Studies on Sodium Boiling Accidents in a Large Fast Breeder Reactor, (II) Radial Propagation of Sodium Boiling by Reactivity Effect

A computer code BORE was developed, with which analyses were performed on channel plugging accidents that would occur on a 1,500 MWe LMFBR. The BORE code calculated the dynamic characteristics of coolant boiling and fuel failure propagation radially in the core, and the requirements of core instrumentation systems were also analyzed. The results show that coolant boiling and/or fuel failure in a channel plugging accident are propagated only to a limited number of adjacent channels when sensors are installed that detect anomalies in channel flow, channel outlet temperature, boiling or reactivity. It is also concluded that the coolant void effect is not serious from the standpoint of safety when the time required for boiling propagation to adjacent channels can be made longer than 0.15 sec.     

Transport and Deposition of Metals in Sodium-Stainless Steel Systems, (VI) Diffusion of Deposited Radioisotopes into Stainless Steel

Abstract

The diffusion of radioactive manganese, zinc and cesium into stainless steel tubes exposed during 700~1,500 hr to liquid sodium containing the radioisotopes was studied by measuring the residual radioactivity on samples with their surface layers successively removed by electropolishing.

Plots of the logarithm of the residual radioactivity in the tube wall drawn against the square of the thickness polished off from the surface were found to fall along straight lines, which changed their slopes in three steps. That of outermost step was most steep, with the residual radioactivity indicating a drop of about one order of magnitude within a layer of a few microns from the surface. The ensuing step presented a much gentler slope, from which the diffusion coefficients were determined. The ratio obtained between the diffusion coefficient and the approximate jump frequency of the diffusing radioisotope was correlated with the reciprocal of absolute temperature, which indicated that the diffusion mechanism of radioactive manganese and zinc differs from that of radioactive cesium.     

Method for Determining Degassing Constant of Xenon in Liquid Sodium

A method for determining the xenon degassing constant from liquid sodium to the cover gas by measuring the ¹³⁵Xe and ^135mXe release rate ratios, which were produced by the decay of ¹³⁵I in liquid sodium, was investigated using the Toshiba inpile fission product loop. Release rate ratios of ¹³⁵Xe to ^135mXe showed values of 9.9~7.6 at a sodium temperature range of 240~390°C and degassing constant of 1.2 × 10^?3-2.6 × 10^?3 sec^?1, which correspond to degassing half-life of 9.7~4.4 min, were obtained. This method is found to be effective for the determination of fission gas degassing constant, which has a corresponding half-life between several minutes to 40 min and will be useful for the study of fission gas transport phenomena in liquid sodium systems.     

Equilibrium Distributions of Actinides and Fission Products in Pyrochemical Separation Systems, (II) LiCl-KCl/Cd System

Equilibrium distributions of actinides and fission products were determined in a LiCl-KCl/Cd system at 800–973 K. The redox potential of the system was controlled with the addition of reductant Li. Different distribution behaviors due to different group elements were observed. The group partitioning in this system was thus supported to be feasible. On reduction, however, the actinide and lanthanide elements were found to be less soluble in the Cd phase and to remain at the interface in the form of their solid intermetallics. A group partitioning process in which reductive extraction is combined with filtration was proposed on the basis of the present observations.     

Transport and Deposition of Metals in Sodium-Stainless Steel Systems, (II)

Making use of the radioisotopes of manganese, zinc and antimony, to represent radioactive corrosion products, and a radioisotope of cesium as representative fission product, the transport and deposition of these radioisotopes in a natural circulation sodium loop have been studied over a temperature range from 630°C down to 280°C.

The movement of the radioisotopes during the loop running were monitored byin situ counting and radiochemical analysis of sodium samples taken from the loop. The distribution of the radioisotopes over the various cross sections along the cooling tube were determined after quick solidification of flowing sodium by chilling.

The results indicate that manganese deposits on the substrate materials presumably by adsorption in the higher-temperature region and by precipitation in the lower-temperature region, zinc presumably by adsorption associated with plating in the lower-temperature region, antimony by coprecipitation with other deposits in the cold settling pocket, and cesium mainly by physical adsorption in the lower- temperature region.

