Kazuo Azekura

Three-dimensional first-order perturbation calculation method for reactivity changes due to core deformations of fast breeder reactors.

A three-dimensional diffusion calculation method has been proposed to rapidly and accurately calculate reactivity changes of LMFBRs caused by assembly displacements in accidental events. The method requires shorter computation times and provides almost the same accuracy as a conventional direct eigenvalue calculation method. In this method, changes in macroscopic neutron cross-sections and diffusion coefficient are defined so that changes in both region volume and material composition can be treated in a mesh-centered finite-difference program under the same coarse mesh division as used for the normal, non-deformed core. Reactivity changes are calculated from the above-mentioned changes by the first-order perturbation method using normal and adjoint neutron fluxes calculated beforehand for the normal core. The method was applied to deformations of a 1,000-MWe LMFBR core. Reactivity changes calculated by the method agreed within 0.4% with those by a conventional direct eigenvalue calculation method, while ...

Application of Coarse Mesh Method to Neutron Diffusion Calculation for Fast Breeder Reactors

The improved coarse mesh method, which was originally derived by Askew and extended by Takeda, has been modified and applied to a 1,000-MWe and a 300-MWe homogeneous FBR core. In the present method, mesh average neutron flux and mesh center neutron flux are distinguished, and transverse neutron bucklings are taken into account. The results of numerical calculations showed that, with the present method, the power distribution and CR worths are appreciably improved for the 1,000-MWe FBR core with large-pitch fuel assemblies. When CRs are withdrawn, the use of the present method reduces the error of power distribution by half for both cores. However, it yields less satisfactory results, particularly with repect to CR worths, for the 300-MWe FBR core with small-pitch fuel assemblies.

An analysis of initiating and transition phases for an unprotected loss-of-flow accident in an axially heterogeneous fast breeder reactor core

An unprotected loss-of-flow (LOF) event has been analyzed for a 1000-MW (electric) axially heterogenous core (AHC) at the end of an equilibrium cycle, using a realistic model to evaluate the AHC safety potential. The SAS3D code was used for the initiating phase analysis, while a phenomenological approach was employed for the transition phase. The SAS3D results showed that the system rapidly approached subcriticality after experiencing a benign power burst, because of axially flattened fuel worth distribution and reduced sodium-void worth particularly around the core center. During the transition phase, fuel-steel discharge into the interassembly gaps, coupled with engagement of the upper axial blanket material in the core region, was found to result in permanent subcriticality and nonenergetic termination of the LOF event.

Criticality Detection Method Based on Fission Product Gamma Radioactivity Measurement

Abstract In this paper, a method is proposed to evaluate the extent of subcriticality of an accident-damaged nuclear reactor. With this method the activity ratio of two fission product (FP) rare gas nuclides and is measured. From the measured value, the value of the nuclides in the fuel region is estimated by correcting for the time lag incurred when the gases diffuse from the fuel region to the measuring point. A simple expression for an effective multiplication factor has been derived that uses the corrected -to-activity ratio and the -to-fission yield ratios of and but that does not require information on the amount or distribution of the fuel material, making the proposed method very simple. This method has the advantage that FP rare gases can easily leak from the reactor core through many openings and gaps, reaching germanium counters without reacting with other materials.