A possibility is indicated of enhancing the tunneling transmission of particles incident on a potential barrier with the aid of an auxiliary potential parametrically in resonance with incident particles. For a simple auxiliary potential, a train of two-step square wells, examples are presented in which the ratio of the transmission coefficients with and without the auxiliary potential can be very large, e.g. (2n+1)^2L, where n is the positive integer and L the number of the period of the auxiliary potential.

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Four air filled beam tubes are present in the small pool type ²³³U fuelled, BeO and water reflected neutron source reactor KAMINI under construction in IGCAR, Kalpakkam. The positive reactivity insertion possible due to the accidental ingress of surrounding water into these empty beam tubes was estimated using the 3-D transport code TRITAC. To save computer time and also to highlight the effects caused by filling of air gaps with water, a similar but symmetrical core with more beam tubes was also studied in detail. Microscopic cross sections were derived from WIMS library and assembly homogenization was performed using a computer code SMAXY. Condensed eight group cross sections were used in TRITAC.
During the study reported here, need was found for the use of tight flux convergence criterion in TRITAC for getting reliable estimates of air to water worths. From the study of the symmetrical core, it is found that the presence of outer water reflector significantly reduces the air to water worth of these beam tubes. Further, the air to water worth of these beam tubes is quite small when the region where water replaces air is, beyond 15 cm from core boundary. Thus it is inferred that the total air to water worth of all beam tubes in the KAMINI reactor will be small compared to the total delayed neutron fraction of ²³³U.

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An improved coarse-mesh discrete ordinates method has been developed for three-dimensional hexagonal transport calculations of high-conversion light water reactors and fast reactors. This method employs a new weighted diamond difference approximation which is obtained by using the neutron balance equations in divided submeshes. The weight is a function of neutron direction and scaler flux, and this method can be easily incorporated into conventional discrete ordinates transport codes.
The present method was applied to hexagonal fuel assembly calculations of high-conversion reactor and fast reactor core calculations, and the results were compared with those of Monte-Carlo calculations. The values of keff and power distributions agreed with each other within 0.5 and 3 %. respectively, verifying accuracy of the present improved coarse-mesh discrete ordinates transport calculation method.

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An experimental small-scale low-pressure setup of a PIUS (Process Inherent Ultimate Safety)-type reactor was used for the examination of the stability during normal operation such as startup and load following operation and of the safety during accidents such as loss-of-feed-water and pump runaway. Automatic feedback pump control system based on differential pressure at lower honeycomb density lock was quite effective to maintain the stratified interface between primary and pool water in the honeycomb density lock during normal operation. The process inherent ultimate safety characteristics of the PIUS-type reactor was confirmed with pump-trip scram at the pump speed limit for the various simulated accidents such as a loss-of-feedwater and pump runaway.

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Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5-10 c) of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1% break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of approximately 5% or more, and might be insufficient for intermediate break areas to maintain adequate core cooling. It is also shown that there might be possibility of core dryout after ACC injection and before LPI injection for break areas less than approximately 2.5%.

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In order to investigate irradiation effects on nodular corrosion resistance of Zircaloy-4, an out-of-pile corrosion test was conducted using Zircaloy-4 specimens cut from the channel box of a fuel assembly irradiated in the BWR (Monticello reactor) up to the neutron fluence of 1.53 x 10²⁶ n/m² (E>1 MeV). The corrosion test was carried out in high pressure steam of 10.3 MPa at 783 K for 24 h. No nodules appeared on the specimens cut from welded areas of the channel box and nodular corrosion resistance tended to be better with increasing neutron fluence. Microstructural evolution in the form of irradiation-induced release of Fe atoms from Zr(Fe, Cr)₂ type precipitates was detected by an analytical electron microscope. It was found that the higher the concentration of dissolved Fe and Cr in the grains of Zircaloy-4, the better the nodular corrosion resistance.

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Sorption experiments of ⁶⁰Co, ⁸⁵Sr, and ¹³⁷Cs were carried out to study sorption characteristics onto a porous tuff. Sorption kinetics and rate controlling step were examined. The effects of particle size on sorption and distribution ratio were also investigated. Desorption experiments were carried out by using extracting reagents to distinguish sorption mechanisms such as reversible sorption in groundwater condition, ion exchange, association with ferro-manganese oxides or oxyhydroxides, and irreversible fixation. Strontium sorbed onto the tuff surface mainly by the ion exchange reaction which was fast and reversible. Cobalt and cesium do not sorb by simple step. The main sorptive binding of cobalt was the association with ferro-manganese oxides and the secondary one was the irreversible fixation. Diffusion into the lattice of rock minerals controlled the sorption rate of cobalt. The main sorption of cesium was the irreversible fixation, secondary one was the ion exchange.

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Mean atomic velocities of uranium, titanium and copper during electron beam evaporation of the metal were measured by a microbalance technique as a function of evaporating metal surface temperature. In the measured temperature range, the mean velocity of uranium was up to 2.1 times the thermal mean velocity v_th, corresponding to the surface temperature of the evaporating metal. Mean velocities of titanium and copper increased to 1.6 and 1.3 times v_th, respectively. This meant that mean velocities of uranium and titanium exceeded the maximum flow speed of an ideal gas in adiabatic expansion, i.e. 1.4 times v_th, while the mean velocity of copper was almost equal to it. The electronic states of uranium and titanium were thermally excited to higher levels, then such excited energy should be converted to kinetic energy during adiabatic expansion. However, copper was rarely excited to higher levels because of the their fewer number. As a result, mean velocities of uranium and titanium were faster than the maximum flow speed of ideal gas.

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Experimental and computational studies have been performed on the temperature coefficients of reactivity in light-water moderated and reflected UO₂ cores with soluble poisons such as boron and gadolinium. Experiments were carried out using the Tank-type Critical Assembly (TCA) in Japan Atomic Energy Research Institute (JAERI). Temperature coefficients of the cores with soluble poisons were measured by changing the temperature of the moderator and reflector from the room temperature to about 60°C. The dependence of temperature coefficients on the core configuration and the concentration of soluble poison was studied with the water level worth method. Temperature coefficients were calculated with a diffusion code CITATION included in the SRAC code system and a perturbation code CIPER for comparison with the experimental data. It was found that the temperature coefficients are always negative in the experimental cores (the water to fuel volume ratio ( V_m/V_f) of 1.83) containing boron as soluble poison. On the other hand, the temperature coefficients become positive in the cores with gadolinium due to the deviation of the gadolinium absorption cross section from the 1/v law and the neutron spectral shift with the increase in temperature.

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In PWRs, loss of decay heat removal (DHR) during reactor shutdown with the reactor coolant system (RCS) partially drained may result in core boiling in a short time. The subsequent RCS pressurization could prevent water flow into the RCS by gravity feed and consequently the core would be uncovered. This paper analyzes U. S. PWR operating experience involving the DHR loss in such reduced inventory conditions.
Between 1976 and 1990, reported were a total of 63 loss of DHR events which occurred during reactor shutdown with the RCS inventory reduced. Review of the event reports indicated that many loss of DHR events in reduced inventory conditions resulted from air entrainment into the DHR pumps due to lowering the reactor water level too far, loss of coolant inventory, increased pump flow and so on.
The coolant heatup rates were evaluated for 12 events with use of the data such as the time elapsed from reactor shutdown actually reported. The calculated results were in reasonably good agreement with the observed ones and showed that core boiling would take place within 1 h even if the DHR loss would occur in the late stage of shutdown (for example, 30 days after the shutdown).

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