[en] The PANTHER code calculates steady-state or time-dependent power distribution in a reactor with a given temperature distribution. The THERMIX-DIREKT code calculates temperature and coolant flow distributions, in steady-state or transient mode, in a system with a given power distribution. It is described how to use the combination of the general purpose modular reactor code PANTHER and the HTR thermal hydraulics code THERMIX-DIREKT. An earlier version of PANTHERMIX consisted of THERMIX-DIREKT plus 2 conversion programs. The jobs and scripts to be edited by the user were very complex in their interactions. Therefore this version of PANTHERMIX has been extended with macros that take care of all these interactions, so the interaction parts of the jobs become much less complex. 6 refs

[en] A methodology for the structural integrity assessment of irradiated flawed components is provided. The disciplines required for the assessment are fracture mechanics, applied mechanics, materials research, and experimental and analytical neutron metrology. A description of the disciplines is given including the input information and resulting output data of each discipline. The deterministic assessment procedure for fracture initiation is based on the LEFM (Linear Elastic Fracture Mechanics) K-concept (K is the stress intensity factor) and the EPFM (Elastic Plastic Fracture Mechanics) J-integral concept (J is the parameter to characterize the singular stress and strain fields around the crack tip in case of yielding) resulting for instance in a failure assessment diagram. The uncertainties in the material data, loading conditions, and neutron metrology results can be addressed by performing a probabilistic analysis. This approach enables to perform a sensitivity and uncertainty analysis. The fracture assessment methodology described in the report can be elaborated, detailed, and validated by performing a case study on flawed irradiated component material, preferably from a real structure component which has operated in a nuclear environment. 11 figs., 14 refs

[en] The research program on the title subject started in 1994 and is planned to be completed in 1998. In this period several technical and scientific aspects of recycling and transmutation are investigated in different projects. The results of the 1996 projects are summarized and described in this report. The 1996 projects concern (1) the chemistry of actinides and inert matrices to test and characterize the matrices and actinide compounds in order to develop uranium-free fissionable materials for the transmutation of actinides; (2) the transmutation of plutonium in light water reactors (LWR) to assess and increase the burnup of plutonium and to assess the safety of plutonium transmutation in LWRs; (3) the radiological consequences of different nuclear fuel cycles; (4) and a reactor physics analysis of new thorium-based reactor systems to study the possibility to reduce the amount of long-living radioactive waste materials by means of the use of thorium-based compounds in a high-temperature reactor (HTR) or accelerators. 15 figs., 6 tabs., 23 refs

[en] For three different power levels, 20, 40 and 150 MWth, the PAP-HTR has been studied. This is an HTR Module concept that has been simplified in such a way that the continuously defuelling system has been eliminated and no defuelling takes place during a period of several years. Two core heatup scenarios have been simulated. It has been shown that in all cases the maximum fuel pebble temperature remains below 1600C, the temperature above which fuel degradation would start to occur, also after the reactor has gone critical again and the power level has been stabilized by itself. Fuel and gas temperature distributions are compared as well. The maximum pebble temperature before recriticality is higher for the loss of coolant (LOCA) scenario than for the loss of flow (LOFA) case, but the equilibrium maximum temperature after recriticality turns out to be higher for the pressurized case, because of the higher equilibrium power level. The equilibrium power level is a much smaller fraction of the nominal power level for the large 150 MWth system than for the smaller systems, due to the lower rate of cooling down of the large system after initiation of the accident. Therefore the equilibrium maximum temperature stays within acceptable limits for the large system too. The effects of the use of thorium fuel on the core height and waste radiotoxicity have been compared with the case of uranium fuel. Although it is widely believed that burnt thorium fuel would be cleaner than spent uranium fuel in terms of radiotoxicity, this did not appear to be more pronounced for this reactor concept than for e.g. PWRs. The relationship of power level and energy price is obvious for this power range. The use of thorium with highly enriched uranium could bring an additional economical advantage because of the lower core height needed for the same power level as the uranium case. With thorium a higher burnup can be attained, through which fuel pebbles can be added at a slower rate. The size of the vessel and the surrounding building, which are important cost factors, could be decreased accordingly. 9 refs

