Volume 66, Issue 1-2

The progress in the technology of conventional power plants with gas, steam and combined cycles suggests investigating to what extent such technology can be applied to power conversion systems for high-temperature reactors (HTR). HTR power conversion systems with steam cycle, offer the possibility of achieving steam parameters equivalent to those of currently constructed or planned conventional plants of similar high efficiency. Gas turbine and combined cycles also offer interesting prospects for obtaining high efficiency. This study, which is restricted to thermodynamic aspects, investigates the potential for increased efficiency in HTR module plants with 200 MWth. Steam and gas turbine cycles, with and without intermediate heat exchangers, and combined gas turbine/steam cycles, with and without intermediate heat exchangers, are investigated.

A passive containment cooling system has been proposed for the advanced heavy water reactor being designed in India. This is to provide long term cooling for the reactor containment following a loss of coolant accident. The system removes energy released into the containment through immersed condensers kept in a pool of water. An important aspect of immersed condenser’s working is the potential degradation of immersed condenser’s performance due to the presence of noncondensable gases. An experimental programme to investigate the passive containment cooling system behaviour and performance has been undertaken in a phased manner. In the first phase, system response tests were conducted on a small scale model to under­stand the phenomena involved. Tests were conducted with constant energy input rate and with varying energy input rate simulating decay heat. With constant energy input rate, pressures in volume V 1 and V 2 reached almost steady value. With varying energy input rate V 1 pressure dropped below the pressure in V 2 . The system could efficiently purge air from V 1 to V 2 . The paper deals with the details of the tests conducted and the results obtained.

Making use of the detonation theory presented in part 1 for melt-water interaction, detonation solutions for different meltwater pairs at different conditions are compared to each other. Discussion is provided on the existence of detonation solutions for water droplet – melt droplet – gas systems. The conclusion is made that even if such solutions can be realized in the nature, which is highly questionable, the resulting detonation pressures will be below 200 bar. This is an important result for judging the risk of the melt-water disperse mixtures in nuclear safety analysis. In addition, the detonation pressures for alumnacontinuous water systems have been found to be stronger then those for urania-continuous water systems, in agreement with the experimental observations and seems to give finally the searched for a long time explanation why alumna-water systems detonate much more violent than urania-water systems.

In order to understand the influences of various factors on the depressurization rate under small break LOCA occurring at the upper plenum in a natural circulation loop, a basic analysis has been carried out. It is shown that the depressurization rate is influenced by system pressure, back pressure, break area, water mass in the system, heating power, operational state of the second loop, and so on. Experimental results of flow and heat transfer are also presented in this paper. The experiment was carried out at the 5 MW nuclear heating reactor test loop HRTL-5. The transient processes are described and heat transfer coefficients are estimated. The results show that flashing improves the natural circulation and the heat transfer, so the wall temperature of the heated rod can be kept at a lower level and the reactor is safe. The flow oscillation, which leads to cyclical variations of important parameters, is described and its development mechanisms are studied. It is pointed out that the interaction of pressure decreasing and bubbles producing caused by flashing is the main reason of the flow oscillation.

The transport of monoenergetic neutrons in spherical geometry with forward scattering and vacuum boundary conditions is considered. The scaled transport equation is solved using the Fn method by considering the pseudo-slab problem. Numerical results for the fundamental and higher order eigenvalues are presented for several significant figures. Some selected results are compared with those obtained using various methods in the literature. Finally, a few remarks about the behavior of the criticality eigenvalue of the neutron transport equation with forward scattering is given.

The performance of water flow measurements with pulsed neutron activation (PNA) have been investigated in the past few years. In this report the following particular Problems were investigated: (a) the effect of collimation on the time-resolved detector signals; (b) the behavior of the background signal, in order to expedite optimum background elimination from the measured signals; and (c) the properties of water mixing, by measuring the geometrical asymmetry of the activated volume as a function of distance from the source as well as of flow velocity. It was seen that with the flow velocities and Reynolds numbers in the present experiments, the mixing is not perfect. Thus, in general, the PNA method will lead to an underestimation of the mass flow due to the effect of the laminar layer close to the wall whose contribution to the total mass flow will be enhanced. The investigations reported here contribute to the improvement of the accuracy of the method.

