Vol 80, No 8 (2017)

Cascade reactors attract particular attention because of their capability of improving the parameters of pulsed reactors and achieving the feasibility of electronuclear facilities. The paper presents the results of three series of experiments on uranium-neptunium cascade assemblies at the Institute of Nuclear and Radiation Physics of the All-Russian Research Institute of Experimental Physics conducted in 2003–2004. The experiments confirmed theoretical conclusions on positive properties of cascade blankets and effectiveness of using neptunium-237 as a means of creating a one-sided connection between the sections.

A reactor with a destructible core (RIR reactor) generating a pulse with an output of ~1.5 × 10¹⁹ fissions and a full width at half maximum of ~2.5 μs was developed and tested at VNIIEF. In the course of investigation, a computational-experimental method for laboratory calibration of the reactor was created and worked out. This method ensures a high accuracy of predicting the energy release in a real experiment with excess reactivity of ~3β_eff above prompt criticality. A transportable explosion-proof chamber was also developed, which ensures the safe localization of explosion products of the core of small-sized nuclear devices and charges of high explosives with equivalent mass of up to 100 kg of TNT.

The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

An algorithm for calculation of prompt fission neutron lifetimes in a nuclear reactor by the Monte Carlo method is described. Evaluation of the importance function is carried out with solution of the neutron transport equation without solving the adjoint equation. The results of the prompt neutron lifetime calculations performed within some critical experiments are presented and compared with the experimental results.

The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.

The methodology of description of kinetic phenomena in isomeric γ-reactors is developed by the example of the ^178m2Hf nuclear isomer. The consideration of processes in the isomer medium is based on an assumption that the use of the resonance NEET mechanism would make it possible to put the isomer in the trigger state and thereby to create the conditions for the chain reaction of γ decays similar to that of fissions in neutron nuclear reactors.

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and ²³⁸U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with ²³⁸U alone should be reprocessed in the usual way.

Abstract—A detailed analysis of interdiffusion processes in emitter shells of electrogenerating channels (EGC) of reactors for space application after long term reactor irradiation is performed. Interdiffusion of Mo and W in the mono- and polycrystalline manufacturing technology with the W layer applied over them with thickness on the order of 10^–4 m with irradiation time from 3000 to 12 350 h and at temperatures from 1350 to 1670°C is studied. One of the mechanisms which facilitate the pore-formation process in the W layer and its separation is demonstrated. A method of calculating the activation energy, temperature, and residual thickness of the W layer on the basis of which the state of the shells can be estimated for a long time from the start of irradiation is proposed.

Calculation studies on the effect of carbon and oxygen impurities on the chemical and phase compositions of nitride uranium–plutonium fuel in the course of burnup are performed using the IVTANTHERMO code. It is shown that the number of moles of UN decreases with increasing burnup level, whereas UN_1.466, UN_1.54, and UN_1.73 exhibit a considerable increase. The presence of oxygen and carbon impurities causes an increase in the content of the UN_1.466, UN_1.54 and UN_1.73 phases in the initial fuel by several orders of magnitude, in particular, at a relatively low temperature. At the same time, the presence of impurities abruptly reduces the content of free uranium in unburned fuel. Plutonium in the considered system is contained in form of Pu, PuC, PuC₂, Pu₂C₃, and PuN. Plutonium carbides, as well as uranium carbides, are formed in small amounts. Most of the plutonium remains in the form of nitride PuN, whereas unbound Pu is present only in the areas with a low burnup level and high temperatures.