Volume 69, Issue 3

The recently developed tracking routine PIJM, for computing improved two-dimensional first-flight collision matrices for the WIMSD code, has been modified to evaluate black Dancoff factors for cluster fuel bundles in unit cells with cylindrical outer boundaries. The routine, similar to the original PIJ but employing an efficient in-plane integration scheme of the Bickley functions appearing in those expressions, is first compared with the accurate Carlvik method in the calculation of Dancoff factors for regular lattices, both with and without a gap surrounding the fuel rods. Dancoff factors for the more general cases of CANDU37 and CANFLEX fuel bundles are next obtained and compared with Monte Carlo based calculations, illustrating the accuracy of the proposed scheme.

The purpose of this paper is to model the additional nuclear power and fuel burn-up needed to compensate non-optimized operation. Technical limitations, such as changes in reactivity, the additional nuclear power needed to compensate deficiencies of the electrical system control, and other related nuclear parameters are evaluated. The equations that define and quantify the additional nuclear power that must be generated by a reactor in order to fulfill the commitment with the national electric system operator are modeled in a simple and precise way.

In the last three years 3 events were observed in which the release rate of fission products from the fuel plates into the pool water increased significantly. Although intervention levels laid down in the permit and safety analysis were never reached, the ALARA principle was the incentive to analyse the situation and take appropriate actions.

The paper presents the comparative study of modern approaches, i. e. “best-estimate” and “partially-conservative” for the accident analyses based on calculations of Maximum Design Basis Accident at the Ignalina Nuclear Power Plant. The calculated critical parameters, values of which lead to the determination of the safety margin, were compared for both analyses. The comparison of the results shows that these values for “best-estimate” calculation are slightly lower in comparison with “partially-conservative” calculation results. That enables to draw a conclusion that in most cases both approaches – either “best-estimate” or “partially-conservative” can be applied. The latter approach looks tempting since in this case only one calculation is sufficient; while in the case of the “best-estimate” approach a high number of calculations is required. Thus, the “partially-conservative” approach requires considerably less computational time. However, when using this methodology the obtained calculated results do not meet the acceptance criteria, the complete analysis by employing “best-estimate” approach is necessary.

The energy deposition was calculated for the targets from carbon to uranium irradiated with intermediate energy protons with the help of different models incorporated in the MCNPX code package (Bertini, CEM and ISABEL) and with the help of the CASCADE/INPE code. The values obtained using the different models and codes are in a good agreement for all targets except uranium. The comparison with the available experimental data for the heat deposition for 0.8, 1.0 and 1.2 GeV protons has been performed. The good agreement is observed for copper, lead and bismuth target. The best result is obtained with the help of the ISABEL model. The systematic dependence of the heat deposition from atomic number of the target was investigated. The contribution of different particles and energy ranges in the heat deposition has been studied at the primary proton energies from 0.3 to 2.5 GeV.

Neutron spectra, yields and energies of particles generated by a uranium electro-nuclear set-up which is now used in Dubna for transmutation research of various isotopes with a proton beam of an energy of 1–2 GeV, were analyzed by means of mathematical modelling based on the Monte Carlo code CASCADE. The influence of space dispersion on the proton beam and its possible deviation from the center of a lead target is investigated. It is shown that the polyethylene radiation shielding that is being used now, does not protect the measuring room from neutrons and γ-rays.

A steady state thermal-hydraulic analysis for the ETRR-2 Research Reactor has been performed using the computer code TERMIC. The analysis has included both, the verification of the core design as well as the performance of the equilibrium core. The uncertainties due to fuel fabrication tolerances, possible deviations in the operational measurements, the errors in the experimental correlations used in the analysis, and the errors due to simplifications made in the calculations have been considered in the analysis. A methodology based on uncertainties propagation has been adopted for the evaluation of hot channel factors. The results of the thermal-hydraulic analysis show that the reactor can be operated at full power level with sufficient safety margins against the thermal-hydraulic critical phenomena.

The MCNP computer Code is used to model the ETRR-2 research reactor. A computer program was designed to evaluate the axial burn-up of the fuel elements. The excess reactivity of the reactor core is calculated for different core configurations and compared with the existing measurements. The thermal flux is also calculated and compared with measurements. Several factors that affect the safety of the reactor such as power peak and the effect of control rod insertion on the reactor power and flux were studied and analysed. The agreement between the MCNP results and the experimentally determined values is good.

The International Atomic Energy Agency (IAEA) through its INPRO Project has developed a methodology to evaluate the innovative nuclear reactors. Here is a preliminary application of this methodology to evaluate the Fixed and Fluidized Bed Nuclear Reactor Concept (FBNR). The FBNR is based on pressurized light water reactor technology. It is a small, modular, and integrated primary circuit reactor. It is an inherently safe and passively cooled reactor concept. FBNR in its first phase of development is a fixed bed nuclear reactor and in its next phase will be a fluidized bed nuclear reactor. FBNR in its advanced versions can use supercritical steam or helium gas as coolant, and utilize MOX or thorium fuel.

Based an new insights into the physics of thermonuclear burn waves a novel nuclear blasting technique is proposed for the excavation of a new Isthmus canal in Central America. It allows the spatial decoupling of a small fission trigger from the main DD fusion charge in the geometrical form of a cylinder with an extremely high length-to-diameter ratio in such a way that absolutely no fission products can escape into the atmosphere. A clean nuclear blasting technique could thus be made available.