[en] Purpose: To supply heats to houses and facilities by using without reprocessing spent fuels in a general reactor. Constitution: A reactor is constructed to incorporate a pressure vessel and a reactor core container installed in the pressure vessed. First fluid for reactor core coolant and moderator such as heavy water or the like is filled in the reactor core container, and second fluid such as light water or the like for a reflector and shielding material is filled between the predssure vessel and the reactor core container. Pipes for transmitting the heat accommodated in the first fluid in the reactor core container to the second fluid are provided at the outer wall of the reactor core container, and heat exchanging means made of pipes in the pressure vessel is provided at the outer wall of the pressure vessel. In this manner, the heat in the reactor core is transmitted through the pipes to the second fluid and further to the exterior. According to this configuration, the spent fuels are used as the fuel assembly to further obtain the degree of burn-up to approx. 7000 MWD/T. (Kamimura, M.)

[en] To overcome the difficulties of the conventional neutron source multiplication method, the authors have developed the Indirect Bias Estimation Method. This method obtains the bias in calculated K_eff using the difference between measured and calculated neutron count rates. This bias in calculated K_eff is used for an adjustment of the calculated K_eff, deriving a truer K_eff. Using neutron diffusion calculations, numerical experiments simulating the neutron source multiplication method were performed to validate this method. It could be shown how closely a true K_eff can be reproduced using this method. It was found that an accurate estimation of subcriticality requires neutron count rate measurements at more than three locations

[en] The Japan Atomic Energy Research Institute (JAERI) is revising and supplementing the first criticality safety handbook of Japan published by the Science and Technology Agency (STA) in 1988, reflecting the opinions of the Working Group chaired by Prof. K. Nishina. The revised version will cover the reactivity effects more precisely and incorporate chemical process data and accident evaluation data. Two entries that will be included in this revised handbook, i.e. fuel grain size effects and partial reflector effects, are discussed in this paper. (Author)

[en] Standard one group constant libraries for the ORIGEN-2 code are prepared using JENDL-3.2 for calculating the amount of nuclides generated in spent fuel of BWR, PWR and FBR, and the adaptability of the libraries is examined for the safety evaluation of the nuclear fuel cycle. This task is performed as a part of the activities of the nuclide generation evaluation working group of Japanese Nuclear Data Committee. (author)

[en] In a step of extracting nuclear fuels by using an organic solvent in a spent nuclear fuel reprocessing step, occurrence of uncontrollable fission chain reactions of fission products such as plutonium and uranium is prevented. Namely, chemically and physically stable organic boron compounds which are soluble only to an extracting organic solvent are added as a soluble neutron poison. In addition, organic boron compounds of high boron and hydrogen content are used, as a fixed neutron absorber, for constituent materials of equipments to be used. m-carbolan or a derivative thereof is used as the organic boron compound. Then, an extractor capable of ensuring critical safety can be designed more rationally without worsening performances of equipments to be used during a spent fuel reprocessing and extracting step. Space for installing a solution storage vessel can be saved, and equipments having complicated shape can be simplified. (N.H.)

[en] To assist succeeding reports which will be presented in this research meeting, following items on the computer code MCNP developed in USA are presented: (1) history of development of MCNP, (2) meaning of the development, (3) progress of study on Monte Carlo codes in the nuclear code committee and (4) expectation to Monte Carlo codes. (author)

[en] A code STEADY-SHIP has been developed to calculate three-dimensional distributions of neutron flux, power and coolant temperature in the reactor core of the nuclear ship MUTSU. The code consists of two parts, that is, a few-group three-dimensional neutron diffusion module DIFFUSION-SHIP and a thermal-hydraulic module HYDRO-SHIP: In the DIFFUSION-SHIP the leakage iteration method is used for solving the three-dimensional neutron diffusion equation with small computer core memory and short computing time; The HYDRO-SHIP performs the general thermal-hydraulic calculation for evaluating feedbacks required in the neutronic calculation by the DIFFUSION-SHIP. The macroscopic nuclear constants are generated by a module CROSS-SHIP as functions of xenon poison, fuel temperature, moderator temperature and moderator density. A module LOCAL-FINE has the capability of computing a detailed rod power distribution for each local node in the core, using the boundary conditions on the surface of the node which were supplied by the STEADY-SHIP whole-core calculation. The applicability of this code to marine reactors has been demonstrated by comparing the computed results with the data measured during the MUTSU land-loaded core critical experiments and with the data obtained during the hot-zero-power tests performed for the actual MUTSU plant. (author)

[en] Heterogeneity effects on reactivity of powdered or slurry fuel were studied through criticality calculations of three-dimensional tiny cells, which were infinitely arrayed and consisted of a spherical fuel pellet of 5 wt% ²³⁵U-enriched uranium dioxide surrounded by water; the diameter of the fuel pellet was varied from 0 (homogeneous) to 6 mm, keeping constant the volume ratio of water to fuel. Reaction rates were calculated by solving the continuous energy transport equations by the Monte Carlo method. The infinite-medium multiplication factor and its four-factors, and their fractional changes from the homogeneous system were obtained. The infinite-medium multiplication factor increased when the system changed from homogeneous to heterogeneous. The results of calculations confirmed that the reactivity increase mainly came from the resonance escape probability p; they also indicated that any uranium-fuel system of grain size less than 100 μm could be treated homogeneous if a 0.3 % increment of reactivity was regarded small enough to be negligible. (author)

[en] It is necessary to consider the interaction of neutron between individual fuel units for the safety analysis of nuclear fuel facilities. The solid angle method has been developed for these analyses, and it requires the effective multiplication factor in a bare fuel unit. A simple evaluation method was examined to calculate the effective multiplication factor in a bare unit from that with water reflector. It is concluded that the ratio of the effective multiplication factor in a fuel unit without to with water reflector is shown by the function of a dimension of a fuel vessel, and this relation is well explained by the one group diffusion theory with calculating migration area, extrapolation length and reflector savings. (author)

[en] In slow source convergence problems, it is often difficult to ascertain whether the source iteration has converged or not. In order to solve this problem, a new 'sandwich method' has been proposed. The essence of this method is that a finally converged eigenvalue k_eff is approached starting from two kinds of initial source guesses which give higher and lower neutron multiplication factors. It is especially important for evaluating nuclear criticality safety to know how to choose a biasing source to obtain an upper limit for k_eff. In this paper, (1) an example is shown to explain the difficulties in ascertaining the source convergence, (2) a method is proposed to obtain the upper and lower limit curves for k_eff by biasing the initial source distribution, (3) the sandwich method is applied to four benchmark problems proposed by the source convergence group of the OECD/NEA Working Party on Nuclear Criticality Safety. Our calculation results show that the sandwich method is an effective means to confirm source convergence in such slow convergence problems. Appendix is prepared to support the method theoretically. (author)