[en] Measurements are presented on diffusive phonon signals in ³He-⁴He mixtures. The results are described by an absorption and scattering parameter α, determined by the effective mean free path, and an absorption rate β, as functions of the molar ³He concentration in the range X< approximately 5 x 10^-3, and of the temperature in the range 50 mK < T < 220 mK. (Auth.)

[en] A 60 MWth, 23 MWe pebble bed HTR plant with indirect Brayton cycle is proposed for the applications of heat and power cogeneration or distributed electricity generation. The reactor will be cooled by helium, whereas for the secondary cycle nitrogen is proposed as a heat carrier. Economic performance is being optimized by simplification of the nuclear part and by the exclusive use of commercially proven systems in the energy conversion part. Cogeneration and maximised electricity production will be the two applications discussed. The pebble bed reactor will be of the cartridge type, refuelled off-line only once every three years. Excess reactivity will be controlled by burnable poison in the reflector. Optimization of burnable poison distribution will be discussed for two core geometries, a cylindrical and an annular one. It is shown that in an annular core burnable poison can be distributed in such a way that the reactor will be able to operate for three years with a sufficiently small overreactivity margin. (authors)

[en] The scenario of a utility in an industrialized country starting new nuclear construction with a single PBMR reactor has been considered. To make the new construction project acceptable by government and society, a maximum effort to obtain sustainability (i.e. minimization of resource use and waste production) will have to be shown. Therefore the usual open cycle for HTR has been abandoned, and the spent fuel will be reprocessed once. The long-lived transuranic (TRU) elements Pu, Np, Am and Cm are all re-fabricated into so-called transmutation fuel elements, and loaded back into the same reactor, in our case a 110 MWe PBMR with low-enriched uranium cycle. In this study, the reactor physical prospects have been investigated: to what extent the amount of TRU could be reduced. In this way, 75% of the initial amount of TRU waste is being destructed, while the time span in which the waste is more radio-toxic than uranium ore is being reduced to one-third. Also, the amount of fresh driver fuel needed is decreases by 25%. A preliminary cost analysis has been performed as well. It shows that there is also a cost advantage of operating the reactor in Deep Burn mode in industrialized countries, where the waste storage fees charged per volume are relatively high. (authors)

[en] In the framework of the IAEA Co-ordinated Research Program (CRP) 'Evaluation of HTGR Performance' for the start-up core physics benchmark of the High Temperature Engineering Test Reactor (HTTR) two-group cross section data for a fuel compact lattice and for a two-dimensional R-Z model have been generated for comparison purposes. For this comparison, 5.2% enriched uranium was selected. Furthermore, a simplified core configuration utilising only the selected type of fuel has been analysed with both the Monte Carlo code KENO and with the diffusion theory codes BOLD VENTURE and PANTHER. With a very detailed KENO model of this simplified core, k_eff was calculated to be 1.1278±0.0005. Homogenisation of the core region was seen to increase k_eff by 0.0340 which can be attributed to streaming of neutrons in the detailed model. The difference in k_eff between the homogenised models of KENO and BOLD VENTURE amounts then only,Δk =0.0025. The PANTHER result for this core is k_eff = 1. 1251, which is in good agreement with the KENO result. The fully loaded core configuration, with a range of enrichments, has also been analysed with both KENO and BOLD VENTURE. In this case the homogenisation was seen to increase k_eff by 0.0375 (streaming effect). In BOLD VENTURE the critical state could be reached by the insertion of the control rods through adding an effective ¹⁰B density over the insertion depth while the streaming of neutrons was accounted for by adjustment of the diffusion coefficient. The generation time and the effective fraction of delayed neutrons in the critical state have been calculated to be 1.11 ms and 0.705 %, respectively. This yields a prompt decay constant at critical of 6.9 s^-1. The analysis with PANTHER resulted in a k_eff =1.1595 and a critical control rod setting of 244.5 cm compared to the detailed KENO results of: k_eff = 1.1600 and 234.5 cm, again an excellent agreement. 5 refs

