HTGR Reactor Physics and Fuel Cycle Studies

The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the HTGR_n project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides. These studies were mainly concerned with the investigation and intercomparison of the plutonium and actinide burning capabilities of a number of HTGR concepts and associated fuel cycles, with emphasis on the use of civil plutonium from spent LWR uranium fuel (first generation Pu) and from spent LWR MOX fuel (second generation Pu).JRC.E.4-Nuclear fuel

HTGR reactor physics and fuel cycle studies

The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the "HTR-N" project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides.These studies were mainly concerned with the investigation and intercomparison of the plutonium and actinide burning capabilities of a number of HTGR concepts and associated fuel cycles, with emphasis on the use of civil plutonium from spent LWR uranium fuel (first generation Pu) and from spent LWR MOX fuel (second generation Pu). Besides, the "HTR-N" project also included activities concerning the validation of computational tools and the qualification of models. Indeed, it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transinutation without multi-reprocessing of the discharged fuel.These validation and qualification activities have been centred round the two HTGR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3D diffusion theory codes yield acceptable results. Special attention, however, has to be given to the modelling of neutron streaming effects. For the HTR-10 the analyses focused on first criticality, temperature coefficients and control rod worth. Also in these studies a good correspondence between calculation and experiment is observed for the 3D diffusion theory codes. (c) 2006 Elsevier B.V. All rights reserved

Direkte Endlagerung ausgedienter Brennelemente DEAB. Aktives Handhabungsexperiment mit Neutronenquellen. Berechnung der Ortsdosisleistung und Neutronenspektren vom POLLUX-Behaelter, Einzelabschirmbehaelter und Versuchsbehaelter in einem Salzbergwerk (TA 1)

Zielsetzung des aktiven Handhabungsexperimentes war, die radiologischen Aspekte aufgrund Neutronenrueckstreuung zu untersuchen, die bei der Handhabung von Brennelementen bzw. hochaktivem Abfall in den Strecken eines untertaegigen Endlagers in Salzformation auftreten. Da Neutronen eine wichtige Rolle bei der Option der direkten Endlagerung spielen, wurde der Schwerpunkt der Untersuchung auf ihren Beitrag zu der totalen Aequivalenzdosisleistung gelegt. Die Aequivalenzdosisleistungen der von direkten und an Salzgestein gestreuten Neutronen bei Handhabung von Pollux-Behaeltern und Einzelabschirmbehaeltern wurde durch einen Versuchsbehaelter, der mit einer Cf-252-Neutronenquelle beladen war, simuliert und experimentell analysiert, um die radiologische Exposition des Betriebspersonals abschaetzen zu koennen. Die Ergebnisse des AHE-Experiments lieferten wertvolle Beitraege zum besseren Verstaendnis der radiologischen Aspekte bzgl. Neutronenrueckstreuung bei der Handhabung von ausgedienten Brennelementen bzw. hochaktivem Abfall in einem Endlager im Salz. Validierte Computerprogramme stehen nun zur Verfuegung, die eine Minimierung der Personendosen des Betriebspersonals ermoeglichen. (orig.)The objective of the Active Handling Experiment with Neutron Soruces was to investigated radiological aspects due to backscattered neutrons when spent fuel and high level waste will be handled in the drifts of an underground repository located in a salt dome. As neutrons play an important role in the direct disposal option, main emphasis was laid on their contribution to the total dose equivalent rate. The radiological exposure caused by direct neutrons and neutrons scattered by the salt rock during handling processes of POLLUX packages and shipping packages was simulated and experimentally analysed with the help of the AHE package. The AHE shielding cask was filled with one Cf-252 line source in order to simulate the neutron fields of a loaded POLLUX cask and a shipping cask loaded with a fuel element canister with respect to neutron energy distributions and local dose equivalent rates. The results of the AHE experiment provided valuable contributions for a better understanding of radiological aspects during handling spent fuel/high level waste in an underground repository. Validated computer codes are available now which make it possible to minimize the occupational dose. (orig.)Available from TIB Hannover: RO 6452(67)+a / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekSIGLEBundesministerium fuer Bildung, Wissenschaft, Forschung und Technologie, Bonn (Germany); European Union (Euro), Brussels (Belgium)DEGerman