No abstract

In the high temperature gas-cooled reactors (HTGRs), the radial and axial heterogeneity resulted from a combination of fuel rods, burnable poison rods, block end graphite and so on causes local power peakings which increase the fuel temperature locally. An method was developed for calculating the local power and the fuel temperature distributions. T h i s method deals with all heterogeneity effects of a whole core in the radial and axial directions with a design code system including a vectorized 3-dimensional diffusion code. The uncertainty of the method had been evaluated through the analyses of the power distribution obtained by critical experiments with the Very High Temperature Reactor Critical Assembly (VHTRC). The difference was less than 3% between the calculated and measured power distributions. From the results, it was confirmed that this method could predict the local power distribution of the HTGR with high accuracy. It was shown that the maximum fuel temperature would be lower than the design limit of 1,495"C for the normal operation and that of 1,600"C for the anticipated operational transients. This method was applied to the evaluation of the fuel temperature of the HTTR.

This report presents the optimization result with respect to the spatial power distribution of the High Temperature engineering Test Reactor (HTTR) core to achieve a high outlet coolant gas temperature of 950°C. At first, the power distribution optimization procedure was developed to achieve a high outlet coolant gas temperature while maintaining the fuel temperature as low as possible. Secondarily, the optimization procedure thus developed was applied for the power distribution design of the HTTR core. The maximum nominal fuel temperature was reduced about 300°C through the optimization and was 1,321 oc. By the power distribution optimization, the maximum fuel temperature was maintained less than the fuel temperature design limit of 1,600°C, even accounting for the temperature increase at the hot spot and the anticipated operational occurrences.

Xenon oscillations of large graphite-moderated reactors have been analyzed by a multi-group diffusion code with two-and three-dimensional core models to study the effects of the geometric core models and the neutron energy group structures on the evaluation of the Xe oscillation behavior.The study clarified the following. It is important for accurate Xe oscillation simulations to use the neutron energy group structure that describes well the large change in the absorption cross section of Xe in the thermal energy range of 0.1~0.65 eV, because the energy structure in this energy range has significant influences on the amplitude and the period of oscillations in power distributions. Two-dimensional R-Z models can be used instead of three-dimensional R-0-Z models for evaluation of the threshold power of Xe oscillation, but two-dimensional R-0 models cannot be used for evaluation of the threshold power. Although the threshold power evaluated with the R-0-Z models coincides with that of the R-Z models, it does not coincide with that of the R-0 models.

The control rod worths of the AVR power plant for the cold and hot conditions were analyzed to verify an analysis method of reflector control rod worths. In the analysis method, the neutron-flux-weighting method was used to obtain the effective group constants of the control rod inserted in the graphite nose. The control rod worths were calculated with the core analysis code system for the High Temperature engineering Test Reactor (HTTR). The experimental values of the control rod worths were 6.47 and 6.81Xdp for the cold and hot conditions, respectively. The differences between the experimental and analytical values for the cold and hot conditions were 2 and l o % , respectively. From these results, it was made clear that the analysis method applied here predicts well the control rod worth of the AVR power plant and is applicable for the nuclear designs of reflector control rods of future small HTGRs.