[en] The Netherlands Programme to Intensify Nuclear Competence (PINK, abbreviated in Dutch) supported the technical and economical evaluation of a direct cycle High Temperature Reactor (HTR) installation for combined heat and power generation. This helium cooled, graphite moderated HTR based on the German HTR-M, is named INCOGEN (Inherently safe Nuclear COGENeration). The INCOGEN reference is a 40 MW HTR design by the US company Longmark Power International (LPI). The energy conversion system comprises a single-shaft helium turbine-compressor (2.3-1.0 MPa) directly coupled with a 16.5 MW generator, a recuperator and low-temperature (150C to 40C) heat exchangers (23 MW). Spherical fuel elements (60 mm diameter) will be added little by little, which keeps the core only marginally critical. Void core volume can accommodate added fuel for several years until defuelling. Analyses of failure scenarios (loss of coolant accident or LOCA, loss of flow accident or LOFA, anticipated transient without scram or ATWS) show no excess of maximum acceptable fuel temperature of 1600C. Scoping analyses indicate no severe graphite fires. Transient analyses of the turbine-compressor system indicate adequate control flexibility. Optimization and endurance testing of the helium turbine-compressor is recommended

[en] The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO₂-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a variety of Pu- and Pu/MA-based fuels (possibly in combination with thorium), and to obtain a significant reduction of the Pu- respectively Pu/MA content, while maintaining, to a large extent, the well-known standard (U-fuelled) HTR safety characteristics. However, this will require some changes in the reactor design. Studies have furthermore shown that fuel with a 'diluted' kernel ('inert-matrix') improves the transmutation performance of the reactor. A study on proliferation resistance, taking into account several possible proliferation pathways, highlights that a prismatic (V)HTR core would be amenable to conventional safeguards accounting and verification procedures, with fuel blocks accounted for individually in the same way as LWR fuel assemblies. However, a modified approach would be needed in pebble bed cores because of the impracticability of accounting for individual fuel spheres. When dealing with minor actinide bearing fuel helium generation is an important issue. Experiments have shown that He will be released from the kernel, but not from fresh kernels. Indeed, fresh fuel has shown a remarkable stability up to 2500 degrees C. For modelling purposes, 100% release of helium from the kernel is justified. The diluted kernel concept was first invoked by Belgonucleaire brings many benefits. The fuel modelling studies have clearly indicated the advantages that can be gained by dilution. Essentially, for a given buffer layer thickness, more volume is available to accommodate the CO and He released. Chemical thermodynamic models have been deployed to design a kernel that will show limited CO production. The most important point here is that substoichiometric Pu or Am oxides are essential. Further improvement can be achieved by chemical buffering of the fuel by the addition of Ce sesquioxide, which takes up oxygen to form the dioxide. Ultimately any coated particle design must be validated in an irradiation test. Though not possible to perform an irradiation programme in the PUMA project, the feasibility of such a programme has been demonstrated, and the initial data needed to launch such a test has been generated. Pu/MA transmuters are envisaged to operate in a global system of various reactor systems and fuel cycle facilities. Fuel cycle studies have been performed to study the symbiosis to other reactor types/systems, and to quantify waste streams and radio toxic inventories. This includes studies of symbiosis of HTR, Light Water Reactor (LWR) and Fast Reactor (FR) systems, as well as the assessment of the technical, economic, environmental and socio-political impact. It is e.g. shown that a Pu/MA-loaded HTR may have a considerable, positive impact on the reduction of the amount of TRU in disposed spent fuel and high level waste.