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PNC TN9410 98-059, 53 Pages, 1998/06
The analysis program code STEDFAST; Space, TErrestrial and Deep sea FAST reactor ・gas turbine system; had been developed in PNC to get the best values of system parameters on fast reactor ・gas turbine power generation systems used as power sources for deep sea, space and terrestrial cogeneration. In this report, we performed a parameter survey analysis by using the code to study characteristics of the systems. Concerning the deep sea fast reactor ・gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor ・gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were perfomed on the base case of a Na cooled reactor of 40kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concening the terrestrial fast reactor ・gas tubine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100 
C for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100MWt. In the comparison of calculational results for Pb and Na of primary coolant material, The primary coolant weight flow rate was naturally large for the fomer case compared with for the latter case because density is very different between them. ...
PNC TN9410 96-070, 52 Pages, 1996/05
Core characteristics calculations were performed on a trasportable deep sea reactor to confirm subcriticality at the state of thermalized neutron flux during a core flooded accident by sea water. The accident conditions were as follows. (1)A sea water leakage accident through a pressure hull occurred at deep sea. (2)The reactor was shut down by insertion of safety and control rods. (3)The primary loop boundary was damaged for some cause and the sea water entered into the core. Two types of fuel core were studied. One is an oxide fuel core using 50% Pu and 50% U of 20% enrichment. The other is a nitride fuel core using U of 97% enrichment like SP-100. The effect of wire spacer was also analysed. The computer program of MCNP was used for the analysis. Calculation results show that the subcriticality is kept by inserting a Re liner into a fuel pin even for the case with the wire spacer, where much volume of sea water exists in the core. The Re has good absorption effect for thermal neutrons. Thickness of the liner was estimated to be 0.15 ㎜ for the oxide fuel and 0.27 ㎜ for the nitride one.
gas turbine power generation system (User's manual); Sekiguchi, Nobutada
PNC TN9520 95-002, 66 Pages, 1995/02
This analysis program code STEDFAST; Space, TErrestrial and Deep sea FAST reactor gas tubine system; is used to get the adequate values of system parameters on fast reactor gas turbine power generation systems used as power sources for deep sea, space and terrestrial cogeneration. Characteristics of the code are as follows. Objective systems of the code are a deep sea, a space and a terrestrial reactors. Primary coolants of the systems are NaK, Na, Pb and Li. Secondary coolant is the mixture gas of He and Xe. The ratio of He and Xe is arbitrary. Modeling of components in the systems was performed so that detailed modeling might be capable in future and that a transient analytical code could be easily made by using the code. A progra㎜ing language is MAC-FORTRAN. The code can be easily used in a personal computer. The code made possible instant calculation of various state values in a Brayton cycle, understanding the effects of many parameters on thermal efficiency and finding the most adequate values of the parameters. From now on, detailed modeling of the components will be performed. After that, the transient program code will be made.
PNC TN9000 94-006, 60 Pages, 1994/07
Main features were studied about an objective deep sea reactor, which will be used as an electric power source at an unmanned deep sea base. The main features determined are as follows. [Thermal power 190 kWt, Fuel Mixed nitride, Cladding material Hasteloy N, Structual material Type 316 Stainless Steel, Coolant NaK, Core height and diameter about 25cm both, Reactor vessel outlet/inlet temperature 605/ 505C, Operation term 10 years.] Some topic subjects of a talk during deep sea reactor research were studied like follows. Availability of electric transmission from the land or a ship is as follows. (1)The electric transmission from the land is limited up to 1,000m in the depth of water and 100km in the distance from the land. (2)The electric transmission from a ship is available only in the days when the sea is calm. Therefore these transmission methods can not be used as the power source for the base. Concerning reliability, reliability analysis were performed about the part of Closed Brayton Cycle Systems of the reactor. Success probability calculated on the part was 0.999942 in the case of continuous four years operation at 20 kWe. Concerning safety, radioactivity contained in the reactor was calculated. The radioactivity was about 1/50,000 of the radioactivity thrown away in the north Atlantic Ocean from 1962 to 1982. Concerning the experience of developping a NaK cooling reactor in U.S., no anormaly was reported to be found in fuel pins and a reactor vessel after about 400 days operation under a reactor outlet temperature condition over about 527C in the test of a ground test reactor FS-3 for SNAP-10A about thirty years ago.
Iida, Hiromasa; Ishizaka, Yuichi*; Y-C.Kim*; *
Nuclear Technology, 107, p.38 - 48, 1994/07
Times Cited Count：10 Percentile：66.21(Nuclear Science & Technology)no abstracts in English
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PNC TN9410 94-093, 52 Pages, 1994/05
Sodium-water reaction under the high pressure condition of several hundred times of atmospheric pressure was analysed based on the occurence assumption of a hypothetical sea water leakage accident through a pressure hull with the rupture of a bellows type accumulator concerning a deep sea reactor. The sodium-water reaction analysis code of SWACS/REG4 was used in the analysis. The basic calculation case of the analysis adopted the water leakage rate of 0.128 kg/s by considering the accumulator rupture time of 1 s and the water leakage rate used in the heat transfer pipe rupture accident analyses of FBR heat exchangers which had been performed by the code. The calculational result clarified no existence of pressure wave in the upper plenum of a reactor vessel, which was due to the attenuation of the wave in the long slender pipe of 1.5 m in length and 2 cm in inner diameter between the accumulater rupture point and the reactor vesssel. In the analysis, survey calculations were also performed by changing the parameter values of the pressure, the water leakage rate and the gas space volume remaining in the pressure hull.