Significant deposition of the radioisotopes was observed in the pocket and the apparent rate constants of the deposition were obtained for manganese, zinc and cesium.     

Prediction of Nongaseous Fission Products Behavior in the Primary Cooling System of High Temperature Gas-Cooled Reactor

In high temperature gas-cooled reactors (HTGRs), some amounts of fission products (FPs) are released mainly from fuel with failed coatings and are transported in the primary cooling system with the primary coolant during normal operation. In that case, condensable FPs plateout on the inner surface of components in the primary cooling system. On the other hand, since the HTGRs use helium gas as primary coolant, the primary coolant is not activated itself and very small amount of corrosion products is generated. Then, γ-ray emitted from the FPs becomes main source in shielding design of the HTGRs, and not only release amount from fuel but also plateout distributions of the FPs should be properly evaluated. Therefore, prediction of plateout behavior in the primary cooling system of HTGRs was carried out based on the calculation result of plateout distribution in High Temperature Engineering Test Reactor. Before the calculation, analytical model was verified by comparison with experimentally obtained plateout distributions and the applicability of the model to predict the plateout distributions in the primary cooling system of HTGR was certified.

This report describes the predicted result of plateout distribution in the primary cooling system of HTGR together with the verification result of the analytical model.     

DHDECMP-TBP/煤油萃取模拟高放废液中的裂片元素工艺研究

实验测定了22％ DHDECMP-42% TBP/OK从模拟高放皮液中萃取Zr、Mo、Cs等裂片元素的分配比，并在微型离心萃取器(转鼓φ10mm)串联台架上进行了从模拟高放废液中萃取裂片元素的工艺研究。采用6级萃取，2级洗涤，6级反萃，流比AF：AX：AS＝1：1．5：0．5，BF：BX＝1：1，实验考察了裂片元素的各级浓度分布。对Zr、Mo的净化系数DF分别为3．2、18；对Rb、Cs、DF、Ba，DF＞10^3。     

Neutronic Analysis for Transmutation of Minor Actinides and Long-Lived Fission Products in a Fusion-Driven Transmuter (FDT)

This study presents the transmutations of both the minor actinides (MAs: ²³⁷Np, ²⁴¹Am, ²⁴³Am and ²⁴⁴Cm) and the long-lived fission products (LLFPs: ⁹⁹Tc, ¹²⁹I and ¹³⁵Cs), discharged from high burn-up PWR-MOX spent fuel, in a fusion-driven transmuter (FDT) and the effects of the MA and LLFP volume fractions on their transmutations. The blanket configuration of the FDT is improved by analyzing various sample blanket design combinations with different radial thicknesses. Two different transmutation zones (TZ_MA and TZ_FP which contain the MA and LLFP nuclides, respectively) are located separately from each other. The volume fractions of the MA and the LLFP are raised from 10 to 20% stepped by 2% and from 10 to 80% stepped by 5%, respectively. The calculations are performed to estimate neutronic parameters and transmutation characteristics per D–T fusion neutron. The conversion ratios (CRs) for the whole of all MAs are about 65–70%. The transmutation rates of the LLFP nuclides increase linearly with the increase of volume fractions of the MA, and the ⁹⁹Tc nuclide among them has the highest transmutation rate. The variations of their transmutation rate per unit volume in the radial direction are quasi-concave parabolic.     

Behaviors of Impurities in Liquid Sodium, (II)

The behavior of hydrogen in liquid sodium containing relatively large amounts of oxygen were investigated with the aid of the hydrogen sensor of niobium membrane. The partial pressure of the hydrogen in liquid sodium at hot zone of a cold-trapped natural circulation sodium loop was measured as functions of temperatures of the cold trap and the hot zone of the loop. It was observed that at constant cold trap temperature the partial pressure of hydrogen in sodium increases with increasing the hot zone temperature. This study also showed that, keeping the temperature of hot zone constant, a logarithmic plot of hydrogen concentration calculated from the equilibrium hydrogen partial pressure vs. the reciprocal temperature of cold trap yields a straight line whose slope is nearly equal to that on the solubility of sodium hydroxide in sodium. Finally, it was observed that the permeation of hydrogen in sodium through a niobium membrane is a process controlled by diffusion. But the permeability for hydrogen in niobium is a few-hundredths smaller than that in literature, owing to the oxide film on the membrane surface.     

