[en] Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

[en] One of the problems in assessing the radiological impact of uranium milling is to determine the rates of release to the air of material from the various sources of radioactivity. Such source terms are required for modelling the transport of radioactive material in the atmosphere. Activity arises from various point and area sources in the mill itself and from the mill tailings. The state of the tailings changes in time from slurry to solid. A layer of water may be maintained over the solids during the life of the mine, and the tailings may be covered with inert material on abandonment. Releases may be both gaseous and particulate. This paper indicates ways in which radon emanation and the suspension of long-lived particulate activity might be quantified, and areas requiring further exploration are identified

[en] Expanding the spent fuel storage pool capacity is a good solution for utilities facing the current shortage in fuel reprocessing capacity. This paper reviews the problems more likely to be found when expanding a spent fuel storage facility by using high density storage racks. Basically three types of problem arise: problems related with the characteristics of the new facility, problems related with the works of expansion, and problems related with the long term storage of large quantities of spent fuel

[en] Argonne National Laboratory has been using the System 2000 data base management system for the past two years. It has been used for technical as well as administrative applications. This paper describes some of the experience gained relating to advantages and disadvantages of data base management systems as well as of System 2000 in particular

[en] This paper describes a planned central spent fuel storage facility in Sweden. The nuclear power program and quantities of spent fuel generated in Sweden is discussed. A general description of the facility is given with emphasis on the lay-out of the buildings, transport casks and fuel handling. Finally a possible design of a Swedish transportation system is discussed

[en] Rio Algom Limited has used barium chloride treatment since 1965 at Elliot Lake to remove radium from effluents. The current system, however, is limited for the removal of suspended or particulate radium-226 and the recovery from the settling ponds of the barium/radium coprecipitate. A research program aimed at removing dissolved radium-226 and also at settling the resultant precipitates with flocculents has as its goal a final effluent of 3 pCi/L dissolved and 10 pCi/L total radium-226. Large scale laboratory tests and field investigations have shown the stringent nature of these targets

[en] The design of an independent spent fuel storage facility with a capacity of 3000 tons of fuel is currently in progress in Sweden. The fuel will be stored in a mined rock cavity. In order to perform a preliminary safety analysis, a number of safety aspects on this specific spent fuel storage facility has been identified and design criteria have been proposed. The location of the fuel storage pools in a rock cavity is to our knowledge unique to the Swedish facility and this specific design feature is reflected in the design criteria. The accident analysis shows that design basis accidents yield very low off-site radiation exposures. A long term total loss of cooling of the stored fuel will have an extremely low probability. An estimation of the maximum fuel temperature in this case shows that the consequences are tolerable

[en] The use of neutron interactions in borehole logging shows considerable promise for fulfilling the role of an in-situ method of elemental analysis. A fundamental problem is related to variability of the rock formation which affects both the neutron spatial and energy distributions. Computations, using Monte Carlo and analytical methods, have been employed in order to make a general appraisal of neutron and gamma transport in earth media. The programme has highlighted requirements for data which have special importance in the interpretation of measurements with neutron sondes for borehole logging

[en] The Griffin-Williams exciton model is used to calculate the preequilibrium contribution to the (n,2n) reaction around 14.5 MeV neutron energy for nuclei throughout the periodic table. The experimental cross sections for 60< A<209 are explained with an r.m.s. deviation of 0.31 by including a statistical evaporation and a preequilibrium component taking into account the competing proton emission. For A<60 the data is not reproduced very well

[en] Drill stem tests were performed to depths of 500 metres during borehole advancement in crystalline rock using a single packer system. Apparent equilibrium shut-in pressures were obtained to provide an estimate of the vertical hydraulic potentials. A downhole pneumatic piston was used to create an instantaneous pressure pulse in the isolated test section and the transient pressure decay was monitored. Equivalent single-fracture apertures and equivalent rock mass hydraulic conductivities, Kerm, were calculated from the transient pressure decay data. Equivalent single-fracture apertures are presented which range between 1 and 125 microns for tests performed in a 150 m deep borehole into a granitic rock mass. It can be shown from modeling studies that diffusion from a fracture into the adjacent matrix, fracture aperture and grounwater velocity through the fractures are the controlling transport parameters for non reactive solute transport through fractured media. Finite difference model results for a constant porosity system illustrating solute breakthrough curves and profiles along the length of a fracture indicate the importance of matrix diffusion in particular. If the average water velocity in a 300 μm fracture in a fractured clay or shale is 3m/d, the average velocity of a conservative solute under the simulated conditions is approximately 08m/d