JPS60202386A - Nuclear fuel aggregate for boiling-water type reactor - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a fuel assembly for a boiling water reactor (BWR).

As shown in Fig. 1, a conventional fuel assembly for BWR consists of, for example, 62 fuel rods aυ and two hollow water rods (13) bundled together in an E18 square lattice arrangement, with a channel box surrounding the bundle. It is composed of (13).

In this case, each lattice spacing of the 8 x 8 square lattice is uniform, and the gaps between each fuel rod in the channel box and the water rod are filled with primary energy from the bottom to the top of the core during operation of the reactor. Cooling water is flushed. By the way, in the operating state of the BWR, a large amount of voids are generated in the primary cooling water flowing inside the channel box (13), so the cooling water density distribution in the radial direction of the aggregate is lower in #1 inside the outer wall of the channel box. Therefore, in a BWR, the distribution of thermal neutron flux in the radial direction of the assembly during operation is as shown in Figure 182. When the neutrons flow into the channel box 0□□□ by diffusion, a neutron flux spatial distribution with a depression in the center is formed as shown by curve N. This means that the utilization rate of thermal neutrons is high in the fuel rods near the channel box 03), but the utilization rate is low in the fuel rods in the center of the assembly. It is true that conventional BVVR fuel assemblies are designed with poor neutron economy in terms of thermal neutron utilization.The reason why water rod α2 is placed in the center of the assembly is to improve this utilization. However, in conventional fuel assemblies, the lattice spacing is constant, so there is a limit to the thickness, and the efficiency improvement effect is not sufficient.Furthermore, fuel assemblies with a constant lattice spacing, for example, reduce the loss of coolant. In the event of a LOCA, the fuel rods in the center of the assembly, where radiation heat transfer is poor, may reach high temperatures9, and some may reach the maximum cladding temperature (PCT).

The present invention was made in order to solve the above-mentioned problems, and the lattice arrangement of fuel rods and water rods is arranged so that the lattice spacing is coarse in areas where the thermal neutron utilization rate is relatively low, and By setting the lattice spacing to be dense in areas where the neutron utilization rate is relatively high, the distribution of thermal neutron flux, which differs from fuel rod to fuel rod in the assembly, is flattened as much as possible and the average thermal neutron utilization rate of the assembly is improved. By increasing the average reactivity, the fuel life is lengthened, and the PCT of each fuel rod is made uniform during LOCA, and the PCT as an aggregate is increased.
BW that can expand the safety margin by lowering CT
The present invention provides a fuel assembly for R.

To explain the present invention together with the embodiment drawings, in FIG. 6, (111X112), (113) and (114) are fuel rods,
), the higher the enrichment and the higher the reactivity, which is the same as in general WR fuel assemblies.

These fuel rods (111) to (114) are attached to a channel box (13) along with two water rods (120).
The fuel rods (111
) (112) has a relatively dense lattice spacing, while the fuel rod (112) in the center of the channel has a relatively low thermal neutron utilization rate.
(114) are arranged at relatively coarse lattice spacing.

As mentioned above, the central region where the water rods (120) are arranged has a coarse grid spacing, and therefore the support grid (140) is used to maintain the spacing of the grid arrangement of the fuel assembly.
), the grid folds are also coarse at the periphery and coarse at the center, so the diameter of the water rod (120) is increased to the extent of the large coarse folds at the center of the support grid.

In a fuel assembly with such a configuration, in an operating BWR core, the fuel density is high in the periphery of the channel box where the thermal neutron flux distribution is relatively high, and the fuel density in the channel box where the thermal neutron flux distribution is low compared to this. Since the fuel density is lower in the center, the thermal neutron utilization rate in the radial direction of the assembly is made uniform, and the water rod placed in the center can be made larger in diameter than the fuel rods. The effect of improving the thermal neutron utilization rate by flattening the neutron flux of the water rod can be made higher than before.

Furthermore, assuming a LOCA, the temperature of the fuel rod itself increases due to the decay heat generated by the fuel rod during LOCA, and in this temperature increase process, cooling by radiation becomes dominant, so the temperature between the fuel rods increases. The fuel rod cladding temperature in the center of the channel box, where the interaction of radiation is large, tends to be high, and the emergency core cooling system (EC
The maximum cladding tube temperature (PCT) that C8) reaches until cooling of the fuel rods begins occurs in the fuel rods located in the center of the channel box of the assembly.

On the other hand, in the embodiment shown in FIG. 6, the lattice spacing, that is, the fuel rod spacing, is wider at the center of the channel box than at the periphery, and therefore the radiation interaction between the fuel rods is reduced to 2 Since it is inversely proportional to the power, the interaction of radiation at the center is smaller than in the conventional case and is uniform as a whole. As a result, we have achieved a reduction in the PCT as a whole, making it possible to create a mechanical dog with a safety margin.

As described above, according to the present invention, the average value of the thermal neutron utilization rate of the BWR fuel assembly is improved, so the average reactivity is improved, and the burnup can be improved compared to the conventional method.
Since the PCT during CA can be substantially reduced, the safety margin can be increased.

[Brief explanation of drawings]

Figure 1 is a schematic cross-sectional view showing a conventional example, Figure 2 is Figure 1 h
FIG. 6 is an explanatory diagram showing the spatial distribution of neutron flux in a vertical cross section taken along the -h arrow. FIG. 6 is a schematic U'i cross-sectional diagram showing an embodiment of the present invention. (111) (112) (113) (114): Fuel rod, (120): Water rod, (150): Channel box, (140): Support grid. Agent Patent Attorney Sanro Kimura

Claims (3)

[Claims] (1) A nuclear fuel assembly to be loaded into a boiling point 1B water reactor, comprising a plurality of fuel rods formed by stacking a plurality of nuclear fuel pellets and enclosing them in a cladding tube, and at least one hollow water rods are bundled in a square lattice shape, and in the radial direction of the aggregate of the square lattice arrangement, the lattice spacing of the portion where the thermal neutron utilization rate is relatively low is set to the lattice spacing of the portion where the thermal neutron utilization rate is relatively high. A fuel assembly for a boiling water reactor characterized by having coarse lattice spacing. (2) The fuel assembly for a boiling water reactor according to claim 1, wherein the lattice spacing at the center of the assembly is coarser than the lattice spacing at the periphery of the assembly. (3) The boiling water reactor according to claim 1, characterized in that the water rods having a larger diameter than the fuel rods are arranged in areas with coarse lattice spacing. IK.