In-Situ Measurement of Fission Product Plateout in In-Pile Gas Loop Using Portable Ge(Li) Gamma-Ray Spectrometer

A nondestructive and quantitative method of measuring fission products deposited on inside surface of the primary duct has been developed. A portable Ge(Li) detector with a lead collimator was used for the external scanning of γ-rays emitted from the fission product plateout in the duct.

Upon termination of the first irradiation cycle for the coated particle fuels in the high temperature in-pile gas loop OGL-1 installed in JMTR (Japan Material Testing Reactor), in-situ measurements of the plateout were carried out by the above method at different points in the primary system.

Identified nuclides in the measured γ-ray spectra were ¹³¹I, ¹³³I, ⁹⁵Zr, ⁹⁵Nb, ¹³⁷Cs, ^110mAg, ⁹⁹Mo, ¹⁴⁰La, ⁶⁰Co, ⁵⁸Co, ⁵⁴Mn, ⁵¹Cr, ⁶⁵Zn, ⁵⁹Fe, ¹²²Sb, ¹²⁴Sb, ¹⁸²Ta, ¹⁸¹Hf, ²²Na and ⁴⁶Sc. The plateout densities of these nuclides were obtained using conversion coefficients determined by detector calibration and calculation, which were from 10^-5 to 10^-2 μCi/cm².

It was shown clearly by the experiment that the present measuring method using an in-situ γ-ray spectrometer is useful in the studies of fission product plateout in the primary cooling system of a high temperature gas cooled reactor.     

Development of a Device for Detecting Fuel Failure in a Pool-Type Reactor Effective Extraction of Fission Rare Gases by Introducing Gas Bubbles

Studies have been made for developing a sensitive device for monitoring possible release of fission products from fuel elements of a pool-type research reactor. Gas bubbles were introduced into the reactor-coolant water to extract dissolved fission rare gases, ⁸⁹Kr and ¹³⁸Xe, and their daughter nuclides, ⁸⁹Rb and ¹³⁸Cs, were counted with high efficiency. Various causes that influence the efficiency of extracting the rare gases and of collecting the daughter nuclides were studied by two methods: (1) a “filter method”, in which the air covering water was sucked and the daughter nuclides were caught onto a filter paper, and (2) a “bottle method”, in which the bubbled gas was sampled into a bottle, and the gas was later washed with water, and the activities on the filter or of the water were counted. The items affecting the efficiency of extraction include the flow rate of gas introduced, the position (the depth and the radial direction) of the gas exit in the reactor water, and whether the coolant was circulated or not. Analysis was made of the effect of the depth on the extraction efficiency of dissolved gas into bubbles. On the basis of these results, a fuel-monitoring device for routine use was tentatively designed, both in the filter and in the bottle methods, and tests were made. The filter method was found to be unsatisfactory for practical use. The bottle method enabled an automatic system which functions well enough to be used as a fuel-failure monitoring device.     

Dynamic Response of Hot/Cold Liquid Interfaces to Pump Speed Perturbations in a Thermal-Hydraulic Loop Simulating a PIUS-type Reactor

The operation of a PIUS-type reactor requires controlling the reactor pump speed to keep stationary the hot/cold liquid interfaces between the reactor coolant and cold borated water. The dynamic response of the interface location to pump speed perturbations is analyzed for an experimental loop simulating a PIUS-type reactor. The transfer function between the pump speed and the interface location is obtained by perturbing and Laplace-transforming the one-dimensional fluid momentum equations. The analytical results agree well with experimental data taken from the same facility. It is shown that the magnitude of the phase lag in the response of the interface location, which needs to be considered in designing a pump speed controller, primarily depends on the fluid inertia in the loop, the density lock flow area, and the density difference between the simulated reactor coolant and borated water.