[en] A finite element stress analysis of the coil under consideration is unavoidable due to the proximity of the stresses to the 0.2% yield strength. The casing of this D-shaped coil is both a hollow body and a looped one. This leads to computing costs and memory requirements which are enormous and preclude any parametric study. To reduce computer costs a newly developed code, FLASH, has been used. It has a hybrid stress model leading to more rapid converge and thick plate elements which allow bending moments to be computed. Only one thick plate is needed across the thickness of the casing and local stress concentrations are obtained from the mean stress and the bending moment. Several models were developed most of which can be set up automatically. Comparisons between the models and with ASKA finite element results from BROWN BOVERI Co. essentially show agreement. The casing of individual conductors has also been investigated with the same code. Both the effect of the Lorentz forces and those arising from the quench pressure due to helium heating on loss of superconductivity have been considered. (Auth.)

[en] The Next European Torus (NET) is foreseen as the next step in the European development towards the controlled use of thermonuclear fusion. Detail design of the shielding blanket protecting the peripherals, more especially the super-conducting coils, is well advanced. A cross-section uncertainty, i.e. a study of the expected inaccuracy due to the nuclear cross-section data, has been done for the neutron-gamma reactions in the insulation of the coils for such a design. As an extension of previous work on the NET shielding blanket (e.g. MCNP calculations), it was deemed necessary to estimate the accuracy attainable with transport codes in view of the uncertainties in microscopic cross-sections. The code used, SENSIBL, is based on perturbation theory and uses covariance files, COVFILS-2, for the cross-section data. This necessitates forward and adjoint flux calculations with a transport code (e.g. ONEDANT, TRISM) and folding the information contained in these coupled fluxes with the accuracy estimates of the evaluators of the ENDF/B-V files. Transport, P₅S₁₂, calculations were done with the ONEDANT code, for a shielding blanket design with 714 MW plasma fusion power. Several runs were done to obtain well converged forward and adjoint fluxes (ca. 1%). The forward and adjoint integral responses agree to 2%, which is consistent with the above accuracy. The n-γ response was chosen as it is typical of the general accuracy and is available for all materials considered. The present version of SENSIBL allows direct use of the geometric files of ONEDANT (or TRISM) which simplifies the input. Covariance data is not available at present in COVFILS-2 for all of the materials considered. Only H, C, N, O, Al, Si, Fe, Ni, and Pb could be considered, the big absentee being copper. The resulting uncertainty for the neutron-gamma reactions in the insulation of the coil was found to be 17%. Simulating copper by aluminium produces a negligible increase in the uncertainty, mainly because the copper is not in a region of large fast flux. (author) 1 fig., 5 tabs., 16 refs

[en] The Next European Torus (NET) is foreseen as the next step in the European development towards the controlled use of thermonuclear fusion. Detail design of the shielding blanket protecting the peripherals, more especially the super-conducting coils, is well advanced. A cross-section uncertainty study, i.e. a study of the expected inaccuracy due to the nuclear cross-section data, has been done for the neutron-gamma reactions in the insulation of the coils for such a design. The resulting uncertainty for the neutron-gamma reactions in the insulation of the coil was found to be 17%. Simulating copper by aluminium produces a negligible increase in the uncertainty, mainly because the copper is not in a region of large fast flux. (author) 1 tab., 15 refs

No abstract available

[en] To corroborate 1-D deterministic shielding calculations on the Next European Torus (NET) vacuum vessel/shield and shielding blanket, 3-D Monte Carlo calculations have been done with the MCNP code. This should provide information on the poloidal and the toroidal variations. Plasma source simulation and the geometrical model are described, as are other assumptions. The calculations are based on the extended plasma power of 714 MW. The results reported here are the heat deposition in various parts of the device, on the one hand, and the neutron and photon currents at the outer boundary of the vacuum vessel, on the other hand. The latter are needed for the detailed design of the super-conducting magnetic coils. A reasonable statistics has been obtained on the outboard side of the torus, though this cannot be said for the inboard side. The inboard is, however, much more toroidally symmetric than the outboard, so that other methods could be applied such as 2-D deterministic calculations, for instance. (author) 4 refs., 44 figs., 42 tabs

