Koichi Chino

Cooling mechanism during transient reflooding of a reactor fuel bundle after loss of coolant

Abstract Cooling efficiency during transient reflooding under loss of normal coolant conditions has been examined with a 7 × 7 simulated fuel rod bundle and jet pump bypass. The bundle contains 49 electrically heated rods with 3600 mm heated length and a pseudo cosine axial power distribution. Water is injected into the lower plenum and the superheated bundle is reflooded from the bottom with some flow diverted to the simulated jet pump bypass. The results show that effective cooling can be maintained.

Analysis of radiant heat transfer in a BWR fuel assembly

Abstract A computer code ‘CIDER’ was developed which analyzes radiant heat transfer in a BWR fuel rod bundle under loss of coolant conditions. In the code, (1) a channel box and fuel rods are considered to be gray bodies, (2) reflection and absorption of radiation beams in the atmosphere is neglected, (3) a fuel rod is approximated by a regular polygonal rod, and (4) radiant heat flux is calculated considering circumferential temperature distribution on each fuel rod surface, which is determined from radial and circumferential heat conduction calculations in a fuel rod. It was found that the conventional model with uniform cladding temperature overestimated heat flux about 30% in a typical situation, or correspondingly underestimated the temperature rises.

Incineration of ion exchange resins using concentric burners

A new incineration method, using concentric burners, is studied to reduce the volume of spent ion exchange resins generated from nuclear power plants. Resins are ejected into the center of a propane-oxygen flame and burned within it. The flame length is theoretically evaluated by the diffusion-dominant model. By reforming the burner shape, flame length can be reduced by one-half. The decomposition ratio decreases with larger resin diameters due to the loss of unburned resin from the flame. A flame guide tube is adapted to increase resin holding time in the flame, which improves the decomposition ratio to over 98 wt%.

Stability of cement-glass packages containing sodium borate salt generated from pressurized water reactor power plants

A new solidification technique using cement-glass, which is a mixture of sodium silicate and cement, was studied for solidification of sodium borate salt of liquid waste generated from pressurized water reactor plants. When the sodium borate salt was solidified with the cement-glass, the resulting package contained eight times more sodium borate than was found in cement because it did not interact with sodium borate. The leaching ratio of cesium ion from the cement-glass package was one-tenth that of cement. Its low leaching ratio was due to the high cesium adsorption ability of cement-glass. The ratio could be theoretically evaluated by considering the cesium adsorption-desorption equilibrium.

Porosity and Ion Diffusivity of Latex-Modified Cement

Latex-modified Portland cement, which would be expected to have low permeability and ion diffusivity, was studied for possible application as a solidification agent for radioactive wastes generated from nuclear power plants. In order to predict the leaching ratio of radionuclides from the cementitious waste forms, the effect of water and latex content in the fresh cement paste on total porosity and ion diffusivity of hardened paste was quantitatively estimated. Total porosity of hardened cement paste decreased with the reduction of water content in the fresh paste and it was also reduced by latex addition. This latter effect could be attributed to the latex emulsion forming a water-proof film and filling the capillary pores. Also Cs ions diffusivity, which is the ratio of the diffusion coefficient in pore water to that in bulk water, showed an exponential correlation with total porosity for both cement materials. An empirical equation, expressing ion diffusivity as a function of total porosity, was derived from the consideration that the water constrictivity in this porous medium could cause an increase of the apparent viscosity of pore water. These results suggested a possibility that the transport behavior of radionuclides through the cementitious matrix could be estimated from the mixing parameters of the original cement pastes.

Stability evaluation for cement package containing radioactive waste

In order to provide stable cement packages, ettringite formation, a major cause of cement deterioration, was studied theoretically and experimentally. A computer program was developed to calculate the chemical equilibrium compositions of a complex cement system. Higher curing temperature and the addition of NaOH were identified as effective methods to avoid ettringite formation. These findings were confirmed by measuring the amount of ettringite in solidified cement by an X-ray diffraction method.

A pelletizing model and its application to radioactive waste treatment

A pelletizing mechanism was studied theoretically and experimentally to transform radioactive waste powders into dense and hard pellets in conjunction with the development of a new volume reduction system, a drying and pelletizing system. A pelletizing model was proposed based on plastic deformation of powder particles. Its validity was confirmed by fundamental experiments. From the model, such pellet properties as density and strength can be predicted over a wide compacting pressure range of a pelletizer using one material constant, the shear modulus of the powder, and one experimental value. This led to the proposal of a method to control the pelletizer and to estimate pellet properties. Usefulness of the method was confirmed experimentally from pilot plant tests.

Properties of a radioactive waste pellet package using cement-glass

This paper reports on radioactive waste slurry generated from nuclear power plants which is dried and compressed into pellets. These pellets are dropped in a polymer-impregnated concrete (PIC) barrier and solidified with cement-glass, which is a mixture of sodium silicate and cement. The mechanical strength of the PIC barrier is about three times higher than that of ordinary portland cement because of added steel fibers. The leaching ration form the package is experimentally studied using {sup 14}C, {sup 60}Ci, {sup 85}Sr, {sup 99}Tc, {sup 125}I, and {sup 134}Cs. Because of the low porosity of the PIC barrier, the leaching rate is controlled and increases in proportion to immersion time. The maximum leaching ratio from a 200-l package is estimated to be 0.004/yr.

Filtration of crud by powdered ion exchange resin.

The filtration mechanism of powdered ion exchange resin, which is used in BWR power plants, was studied through filtration experiments of two type resins. A filtration model was proposed to calculate the increase in pressure drop of a powdered resin layer by crud filtration.According to observations of the sectional appearance of the powdered resin layer, crud was adsorbed in it, and this amount decreased exponentially with increased depth. This meant that filtration in powdered resin could be evaluated by a conventional model for deep bed filtration. But this model includes no consideration of contraction of the powdered resin layer, because it was developed for a deep bed without contraction.Therefore, the effect of contraction was measured by basic compaction experiments, and a new model was proposed to evaluate this effect by modification of the conventional one. The new model provided calculation results which agreed within 5% of the experimental ones.