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.     

Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16–20 November 2003]. The molten salt fuel is a ternary NaF–LiF–BeF₂ system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF₃, etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP’ 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as to describe the molten salt reactors. For the adaptation to molten salt reactor, a complete equation of state (EOS) for this liquid fuel had to be developed and implemented into the SIMMER-III code. Through those simulations it was concluded that the thermal hydraulic behaviour appeared to be very important in molten salt reactors concerning design, operation and safety. A flow distribution plate design was found effective to optimize the flow pattern in the core region. Further investigations are under way to obtain optimal flow fields without exceeding design limits.     

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.     

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.     

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.     

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.     

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.     

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.     

Experimental investigation of solid–liquid mixtures freezing behavior in flow channels

The penetration and freezing of hot-core material mixtures through flow channels during core disruptive accidents (CDAs) within a sodium-cooled fast reactor is one of the major concerns confronting safety designers of the next-generation reactors. The main objective of this study is to investigate those fundamental characteristics of penetration and solidification involved in channeling molten metal and solid particle mixtures over cold structures. In this study, a low-melting-point alloy (viz., Bi–Sn–In alloy) and mixtures with solid particles (of copper and bronze) were used as a simulant melt, while L-shape metal (of stainless steel and brass) and stainless steel fuel pin bundle were used as cooling structures. Two series of basic experiments were performed to study the effect solid particles have on penetration and cooling behavior under various thermal conditions of melt by varying solid particle volume fraction, structure temperature and structural geometry. Melt flows and distributions were recorded using a digital video camera and subsequently analyzed. The melt penetration length into the flow channel and the proportion of melt adhesion on structural surfaces were measured. Results indicate that penetration length becomes shorter for molten-metal/solid particle mixtures (mixed melts) compared with pure molten metal (pure melt) as well as decreases with increasing solid particles volume fraction of the melt. The present study will contribute to a better understanding of the basic thermal-hydraulic phenomena of melt freezing in the presence of solid particles and to provide an experimental database for validation and improvement of the models of fast reactor safety analysis codes.