Core Design Method

The core design procedure consists of two parts, nuclear design and thermal-hydraulic analysis. The former is based on the fine-mesh multi-group neutron diffusion solution. The latter is based on single channel analyses for the average and hot channels of all the fuel assemblies. This approach is the same as that in the Super LWR design. 

The nuclear design limits and the nuclear performance are examined in the nuclear design part, while the core outlet temperature and MCST are examined in the single channel analyses. Detailed descriptions for each part are discussed below. 

The nuclear design method is based on the Tri-Z fine mesh finite difference solution of the neutron diffusion equation with multi-neutron energy groups. SRAC2002 is used here. It includes the major neutron data library JENDL-3.3, which contains 107 group neutron cross sections for more than 300 nuclides. 

The unit cell depletion calculation reflects spatial and neutron energy group self shielding effects coming from the spatial distribution of group neutron flux and change of neutron spectrum. At first, one representative cell containing a fuel rod 

Guide tube Wnipperuibe and assembly gap for given conlanl density and biinuip 

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The initial conditions and criteria for MCST in abnormal transients and accidents are shown in Fig. 1.37. The maximum peak temperature at the steady state condition, 740°C, has changed with improvement of the core design method and data as already described in Sect. 1.3.4. But when 740°C is taken, the temperature difference between the limits, 110°C for abnormal transients and 520°C for accidents are the margins. 

M. Kadowaki, Rationalization of Core Design Method and Improved Core Design of Super LWR, Master s thesis. University of Tokyo (2009) (in Japanese)... 

Core Design Method and 1,000 MWe Class Core Design 467... 

In this section, the core design method for the Super FR is introduced, and then an example of a 1,000 MWe core is designed based on the fuel rod design introduced in Sect. 7.4. 

The points 1 to 5 belong to the targeting procedure already presented. The points 6 to 9 form the core of the Pinch Design method that will be explained in the subsequent subchapters. Particularly the point 7 deserves more attention. 

Many early literature examples do not employ quantitative design methods, relying on the use of a privileged core and pragmatic choice of reagent with which to decorate the core. ... 

SAV requirements may be documented using formal methods (i.e., using mathematical notation), semiformal notation (i.e., a mix of natural language with mathematical notation), natural language or a combination thereof. Several notations are also available that are coupled with speeific requirements identification methods [e.g.. Consortium Requirements Engineering (CORE) or Structured Systems Analysis and Design Method (SSADM) as desalbedby Jackson (1988)]. 

Three core questions arise when reviewing the currently available design methods for RD processes. 

This section is composed of four subsections with the following topics (1) noise analysis, (2) low-noise core-amplifier design, (3) dead-time compensahon technique, and (4) input offset voltage cancellation. Considering the four design issues, we will study the VLSI bioinstrumentation design method, analysis, and implementahon. 

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). 

Waltar, A. E. and A. B. Reynolds. 1981. Fast Breeder Reactors. New York Pergamon Press. This book discusses the basic principles and methods as well as design features of fast breeder reactors. Each chapter ends with references and problems to be solved. Appendices include fast reactor data, comparison of homogeneous and heterogeneous core designs, and a list of symbols. 

Proc. NEACRP Specialists Mtg. on Application of Critical Experiments and Operating Data to Core Design via Formal Methods of Cross-Sections Data Adjustment, JACKSON-HOLE (1988) - NEACRP-L-307. 

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