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The feasibility of rhenium (Re) production by irradiating tungsten (W) metal in a medium size fast reactor was evaluated by using a Monte Carlo code. The fast reactor can produce about 50 kilograms of Re per every 3 years, which corresponds 10% of Japanese domestic production. The specific activity of Re can be reduced below the exemption level or even the natural Re level if W and osmium is separated after the irradiation. The use of ZrD1.7 moderator reduces the specific activity by half compared to that of ZrH1.7 case, and even the no moderator case is permissible to produce the production of Re which has lower specific reactivity than that of natural Re.

The correlation of dose rates inside of the typical Japanese house to different distributions of Cs-137 activities outside are calculated by using the Monte Carlo code MVP with its attached gamma cross section library for aiming at the planning of decontamination strategies and setting the priorities in order to decrease the amounts of waste soil effectively, which will be disposed and stored for a long time with shield. The study is summarized that (1) the decontamination of the roof is not effective in decreasing the dose rate of indoors when surrounding field is contaminated, (2) the plowing of wide area will decrease the dose rate by one tenth even if the depth is only 5 cm, (3) the dose rate of the 2nd floor is higher than that of the 1st floor in some decontamination cases.

A new method is proposed to separate the sodium void reactivity of step type FBR cores to components including non-leakage terms and a leakage term by using a newly developed perturbation code MCPERT where fluxes and adjoint fluxes are derived from a group-wise Monte Carlo code. The step type FBR core is a core where the height of the inner core is smaller than that of the outer core and a large sodium plenum region is located above the core so as to decrease the sodium void reactivity. The conventional diffusion perturbation method cannot treat such a large void region due to the diffusion approximation, while the Monte Carlo code can treat it exactly. In this study, a group-wise Monte Carlo code GMVP with a 70-group constant set JFS-3-J3.3 is employed to evaluate the neutron fluxes and adjoint fluxes which are used as inputs to the MCPERT code to evaluate the non-leakage terms. The leakage term is derived from the difference of the total sodium void reactivity evaluated by the direct calculation of GMVP and the summation of the non-leakage terms. It is found that the proposed method can provide the result approximately consistent to the ratio of the reactivity components derived from the conventional method.

The tolerance capability against ATWS for the FBR core with metallic fuel can be improved by employing a fuel with high thermal conductivity (HTC fuel) instead of the conventional metallic fuel, UPu-Zr. To investigate the self-controllability for the HTC-fueled core with U-Pu-Al alloy fuel, having one order of magnitude higher thermal conductivity than that of the U-Pu-Zr, the core employing the U-Pu-Al fuel was evaluated against ULOF and UTOP. Based on the systematic calculation, it was found that the larger temperature margin between the steady state and ULOF/UTOP conditions caused the excellent tolerance capability against ULOF and UTOP for the HTC-fueled core compared with that for the Zralloy-fueled core. Also, the conditions of the core reactivity coefficients required for neither fuel melting nor coolant boiling were investigated by using a “self-controllability map” consisting of effective fuel and coolant reactivities. As a result, the self-controllable region was found to be expanded especially ...

Metal hydrides have high hydrogen atom density, which is equivalent to that of liquid water. An application of the hafnium hydride has been investigated as a neutron absorber in the Fast Breeder Reactors (FBRs). Fast neutrons are efficiently moderated by hydrogen in Hf hydrides and are absorbed by Hf. Since three isotopes of Hf have large cross sections, increase in the life of control rod is considered by Hf hydride. Results of design study of the core with Hf hydride control rods shows that the long lived hafnium hydride control rod is feasible in the large sodium-cooled FBR. Results of irradiation test conducted in BOR-60 has demonstrated the integrity of the capsules during irradiation. Na bonded capsule has an advantage in confinement effect of hydrogen compared with He bonded one. An application of hydride technique to transmutation target of MA was also discussed. MA hydride target is able to enhance the transmutation rate in FBR.

An inherently safe core concept with metallic fuel for sodium cooled fast reactor is proposed that has a negative void reactivity at the loss of coolant events without scram as well as a small excess reactivity during the operation cycle. The relationship of sodium void reactivities and burn-up reactivities to different core configurations has been studied quantitatively to clarify the core concept for large metallic fuel reactors. It has shown that the sodium void reactivity is greatly dependent on the core shapes while the excess reactivity is on the fuel compositions. It has also indicated that the core configuration that enables to enhance the neutron streaming through the region above the active core at coolant voiding is most effective to decrease sodium void reactivity. A 3000 MWt core composed of the flat inner core and annular outer core where the fuel volume fraction is relatively high and the sodium plenum is placed just above the active core region has been selected as a candidate core. The maximum excess reactivity of the candidate core at UTOP is about 0.4 $ and it can be reduced to approximately zero by power or inlet temperature adjustment during the operation cycle, meanwhile the sodium void reactivity is as low as -1.3 $ in negative that is enough to prevent ULOF sequences.