[en] The Next European Torus (NET) is foreseen as the next step in the European development towards the controlled use of thermonuclear fusion. It has essentially two goals: the study of plasma physics with a long burning, ignited D-T plasma and the technological development of blankets, materials, tritium breeding, etc. The neutron load is already considerable, needing detail design of the shielding blanket protecting the peripherals, more especially the super-conducting coils. Monte-Carlo shielding calculations have been done with the MCNP code for the heat generation in the pressure vessel and for the radiation load on the outside. A very detailed 3-D geometrical model has been set up to obtain both the neutron and gamma loads on the coil insulation, as this influences both the cooling power and the life time of the coils. At present only shielding is considered, but the model is a template for later tritium breeding modules. It includes detailed representations of both the plasma as an extended neutron source and of the large vacuum pumping ducts. Preliminary results for a plasma power of 714 MW, show that the total heat deposited in the blanket and vessel is 810 MW, considerably more than the energy of the neutrons at birth. It corresponds to 20.0 MeV per source neutron. This high value is due to the preponderance of stainless steel and absence of lithium. The maximum currents escaping from the vessel are 3 x 10¹⁰ neutrons/cm² sec and nearly 9 x 10⁹ photons/cm² sec. (author). 2 refs.; 4 figs.; 2 tabs

[en] The response of an LMFBR roof cover to HCDA loadings is examined using a combined 2D/3D modelling approach. A generic 3D roof design of box-type construction is adopted and analysis under specimen loads carried out using the finite element program ADINA. The reactor tank and all internal components below roof level are assumed axisymmetric with the containment code SEURBNUK-EURDYN employed to follow the accident progression. An interface between SEURBNUK-EURDYN and ADINA is provided via a 2D simulant roof model, chosen to match the principal response characteristics of the 3D roof, to enable any interaction effects occurring during impact to be assessed. (author)

[en] A finite element stress analysis of the coil under consideration is unavoidable due to the proximity of the stresses to the 0.2% yield strength. The casing of this D-shaped coil is both a hollow body and a looped one. This leads to computing costs and memory requirements which are enormous and preclude any parametric study. To reduce computer costs a newly developed code, FLASH, has been used. It has a hybrid stress model leading to more rapid convergence and thick plate elements which allow bending moments to be computed. Only one thick plate is needed across the thickness of the casing and local stress concentrations are obtained from the mean stress and the bending moment. Several models were developed most of which can be set up automatically. Comparisons between the models and with ASKA finite element results from BROWN BOVERI Co. essentially show agreement. Parametric variations were done to study the effects of boundary conditions. The casing of individual conductors has also been investigated with the same code. Both the effect of the Lorentz forces and those arising from the quench pressure due to helium heating on loss of superconductivity have been considered. (author)

[en] A hybrid blanket for a reversed-field pinch (RFP) reactor is presented for breeding ²³³U from thorium. This study focuses on the shellblanket design and assumes a plasma like that used by Culham Laboratory (CL) in its designs. The ²³³U bred helps the important neutron economy and allows the tritium breeding ratio to be 0.96 at beginning of life for a mean of 1.06 for a 9 MW . yrm² burnup. A thick conducting shell is assumed for discharge stability and field reversal. This need for a good conductor requires that only pure copper or aluminum or alloys thereof be used. Two designs were investigated, one with a pure copper first wallshell, the other with an aluminum alloy. In these designs 3.4% of the thorium is converted to ²³³U. In both designs the low metallurgic temperature limit means the large amount of power deposited on and in the shell is not attractive thermodynamically. The resulting large temperature differences in the shell cause high mechanical stresses. The design as it stands is not feasible from the point of view of radiation damage to materials

[en] The Swiss Heating Reactor (SHR) is a 10-50 MW(th), low pressure-low temperature, boiling water type reactor. Its concept is based 1. on the characteristics of the heat market in Central Europe and 2. on the typical community sizes in this region, representative of the heat consumers. The SHR is designed to operate safely without permanent presence of shift personnel and to be built right within urban areas. Design goals like these have called for a relatively simple and extremely reliable reactor and plant, characterized by 'fail-safe' systems and 'passive' processes. These are now implemented in the SHR concept. The safety assessment for the SHR was performed in a dual manner, using the deterministic and the probabilistic methods. This paper gives certain details of these safety and radiation protection evaluations and summarizes the results of the corresponding work. (orig.)