[en] Numerous publications are cited, some with abstracts, on the subjects of ground water motion, neutron transport, and reactor physics. (E.C.B.)

No abstract available

No abstract available

[en] Formulae are presented for calculating the radial distributions and averages of the fluxes and currents of radiation on a plane, circular aperture due to direct contributions from a plane, uniformly distributed source of radiation of circular shape located co-axially with the aperture. The source emission is either isotropic or of the cosine type. Limiting forms of the fluxes and currents, the mean solid angle of emission from source to aperture and certain identities that arise are also discussed. Some graphical illustrations are given

No abstract available

[en] Results are presented of neutronic evaluations of activations induced in foils irradiated at various positions within the INRS irradiation facility; comparisons with the results of measurements are made. In the measurements, DT-generated neutrons resulting from accelerated deuterium (60 keV) striking a tritiated target located at the centre of the facility were used. The activities considered were those induced via the ²⁷Al(n,p)²⁷Mg, ²⁷Al(n,a)²⁴Na, ⁶³Cu(n,2n)⁶²Cu, ¹²⁷I(n,2n)¹²⁶I, ⁹⁰Zr(n,2n)⁸⁹Zr, sup(115)In(n,γ)sup(116m)In, ¹⁹⁷Au(n,γ)¹⁹⁸Au, ²³²Th(n,γ)²³³Th and ²³²Th(n,f)⁸⁹Rb (F.P.) reactions. The calculations were made using the MORSE Monte Carlo neutronics code in the 25-neutron-group approximation for determining the fluxes, and a specially prepared 25-neutron-group ΣactivitiesΣ library of reaction cross sections for determining the resultant reaction rates

[en] A vessel of type 316L stainless steel will be used for the long-term storage of tritium. The tritium may be gas, or fixed as a metal tritide. We have assembled the data and equations needed to estimate the efficiency of the containment. If the vessel wall is 6 mm thick, then at the design storage temperature, < 348 K, negligible tritium will be released by permeation through the vessel wall. In hypothetical accidents, the vessel and contents will be heated 2-12 h to temperatures up to 1073 K. Permeation of tritium through the vessel wall is very rapid at this temperature, although heating to 573 K would be of little concern. Techniques for estimating tritium loss from the vessel over the complete range of conditions envisaged are listed in the report

[en] The role of collision/escape probabilities in establishing the prompt response of bismuth germanate and NaI(Tl) scintillators is investigated. A possible approximation to the collision probability is discussed. Exact expressions for the probability of escape of radiation from finite, right circular cylindrical and rectangular parallelepiped-shaped detectors are derived from an exponential-like spatial distribution of the source in the detector medium and a parametrized angular distribution of emission of the source. Fairly extensive tables of collision probabilities calculated for these bodies are given. For purposes of comparison some results obtained previously in a Monte Carlo calculation are also presented and discussed

[en] Results are presented of evaluated neutron fluxes, and neutron induced activations of isotopes of Al, Fe, Co, Zr, Cu, Th, U and Pu in various graphite-thorium (metal) assemblies. The systems considered are spherically symmetric with the 14 MeV, DT neutron sources for the irradiation located at the system centre. The systems are idealizations of material arrangements in proposed experimental activation-analysis studies of graphite-thorium blankets. (author)

[en] Results of neutronic evaluations of fissile fuel breeding in a variety of DT fusion hybrid-reactor blankets are presented. The blankets are of the fast-fission or fission-suppressed rather than fission-enhanced designs, i.e. in the blankets considered emphasis is on fissile fuel rather than power production. For ²³³U breeding, when Li metal is the coolant for the first wall and the graphite moderator and the tritium breeding constituent of the blanket, the number of atoms of ²³³U produced per fusion in blankets that could be of practical interest is in the range 0.5 - 0.68, with the lower value applying to water-cooled ThO₂ fertile fuel, the upper to gas-cooled Th-metal fuel located next to the reactor first wall. Neutron multipliers like Pb or Be can increase the production to about 0.74. For ²³⁹Pu breeding, the production ratio in practical blankets is 0.6 - 1.64, with the best results being for gas, Na- or Li-metal-cooled U-metal fuels located adjacent to the first wall (the U is depleted uranium). Gas-cooled U-Th-metal blankets, optimized for ²³³U breeding, yield 0.76 atoms of ²³³U and 0.38 atoms of ²³⁹Pu. The blanket energy multiplication factors are in the range 1.6 - 2.5 for Th blankets, 2.5 - 9.0 for U blankets and approximately 5.5 for the U-Th-metal blanket. The tritium breeding ratio in all blankets is 1.075. Blankets with other first wall, coolant and tritium breeding constituents are also considered. The fusion power requirements of hybrids that could supply the fuel needs of thorium-burning CANDU power reactors, and the allowed costs for building the hybrids are indicated