CRIEPI and Toshiba Corp. have been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. A conceptual design of 4S ( S uper- S afe, S mall and S imple) reactor is proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void reactivity are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core lifetime is more than 10 years. The 4S reactor is a metallic fueled sodium cooled fast reactor. A target of an electrical output is 10–50 MW. A remarkable feature of 4S is that its reactivity is not controlled by neutron absorber rods but by neutron reflectors to cope with a long core lifetime and a negative coolant void reactivity. This study includes a design consideration of 4S. Design discussions are mainly focused on various core designs to meet above requirements. A tall core active height is adopted to gain long core lifetime. An averaged fuel burn-up is tried to be increased up to 100 GWd/ton from a point of economic view. The reference 4S designs are 10 MWe (30 years core lifetime) and 50 MWe (10 years core lifetime).

A hydride target including minor actinides (MA) is able to enhance the transmutation rate in a fast breeding reactor (FBR) without degradation of core safety characters. Fast neutrons generated in the core region of the FBR are moderated in the MA-hydride target assemblies and then efficiently absorbed by MA. The MA-hydride target pin has been designed in the light of recent research of hydride materials. This chapter shows the feasibility of MA transmutation by an existing reactor, Monju.

CRIEPI (Central Research Institute of Electric Power Industry) has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application to dispersed energy supply and multipurpose use, with Toshiba Corporation [1,2,3,4]. Electrical output of the 4S reactor is from 10MW to 50MW, and burn-up reactivity loss is regulated by neutron reflectors. The reflector that surrounds the core is gradually lifted up to control the reactivity according to core burn-up. 30year core lifetime without refueling can be achieved with the 10MW 4S (4S-10M) reactor. All temperature feedback reactivity coefficients, including coolant void reactivity, of the 4S-10M are negative during the 30year lifetime. A neutron absorption rod is set at the center of the reactor core with the ultimate shutdown rod. The neutron absorption rod used during the former 14 years is moved to the upper part of the reactor core, and the operation is continued through the latter 16 years. The pressure loss of the reactor core is lower than 2kg/cm2 to enable effective utilization of the natural circulation force, and the average burn-up rate is 76GWD/t. To suppress the influence of the scale disadvantage, loop-type reactor design is one of the candidates for the 4S-10M. The size of the reactor vessel can be miniaturized by adopting the loop type design (4S-10ML). In the 4S-10ML design, integrated equipment which includes primary and secondary electromagnetic pumps (EMPs), an intermediate heat exchanger (IHX) and a steam generator (SG) is adopted and collocated by the reactor vessel. The decay heat removal systems of 4S-10ML consist of the reactor vessel air cooling system (RVACS) and SGACS (a similar system to the RVACS, with air cooling of the outside of the integrated equipment vessel). They are completely passive systems. To decrease the construction cost of the reactor building, a step mat structure and the horizontal aseismic structure are adopted. 4S-10ML has unique features in the cooling systems such as integrated equipment and two separate passive decay heat removal systems which operate at the same time. To evaluate the design feasibility, the transition analyses were executed by the CERES code developed by CRIEPI [5]. In this paper, the design concept of 4S-10ML, and the results of the plant transition analyses are described.

CRIEPI has been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. And a conceptual design of 4S (Super-Safe, Small and Simple) reactor was proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void coefficient are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core life time is more than 10 years; (5) Its construction, maintenance and operation are expected to be very simple by eliminating active components from inside of a reactor vessel. The 4S reactor is a sodium cooled fast reactor and its reactivity is not controlled by neutron absorber rods but by neutron reflectors. An electrical output is 50 MW. This paper describes a design modification to enhance the feasibility from the previous 4S design. A core active height can be shortened to 1.5 m from 4.0m to keep the reactivity characteristics. An averaged fuel burn-up is up to 70 GWD/ton and a pressure drop at the core region is less than 0.1 MPa. A reactivity control system is modified according with the core design change. As for the steam generator design, sodium-water reaction accidents must be taken into account as a design basis event for the utilization of the secondary sodium coolant. Therefore, a modified plate type heat exchanger is proposed as a steam generator. It may be possible to develop a compact steam generator, which is free from sodium-water reaction accidents and to eliminate the secondary sodium systems. The 4S reactor without secondary system has been proposed as a candidate design.Copyright © 2002 by ASME

Gamma-dose analysis was performed on an absolute-value basis at the zero-power plutonium reactor (ZPPR)-13B/4 critical assembly, which was one of the benchmark radially heterogeneous 650-MW (electric) liquid-metal reactor cores, to validate the current data and methods applicable to power reactor design calculations. Discussions of particular aspects inherent to the critical measurements, such as heterogeneity in the ZPPR plate cells and delayed fission product gamma effect, are included. The results show that the gamma-energy deposition distribution was reproduced within an accuracy of {approximately} 10% in cores that had complicated fuel and internal blanket layouts.

Design studies are performed for a commercial liquid-metal fast breeder reactor core that can achieve a burnup of 200 GWd/t. A plutonium-type asymmetric parfait core with two different plutonium-enriched zones in the axial direction as well as in the radial direction is studied. This core concept solves core design problems related to high burnup, and it is possible to achieve a burnup of 200 GWd/t with this concept. A core with ductless fuel assemblies suitable for high burnup is also studied. An axially heterogeneous core was selected from among various concepts. It is possible to realize a core with a burnup of 200 GWd/t, a compact size, and a lower core pressure drop than the demonstration reactor design.