Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O_2−x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of “best estimates” provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.     

First 3-D calculation of core disruptive accident in a large-scale sodium-cooled fast reactor

The SIMMER-IV computer code is a three-dimensional fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. The present study has attempted the first application of SIMMER-IV to a core disruptive accident in a large-scale sodium-cooled fast reactor. A principal point of this study was to investigate reactivity effects with fuel relocation under three-dimensional core representation including control rods. The calculation has indicated that the fuel discharge from the core was disturbed by a significant flow resistance at the entrance nozzle in the current design. Additional static neutronic calculations have been performed to compare basic neutronic characteristics between different scale cores. The static neutronic calculations have clarified that the outward fuel compaction within the inner core increased the reactivity in the large-scale core unlike the small-scale core.     

Mitigation of core disruptive accident energetics in burner cores

As part of the CAPRA Program (Consommation Accrue de Plutonium dans les RApides) the feasibility of fast reactors is investigated to burn plutonium and also to destruct minor actinides. The design of CAPRA cores shows significant differences compared to conventional cores. Especially the high Pu-enrichment increases the recriticality risk and the associated energetics levels of secondary excursions. Other features of the core have the potential to mitigate this risk again. Of special importance are the numerous diluents in the core which might both prevent coherent liquid fuel compactive motions and can also be used as dedicated fuel discharge paths. The early release of fuel could prevent the escalation to large whole core pools with their energetics potentials.     

Accident analyses and passive measures reducing the consequences of a core-melt in CAPRA/CADRA reactor cores

As part of the Combustion Améliorée du Plutonium dans les Réacteurs Avancés/Consommation D'Actinides et de Déchets dans les Réacteurs Avancés (CAPRA/CADRA) program the feasibility of reactor systems with different neutron spectra and coolants is investigated to burn plutonium and also to destruct minor actinides and long lived fission products. In this paper, we deal with reactor cores with fast spectrum and metal cooling. The design of this type of CAPRA/CADRA cores shows significant differences compared e.g. to conventional fast reactor cores. The high Pu-enrichment and the high minor actinide load have an important influence on the core meltdown behavior and the associated recriticality risk. To cope with this risk, inherent design features and special measures/devices are investigated for their potential of early fuel discharge to reduce the criticality of the reactor core. An assessment of such measures/devices, which could provide an additional line of defense against severe accident development, is given. Within the CAPRA/CADRA program, also accelerator driven subcritical systems are investigated for performing the task of transmutation and incineration. In these fast neutron systems with a strong external neutron source, the kinetic behavior is different to a critical core and new strategies and measures for accident prevention have to be investigated.     

Transient analyses for accelerator driven system PDS-XADS using the extended SIMMER-III code

Transient analyses for Preliminary Design Studies of an Experimental Accelerator Driven System (PDS-XADS) were performed with the reactor safety analysis code SIMMER-III, which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors so as to describe the XADS specifics such as subcritical core, strong external neutron source and lead–bismuth–eutectic (LBE) coolant. As transient scenarios, the following cases were analyzed in accordance with the PDS-XADS program: spurious beam trip (BT), unprotected beam overpower (UBOP), unprotected transient overpower (UTOP), unprotected loss of flow (ULOF) and unprotected blockage (UBL) in a single fuel assembly. In addition, to cover some core-melt situations and investigate the potential for recriticalities, so-called snap-shot analyses with ad hoc postulated severe blockage conditions were also investigated.The simulation results for BT and UBOP showed that immediate fuel damage might not take place under short-time beam interruption or a 100% increase of the external neutron source. Concerning UTOP, it was found that a reactivity jump of 1 $ would not lead to damage of the fuel and the cladding. The ULOF simulation showed that the remaining natural convection of the coolant would prevent the cladding from disruptions. In the simulation of UBL in a single fuel assembly, it was shown that no cladding failure might be expected, due to the radial heat transfer and the coolant flow in the hexcan gap. Under an artificial suppression of the radial heat transfer for this UBL case, a pin failure occurred in the simulation but subsequent fuel sweep-out into the upper plenum region would bring a reactivity reduction and no power excursion. The severe accident simulations starting from postulated blockage above an already disrupted core showed that a severe recriticality could be avoided by the fuel sweep-out into the dummy-assembly or hexcan gap regions.The present simulation results showed that the current PDS-XADS design has a remarkable resistance against severe transient scenarios even in core-degradation conditions.     

