to reduce the demand of electricity. A heat transport technique using a chemical process as the heat media is adapted for the latter. In the chemical system, hydrogen and carbon monoxide are formed from methane and steam by adding nuclear heat. Then the gases are transported through a pipe and changes to the original materials generating heat at the consuming spot. As the result of present study, it is clarified that the co-generation system has a possibility to have an advantage over the case of all electrification in total weight of system when the energy demand increase to MWe level. Finally, future R&D items ware mentioned to reduce the weight of this co-generation system and to increase the effectively.
; Haga, Kazuo
PNC TN9410 91-205, 55 Pages, 1991/05
A transportable reactor has been studied in one of activities of frontier research in PNC. Since the reactor is going to be used at an secluded place in the earth, on the surface of the moon or at the deep sea bottom, the operation of the reactor requires wireless communications. Based on the present status of technology, a preliminary study has been performed in this report on a wireless operation method of the reactors on the moon and at the deep sea bottom. A wireless operation system of the reactor on the moon is supposed to exist technically on the extension of a present space communication system and a difficult problem does not seem to remain at the development stage of the wireless operation system. Concerning the wireless operation system of the reactor at the deep sea bottom, a few problems remain to be solved in the field of acoustic communications in sea water. However they seem to be solved technically in future. It takes about three seconds for an electric wave to go and come back between the reactor on the moon and the earth. It takes also about four seconds for an acoustic wave to reach the reactor at the deep sea bottom from the sea surface. Therefore, urgent control of the reactor by wireless communications is impossible in both cases. The urgent control must be performed by the reactors themselves.
Nomura, Norio; Haga, Kazuo;
PNC TN9410 91-100, 73 Pages, 1991/03
Liquid Metal cooled Fast Reactor is a good candidate of a large-scale energy supply system to a manned lunar base because of the compact structure and being free of refueling. A 300 kWe transportable reactor SPECTRA-L is being studied as the power source on the moon. Because the reactor system is launched by a launch vehicle, safety evaluation is necessary to the launch failure accident. We examined (i) the possibility of recriticality, and (ii) the influence of nuclear fuel leakage to the environment in the case of reactor damage. The followings are the main findings of this preliminary study. (1)Under-criticality is maintained even the reactor falls into water or crashes against the earth. (2)The external exposure dose by a radiation cloud of released fuel is less than the natural radiation. (3)The internal exposure dose by inhaling the cloud is less than 1 mSv/year which is a reco㎜ended dose limit to the public. (4)The surface radioactive density increased of land by the accident is be less than 0.4 Bq/square centimeter, which is a limit for things contaminated by alpha radiation to be transported from a controlled area, regardless the whether. This estimation is based on a leakage of five percent fuel, but the exposure dose would be far less than the estimated from the following reasons. (1)Nuclear fuel is in a ceramic form called pellets, and they are inserted in fuel cladding, and contained in the coolant material (metal) and the reactor vessel. (2)The nuclear fuel does not break into fragment as aerosol by the accident.
Otani, Nobuo*
PNC TN9410 90-083, 70 Pages, 1990/07
Core physics was studied on the High Temperature Fast Reactor (HTFR) whose prime objective is to produce hydrogen. Core of HTFR consits of nitride or oxide fuel, and thermal power of a commercial HTFR is assumed to be 300 to 400 MWt. The analysis in this report aims at the core design having negative or small positive void reactivity from view point to attain safety if the reactors, The method of decreasing sodium void reactivity coefficient was to increase neutron leakage through the large surface area of the core by adopting its shape of a pan cake (core height/core diameter=1/2 to 1/3). Result of the analysis revealed that, total void coefficients is negative for all cases analyzed with U fuel. However almost all the cases analyzed had positive void reactivity coefficients for MOX fuel. Burn-up calculation was peformed for U fuel core. Calculational results showed that the excess reactivity of about 5% was necessary to compensate reactivity decrease due to the burn-up during a year. The above calculations were performed using the CITATION code.
Morii, Tadashi*; *; *
PNC TN9410 86-124, 61 Pages, 1986/12
PNC-TN9410-86-124.pdf:3.08MBPNC-TN9410-86-124TR.pdf:3.23MB
On Sept. 27, 1985, a large scale sodium spray fire test (RUN-E1) has been conducted in an air atmosphere using the SOLFA-2 test vessel (100m
made from SUS) of the SAPFIRE facility. The major test conditions are as follows. (Spray Rate : 510 g/sec) (Spray Period : 1800 sec) (Spray Inlet Temperature : 505 C) (Spray Falling Height : 4 m) As a sodium spray started, the gas pressure and temperature rose rapidly and reached to the maximum values 1.24kg/cm
-g and 700C, respectively, after about 1.2 minutes. The oxygen in the test vessel was consumed completely after 4 minutes. From oxygen consumption rate during this time, burning rate of sodium was calculated to be 160g-Na/sec that was equivalent to about 30% of the sodium spray rate (under the assumption of 100% NaO production). Many thermo-couples installed in a spray corn region have been failed due to their exposure to the high temperature above 1000 C, which suggested the existence of a burning zone around the sodium droplets. No remarkable distribution of oxygen concentration was observed in the vertical direction of the vessel during a spray, indicating that the gas within the vessel was well mixed by natural convection due to gas temperature difference between the outside and the inside of a spray corn. Aerosol concentratian has reached the maximum value of 17.5g-Na/m after 5 min and decreased below 1 g-Na/m after 20 min.