Reactor core design

The design of the BWR core and fuel is based on the proper combination of many design variables and operating experience. These factors contribute to the achievement of high reliability, excellent performance, and improved fuel cycle economy. 

Several important features of the BWR core design are summarized below. 

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K. Kunitomi, S. Katanishi, S. Takada, X. Yan, N. Tsuji, Reactor Core Design of Gas Turbine High Temperature Reactor 300" Nuclear Engineering and Design, Vol. 230, p. 349 (2004). 

In support of the BN800 reactor core design validation 6 vibro-packed and 8 pelletized MOX fueled sub-assemblies have been tested in the reactor and 4 more sub-assemblies are being tested. 

By present time the 11.3% h.a. peak fuel bumup BN600 reactor core design has been elaborated and is at the stage of approval and finalization. The activities on development of the 12% h.a. peak bumup core have been started. 

The evaluation of the reactor core design is discussed in the following sections. 

BFS-1 critical facility was used to continue studies on the characteristics of fast reactor cores designed for the weapons grade plutonium utilization and minor actinides burning, for instance, the effect of neptunium introduction into fuel. The first stage of these studies made on the insert of the BN-800-Superphenix reactor fuel with up to 14% of depleted uranium dioxide replaced by neptunium dioxide was accomplished in 1995 (BFS-67 critical assemblies set). 

Prismatic graphite-block fuel with traditional refueling. The LS-VHTR would be fueled with prismatic fiiel and refueled when shut down. This particular fuel geometry provides a large latitude for the reactor core designer in the choice of (1) fuel-moderator-coolant ratios and (2) core geometry. 

The configuration and other important characteristics of a fast reactor core design are strongly influenced by the nuclear parameters of the fuel and other core materials. It is important, therefore, to have some assessment of the effect of uncertainties in the basic nuclear data. 

An extensive survey was carried out to identify those uncertainties in nuclear data which were expected to be of significance from the point of view of fast reactor core design. This work was presented in papers by Greebler et al. 8-10) in this study, the discussion is restricted to the effects of only those changes in nuclear parameters which appear to be of greatest importance, and for which strong experimental evidence exists. 

Because all fuel vendors bid on reloads for each others reactor core designs, as well as their own, all design features of each core must be public knowledge. Each reload constitutes one-third of the core, so it must be compatible neutronically and mechanically to fuel already in the core to avoid power generation distortions within the core and equipment mismatches. This applies to all types of reactors, not just PWRs. 

XX-19JHONQ S.G, GREENSPAN, E., KIM, Y.I., The encapsulated nuclear heat source (ENHS) reactor core design. Nuclear Technology, Vol. 149, 1, paper 22—44 (January 2005). 

Inadequate composition and selection of the characteristic items were identified as possible reasons for this unsatisfactory situation. No clear concept of the design characteristics had been established as to what kind of information should be provided. No definitions or instructions for the data providers had been available, which was discouraging from reporting. Some items were not clearly expressed and no options were provided. Some items were too detailed and it was difficult to obtain them. Other items varied with reactor core design changes or with operation and fuel cycle strategy, for example fuel enrichment, so it was difficult to enter just a single value. 

Finally, it should be mentioned that calculations of the diffusion length for systems with holes formed as interstices between random arrangements of solid bodies in point or line contact have also been carried out by Behrens and are reported in the reference paper. These results may be applied in the study of reactor core designs utilizing pebble beds and other similar configurations. 

TWINKLE is a multidimensional spatial neutron kinetics code, whieh is patterned after steady-state codes currently used for reactor core design. The code uses an implicit finite-difference method to solve the two-group transient neutron diffusion equations in one, two, and three dimensions. The code uses six delayed neutron groups and contains a detailed multi-region fuel-clad-coolant heat transfer model for calculating point-wise Doppler and moderator feedback effects. The code handles up to 2000 spatial points and performs its own steady-state initialisation. Aside from basic cross-section data and thermal-hydraulic parameters, the code accepts as input basic driving functions, such as inlet temperature, pressure, flow, boron concentration, control rod motion, and others. Various edits are provided (for example, channel-wise power, axial offset, enthalpy, volumetric surge, point-wise power, and fuel temperatures). 

Each of the material parameters is a constant or can be calculated from information about core construction such as fuel and absorber number densities. Once these values are known, the-core diameter, a, can be calculated using equation (9,8). Reactor core design involves an iteration of setting and resetting core geometry and materials until a satisfactorily sized critical core is obtained. 

IFR reactor core design research and development The reference ALMR power plant described in Section 2 is a paiticulaily well-developed adaptation of the IFR reactor technology. As noted, it incorporates a number of specific innovations and advances and is solidly based on liquid metal fast breeder reactor experience in the U.S., France and U.K. It also benefits from experience obtmned in the former Soviet Union and from technology development in Japan, Italy, Germany, and many other countries. In addition, the IFR metal fuel benefits from years of experience with metallic fuel in the EBR-II, and incorporates fundamental advances in fuels technology. 

XVI-6] K. KUNITOMI, S. KATANISHI, et al. Reactor core design of gas turbine high temperature reactor 300, Nuclear Engineering and Design (submitted). 

Based on the US experience with nuclear steam reheat, it may be concluded that the nuclear steam reheat is possible and higher thermal efficiencies can be achieved however, this implementation requires more complicated reactor core design and better materials. 

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