The experimental results of measurements of the astrophysical S-factor for dd-reaction at keV-energy range collision energies using liner plasma technique are presented. The experiments were carried out at the high current generator of the Institute of High-Current Electronics in Tomsk, Russia. The measured values of the S-factors for the deuteron collision energies 1.80, 2.06 and 227 keV are Sdd=(114±68),(64±30),(53±16)b⋅keV, ${{S}_{dd}}=(114\pm 68),(64\pm 30),(53\pm 16)b\cdot keV,$respectively. The corresponding cross sections for dd-reactions, described as a product of the barrier factor and measured astrophysical S-factor, are σddn(Ecol=1.80keV)=(4.3±2.6). $\sigma _{dd}^{n}\left( {{E}_{col}}=1.80\text{keV} \right)=(4.3\pm 2.6).$ 10−33 cm2;σddn(Ecol=2.06keV)=(9.8±4.6)⋅10−33 cm2;σddn( Ecol= ${{10}^{-33}}~c{{m}^{2}};\sigma _{dd}^{n}\left( {{E}_{col}}=2.06\text{keV} \right)=(9.8\pm 4.6)\cdot {{10}^{-33}}~c{{m}^{\text{2}}};\sigma _{dd}^{n}\left( {{E}_{col}}= \right.$

The neutron yield in the interaction of protons with lead and uranium targets has been studied using the LAHET code system. The dependence of the neutron multiplicity on target dimensions and proton energy has been calculated and the dependence of the energy amplification on the proton energy has been investigated in an accelerator-driven system of a given effective multiplication coefficient. Some of the results are compared with experimental findings and with similar calculations by the DCM/CEM code of Dubna and the FLUKA code system used in CERN.

A specific subcritical reactor system has been suggested and the stability of its parameters against small changes in the geometrical set-up has been estimated using the model calculation code DCM/CEM from Dubna. The dimensions of the system are 200 cm in diameter and 170 cm in length and it supplies a thermal power of 900 MW with k eff =0.943, using a 20 mA proton beam of an energy of 1 GeV. Considering the thermal → electricity power and electricity → beam power conversion efficiencies, the electric power amplification is about 8. Energy deposition and neutron energy distribution in the fission core are also studied. Some properties, such as the heat production per unit volume, are rather similar to modern fast breeders. The neutron multiplication factor k eff is very sensitive to small changes in the geometrical set-up within the fission core.

The concept of a fluidized bed nuclear reactor driven by an energy amplifier system is described. The reactor has promising characteristics of inherent safety and passive cooling. The reactor can easily operate with any desired spectrum in order to be a plutonium burner or have it operate with thorium fuel cycle.

Fuel swelling is a key parameter to be studied in the process of qualifying a fuel for use in reactors. Generally, test irradiations are performed to quantify the swelling. However, all reactor conditions can often not be simulated in tests. This situation occurs, for example, when a fuel shall be used in a new reactor where the fission rate is higher than in existing reactors available for the qualification tests. The swelling under the conditions of the new reactor must be predicted based on the data found under these test conditions. For this prediction a formula is required. Such a formula is derived for one of the most popular fuels for high flux reactors: U 3 Si 2 .

The EC-JRC International Workshop on Promotion of Technical Harmonisation on Risk-Based Decision Making, held at Stresa & Ispra, Italy, 22–25 May 2000, was an experts meeting to discuss the possible need and perspectives of developing an internationally accepted generic “standard” for risk-based decision making. This paper briefly describes the workshop background, its organisation and summarises its main results and conclusions; it reflects the personal opinions of the authors and in no way commits the European Commission.