Extension of the SIMMER cross-section processing scheme for core degradation analyses of thermal systems

The SIMMER multi-physics code system was initially developed for safety analyses of liquid metal-cooled fast reactors. For these reactors material homogenization was considered to be an acceptable approximation in the neutronics part of the code. In order to increase the application range of SIMMER for analyses of transient phenomena in thermal reactors, the SIMMER cross-section processing scheme and the related data libraries have been extended. New options allow taking into account heterogeneity effects in e.g. water-cooled systems with pin-type or plate-type fuel, where these effects may play an important role. By now, the new extension was checked for intact core geometries. Accident initiators in light water reactors, e.g. reactivity induced accidents (RIAs), may lead to severe core degradation. Therefore, the fuel sub-assembly geometry may dramatically change during an accidental transient. In this paper we investigate the capability of the extended SIMMER cross-section processing scheme to take into account heterogeneity effects for distorted fuel sub-assembly configurations in water-cooled systems. The new SIMMER option improves the accuracy of reactivity calculations if parameters for heterogeneity treatment are evaluated and implemented for degraded configurations in the cross-section processing scheme. It is shown in the paper that only a few distorted configurations may be considered and the parameters can be obtained by interpolation in intermediate cases.     

Effect of surgical technique on meniscus transplants. A histological, animal experiment study

Heparan sulphate binding to cells of the gastric pathogen Helicobacter pylori at pH 4-6 is common. Binding was inhibited by various unlabelled sulphated polysaccharides and at high ionic strength and pH, but not by carboxylated or non-sulphated compounds. The inhibition by various sulphated compounds such as dextran sulphate and carrageenans was related to the sulphate content and not to the carbohydrate polymer backbone. The IC50 values for heparin and dextran sulphate for H. pylori strain 25 were calculated as 3.55 x 10(-7) M and 5.01 x 10(-6) M respectively. Heparin-binding proteins of H. pylori are exposed on the cell surface, as shown by biotinylation of cell-surface proteins before separation of outer membranes and by an indirect immunofluorescence assay. The strongest biotin-heparin binding by H. pylori was observed with a polypeptide in the 55-60 kDa region.     

Heterogeneous cores for improved safety performance: A case study: The supercritical water fast reactor

Light water reactor (LWR) technology is nowadays the most successful commercial application of fission reactors for the production of electricity. However, in the next few years, nuclear industry will have to face new and demanding challenges: the need for sustainable and cheap sources of energy, the need for public acceptance, the need for even higher safety standards, the need to minimize the waste production are only a few examples. It is for these very reasons that a few next generation nuclear reactor concepts were selected for extensive research and development; super critical water reactors are among them. The use of a supercritical coolant would allow for higher thermal efficiencies and a more compact plant design, since steam generators, or steam separators and driers would not be needed, hence achieving a better economy. Moreover, because of the high heat capacity of supercritical water, relatively less coolant would be needed to refrigerate the reactor, therefore the feasibility to design a water cooled fast reactor: the supercritical water fast reactor (SCFR). This system presents unique features combining well-known fast and light water reactor characteristics in one design (e.g. a tendency to a positive void reactivity coefficient together with loss of coolant accident – LOCAs as a design basis accident). The core is in fact loaded with highly enriched MOX fuel (average plutonium content of 23%), and presents a peculiar and significant geometrical and material heterogeneity (use of radial and axial blankets, solid moderator layers, 12 different enrichment zones). The safety analysis of this very complex core layout, together with the optimization of the void reactivity effect through core design, is the main objective of this work.     

Evaluation of a steam generator tube rupture accident in an accelerator driven system with lead cooling

A postulated steam generator tube rupture (SGTR) accident in a lead cooled accelerator driven transmuter (ADT) is investigated. The design of the ADT without intermediate loops bears the risk of water/steam blasting into the primary coolant. As a consequence a nuclear power excursion could be triggered by steam ingress into the ADT core which has a significant positive void worth. A thermal coolant–coolant interaction (CCI) might initiate a local core voiding too and additionally could lead to sloshing of the lead pool with mechanical impact of the heavy liquid on structures. The steam formation will also lead to a pressurization of the cover gas. The problems related to an SGTR are identified and investigated with the SIMMER-III accident code.