[en] Objectives, means and results of Swedish light water reactor safety research during the 1970s are reviewed. The expenditure is about 40 Million Swkr per year excluding industry. Large efforts have been devoted to experimental studies of loss of coolant accidents. Large scale containment response tests for simulated pipe breaks were carried out at the Marviken facility. At Studsvik a method for testing fuel during fast power changes has been developed. Stress corrosion, crack growth and the effect of irradiation on the strength ductility of Zircaloy tube was studied. A method for determining the fracture toughness of pressure vessel steel was developed and it was shown that the fracture toughness was lower than earlier assumed. The release of fission products to reactor water was studied and so was the release, transport and removal of airborne radioactive matter for Swedish BWRs and PWRs. Test methods for iodine filter systems were developed. A system for continuous monitoring of radioactive noble gas stack release was developed for the Ringhals plant. Attention was drawn to the importance of the human factor for reactor safety. Probabilistic methods for risk analysis were applied to the Barsebaeck 2 and Forsmark 3 boiling water reactors. Procedures and working conditions for operator personnel were investigated. (GBn)

[en] Swedish calculations on heavy water lattices are based essentially on a four-factor two-group model. The present standard scheme, BURNUP 5, calculates cell parameters and two-group constants as a function of irradiation. It is used in combination with two-group, one- or two-dimensional diffusion theory codes to predict reactivity, reactivity coefficients, burnup and other data of interest to reactor design. A comparison of measured and calculated integral data has been made for a large number of lattices, including uranium oxide clustered fuel at various lattice pitch, pin pitch, coolant void content and moderator temperature. The overall agreement is shown to be good, in general within 1% in k_eff, although systematic deviation seems to occur for low volume ratios and high moderator temperatures. The situation for slightly enriched and for irradiated fuel is uncertain due to lack of experimental data. Special investigations have been made of void effects, the shielding factor for resonance absorption in clusters and spectrum hardening due to selective absorption in the fuel, including comparisons with experiments. The Monte Carlo method has been used for studying fast and thermal spectrum effects in heterogeneous media. A two-dimensional fuel depletion code capable of handling intermittent fuelling schemes of practical interest has been developed for the MERCURY computer. For the treatment of non-uniform lattices a series of heterogeneous codes based on source and sink theory have been developed. Calculated and measured static control rod effects are shown to be in acceptable agreement although the experimental information is very limited. (author). 67 refs, 17 figs, 6 tabs

[en] A new formula for the fast fission factor is derived, which takes proper account to fast capture. The fission neutron spectrum is divided into two groups with constant fission cross section in one group and zero fission cross section in the other. The average total, elastic, inelastic and capture cross sections in the two groups are calculated. Different assumptions regarding anisotropic and inelastic scattering are investigated. The effects of backscattering from the moderator and fast fission in neighbouring fuel elements are pointed out. Formulas for the fast fission ratio and for the fast conversion ratio are derived. The calculated fast fission ratios are compared with experimental values. Curves are given for the fast fission factor in uranium metal and uranium oxide

[en] An infinitely long circular cylinder of radius a is surrounded by an infinite moderator. Both media are non-capturing. The cylinder emits neutrons of age zero with a constant source density of S. We assume that the ratios of the slowing-down powers and of the diffusion constants are independent of the neutron energy. The slowing-down density is calculated for two cases, a) when the slowing-down power of the cylinder medium is very small, and b) when the cylinder medium is identical with the moderator. The ratios of the slowing-down density at the age τ and the source density in the two cases are called ψ_V, and ψ_M respectively. ψ_V and ψ_M are functions of y=a²/4τ. These two functions (ψ_V and ψ_M) are calculated and tabulated for y = 0-0.25

[en] The mutual shadowing of fuel lumps in a moderator is studied assuming that the source density of resonance neutrons is constant in the moderator, that one collision with a moderator atom removes a resonance neutron from the resonance interval and that the fuel is black to resonance neutrons. Formulae for the shadowing factor are given for square or circular tubes, parallel plates of finite or infinite width and parallel circular cylinders, and some numerical results are presented

[en] A brief summary of Swedish research and development work on the utilization of plutonium is presented. As a background, plutonium production from Swedish reactors in the years 1970 - 1980 has been estimated, based on an actual forecast for installed nuclear capacity during the same period. (author)

[en] This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of pressurized water reactors and boiling water reactors. The main focus of the work is on the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are described and their implications for light water reactor safety discussed. The book contains 178 figures and 92 tables as well as cross-references and an index

[en] This article reviews principles and practices for reactor safety, radiation protection, and waste management in Sweden. The legislative framework and organizational structure are briefly described. The application of safety concepts to the design and operation of nuclear facilities is examined. Operating experiences and plant modifications for improving safety are summarized. Current safety issues, including probabilistic safety studies, filtered containment venting, and concepts for final waste disposal are reviewed. The lines of safety research are indicated, and some major projects are highlighted. Finally, a novel concept of an intrinsically safe light-water reactor which has been studied in Sweden for some time, is referred to

[en] In a comparison between pressure tube and pressure vessel type reactors for pressurized D₂O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D₂O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

[en] The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)