Tomoyasu Mizuno

Tensile Properties of 11Cr-0.5Mo-2W, V, Nb Stainless Steel in LMFBR Environment

The tensile strength of ferritic-martensitic HCr-0.5Mo-2W, Nb, V stainless steel (PNC-FMS), which had been developed for core component applications in LMFBR by Japan Nuclear Cycle Development Institute, was evaluated for the effects of thermal aging, sodium exposure, and neutron irradiation. The tensile strength of thermal aged specimens (~1,023K, ~12,000h) decreased at aging conditions above the initial tempering parameter, and the aging effect was considerably enhanced for the wrapper tubes tempered at lower temperatures. The tensile strength of sodium exposed specimens (~973K, ~10,000h) decreased more than aged specimens due to decarburization, and the effect of decarburization was greater in thin wall cladding tubes. Evaluation of the contribution of both thermal aging and decarburization effects on the tensile strength of cladding tubes irradiated in JOYO (~1,013K, ~6,030h, ~29dpa) suggested that the radiation showed smaller effect on tensile properties than thermal aging and decarburization. By using the derived correlations for thermal aging and decarburization effects, the tensile strength decrease for PNC-FMS after long period (30,000 h) in LMFBR environment was quantitatively calculated.

Advanced MOX Core Design Study of Sodium-Cooled Reactors in Current Feasibility Study on Commercialized Fast Reactor Cycle Systems in Japan

Abstract Sodium-cooled mixed-oxide core design studies are performed with a target burnup of 150 GWd/t and possible measures against the recriticality issues in core disruptive accidents. Four types of core are compared from the viewpoints of core performance and reliability. Results show that all the types of core satisfy the target and that a homogeneous core with an axial blanket partial elimination subassembly is the superior concept, although experimental demonstration is required of molten fuel motion for mitigation of recriticality following fuel melting and loss of fuel pin integrity.

MINOR ACTINIDE-BEARING OXIDE FUEL CORE DESIGN STUDY FOR THE JSFR

In the Fast Reactor Cycle Technology Development (FaCT) project, a sodium-cooled fast reactor (SFR) with mixed-oxide (MOX) fuel and an SFR with metal fuel were selected as the primary and the secondary candidates, respectively, for the Japan Sodium-Cooled Fast Reactor (JSFR). The present study focuses on the effects of transuranium (TRU) composition in the design for the JSFR core with MOX fuel. In the transitional stage from light water reactor (LWR) to fast breeder reactor (FBR), there is the possibility for FBR fuel to have high minor actinide (MA) content due to the recycling of LWR spent fuel. High MA content affects core and fuel designs as follows: the neutronic reactivity characteristic changes; the linear power limit is reduced because of decreases of the melting point and thermal conductivity in the fuel; the gas plenum length is extended because of an increase in He gas generation. Thus, to evaluate the effects quantitatively, design studies for cores with two TRU compositions were conducted: an FBR multirecycle composition with ~1 wt% (in heavy metal) of MA content and an LWR recycle composition for which 3 wt% of MA content was assumed as a tentative target. The results show that the change from the FBR multirecycle composition to the LWR recycle composition leads to a sodium void reactivity increase of 10%, a linear power limit decrease of 1 to 2%, and a gas plenum length increase of 5%. As a result, the effects of TRU composition on the core and fuel designs were revealed to be benign.

A conceptual design study of a small natural convection lead-bismuth-cooled reactor without refueling for 30 years

A small fast reactor has the potential to be utilized as a power source applicable to diversified social needs and to reduce capital risks. At remote sites where the population is small and plants cannot be economically connected to a power grid, power sources without refueling whose capacities are <50 MW(electric) are required because the fuel transfer cost is expensive at such sites. In the present study, a small lead-bismuth-cooled core with a 30-yr lifetime has been developed, and a simple plant system without refueling has been sketched. The dimensions of the major components are determined to evaluate its economic potential. Transient analyses of anticipated-transient-without-scram events show that the design has passive safety features suitable for a remote power source.

HLMC Fast Reactor With Complete Natural Circulation

To seek for a promising concept of a heavy liquid metal coolant (HLMC) fast reactor plant, Japan Nuclear Cycle Development Institute (JNC) and the electric utilities conducted conceptual design study on various types of plant concepts and compared these concepts based on technical feasibility and economical perspective. The Pb-Bi cooled complete natural circulation reactor concept may attain high safety level and construction cost goal (¥200,000/kWe).Copyright