Volume 66, Issue 5-6

A new idea for the transmutation of minor actinides, long (and even short) lived fission products is presented. It is based an the property of neutron flux compression in nuclear (fast and/or thermal) reactors possessing spatially non-stationary critical masses. An advantage factor for the burn-up fluence of the elements to be transmuted in the order of magnitude of 100 and more is obtainable compared with the classical way of transmutation. Three typical examples of such transmuters (a subcritical ringreactor with a rotating reflector, a sub-critical ring reactor with a rotating spallation source, the socalled “pulsed energy amplifier”, and a fast burn-wave reactor) are presented and analysed with regard to this purpose.

Air sampling for radioxenon has been recommended as part of the verification system for a Comprehensive Nuclear-Test-Ban Treaty. Xenon isotopic activity ratios are used as a basis for confirming nuclear detonation, and discriminating between nuclear reactor accidental release and nuclear explosion. The primary purpose of this work is to use the available data on explosions, accidental reactor releases and laboratory experiments and simulations, in order to define the most probable scenarios for nuclear reactor accidental release, and for emission from underground nuclear detonations, to calculate the relevant radioxenon isotopic activity ratios in each scenario and to assess their use as event characterizing tools.

As part of a larger nuclear regulatory investigation project, a contribution to the development of methodologies for the fire hazard assessment has been provided. Within the framework of probabilistic fire risk analyses, a methodology for the selection of critical plant areas as starting point for the detailed analysis has been developed. This method is mainly based on the systematic compilation of comprehensive fire related information for each individual compartment. The primary information gathered is used for ranking of compartments in terms of fire load, conditional fire occurrence frequency, and possibilities of fire propagation. For the implementation and application of the methodology within the framework of probabilistic safety assessment studies, an ACCESS database has been developed, both for compilation of the raw data as well as for further data analysis.

Probabilistic safety assessments (PSA) have been performed for all German nuclear power plants in operation. These assessments are mainly based on the recent German PSA guide and an earlier draft, respectively. However, comparison of these PSA show differences in the results which are discussed in this paper. Lessons learned from this comparison and further development of the PSA methodology are described.

The basic parameters of the accelerator facility for a proton beam of 100 mA and 1 GeV using a superconducting separate orbit cyclotron, are given. By means of such an accelerator system one can activate a subcritical 3 GW thermal power nuclear reactor of the well known RMBK–type 1000. The system is stable against any “Chernobyl–type” power excursion. When this accelerator system will be coupled to existing RMBK–type 1000 power stations, the resulting electricity production would become substantially more safe. Associated problems and design parameters are discussed.

The simulation of the nuclear processes, occurring in media irradiated with high-energy particles, promote the elaboration of high-energy transport files. These files being used in versatile computer codes, such as MCNP, allow to simulate with high accuracy particle transport. High-energy transport files are very important for the investigation of accelerator-driven systems. The detailed description of the methods and codes used for the elaboration of the file for 209 Bi is presented in this paper.

The behaviour of neutron and delayed neutron precursors in subcritical systems is considered in time-dependent physical models at different levels of detail. Point kinetics and a simple space model in one-group diffusion theory are assumed at first, in order to be able to construct analytical solutions and to gain a full insight into the problem. More realistic calculations are then performed using a numerical multigroup diffusion code in two-dimensional cylindrical geometry. The results presented clearly show the appearance of a decoupling phenomenon between the evolutions of neutrons and delayed neutron precursors, which is enhanced by the subcriticality level of the multiplying system.

Fission gas release from the fuel is an important aspect of mechanical fuel rod design. It is required by design rules that gas release be limited. To predict the release in advance for arbitrary power conditions, a reliable computational tool must be available. A procedure based on the finite element method for evaluating the gas release is presented. The differential equation governing the diffusion process is approximated by a system of one-dimensional elements leading to small equation systems with few unknowns. These systems can rapidly be solved allowing the computation of the diffusion and gas release for any power history in short time. The results obtained with the finite element approach are compared with analytical results for constant conditions. Additionally, the gas release under varying conditions is studied and the influence of the grain size on the gas release is discussed. Finally, some methods are described that enable the estimation of gas release based on the temperature histories of grain or fuel pellet centre.

The ET-RR-1 research reactor has been operated at 2 MW since 1961 using EK-10 fuel elements with 10% enriched uranium. The reactor has been used for nuclear applied research and isotope production. In order to upgrade the reactor power to a reasonable limit facing up-to-date uses, core conversion by a new type of fuel element available is necessary. Two fuel elements in plate type are suggested in this study to be used in the ET-RR-1 reactor core rather than the utilized ones. The first element has a dimension of 8×8×50 cm and consists of 19.7% enriched uranium, which is typical for that utilized in the ET RR-2 reactor, but with a different length. The other element is proposed with a dimension of 7×7×50 cm and has the same uranium enrichment. To accomplish safety requirements for these fuel elements, thermal-hydraulic evaluation has been carried out using the PARET code. To reach a core conversion of the ET-RR-1 reactor with the above two types of fuel elements, neutronic calculations have been performed using WIMSD4, DIXY2 and EREBUS codes. Some important nuclear parameters needed in the physical design of the reactor were calculated and included in this study.

Time-dependent, one-speed neutron transport equations with strong forward and backward scattering together with isotropic scattering are studied in homogeneous slabs and spheres. First, a simple formal equivalence between the transport equations for a critical and for a time-decaying system is established. Then, the transport equation is converted into a more conventional one. The FN method of solving the resulting transport equation is applied to the calculation of the critical parameters and decay constants for the fundamental mode of the flux distribution and one-speed neutrons in spheres and infinite slabs. Numerical results are given for a number of significant figures and compared with those already available in the literature.