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Preliminary Results on PDS-XADS Fuel Rod Analysis: AISI 316L vs. T 91 Cladding Material
No full textIn this document are reported preliminary results on PDS-XADS fuel rod performance when AISI 316L or T 91 steel as cladding material is adopted. For both options a Transuranus deterministic analysis was performed under Design Basis Conditions - Category I (Normal Operations). A custom version of Transuranus was developed by implementing the main T 91 physical properties. In particular, a cladding swelling and an irradiation creep behaviour, consistent with open literature, have been inserted. Compared to AISI 316L, T 91 cladding improves PDS-XADS fuel rod performance: safety margins on assumed design limits are increased, Fission Gas Release and clad strains are lowered. At End Of Life, a fuel-cladding radial contact is observed in T 91 option, while AISI 316L clad maintains an open-gap fuel rod geometry. This aspect, which is strongly dependent on the assumed fuel and cladding swelling modelling, may play an important role in mitigating the above statements, for this reason further investigations are required
D67 App. D - Fuel Rod Thermal and Mechanical Analysis
No full textThis report is focused on the PDS-XADS fuel rod thermal and mechanical performance analysis. The Design Basis Conditions - Category I (Normal Operations) were assumed. In particular two cases have been considered: case A (nominal conditions), case B (linear heat rate and fast neutron flux increased by 20% with respect to case A as well as irradiation time increased by 50%). Results confirm that design limits are fully respected far both cases. The Helium pre-pressurisation effect was specifically investigated. The analysis suggests that increasing the fabrication tilling gas pressure would be beneficial for the overall fueI rod performance. Further analysis has been performed to investigate the PDS-XADS fuel rod behaviour when a zero lower plenum volume is assumed. In this case a significant difference is found under conditions B at the End Ot Life, nevertheless design limits are still respected
TRANSURANUS Modelling of Irradiated Inert Matrix Fuels from Halden IFA-652 Experiment
No full textInert matrix fuels are a possible option to reduce plutonium stockpiles by burning it in LWRs. These fuels, which host plutonium in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. In last ten years ENEA has been studying innovative U-free fuels focusing its investigation on Calcia-Stabilised Zirconia (CSZ) and thoria matrices. The high resistance to radiation damage and leaching, which characterizes the designed compounds, support a once-through fuel cycle strategy, while a decrease of fuel thermal conductivity with respect to std. UOX fuel (markedly for CSZ matrix), is a main drawback for the fuels under consideration. A first inreactor testing of these innovative fuels, has been performed through the IFA-652 experiment in the OECD Halden HBWR. Irradiation conditions were similar to the Westinghouse AP-600 fuel with 45 MWd/kgU eq burnup as target. The test-rig consists of a six-rod bundle loaded with IM, IMT and T fuel. IM and T fuel are based, respectively, on CSZ and thoria matrices, the fissile phase being HEU oxide (93% 235U enriched). IMT is a ternary fuel composed by CSZ+thoria matrix and HEU oxide as fissile phase. The thoria content in IMT fuel is adjusted so as to improve the reactivity feedbacks. Pins are equipped with fuel temperature thermocouple, internal pressure gauge and fuel stack elongation sensor. In this paper, a TRANSURANUS modelling of inert matrix fuels is discussed. In particular, a comparison of the code predictions with IM IFA-652 experimental data is performed up to 40 MWd/kgUeq average discharge burnup. Main issues of under-irradiation response (thermal conductivity and its degradation with burnup, densification-swelling behaviour and FGR) are addressed. It is expected that these results together with future PIE and parallel experiments findings, may highlight prominent features of these innovative nuclear fuels
Under Irradiation Issues of the CSZ-based Inert Matrix Fuels from IFA-652 Halden Experiment
No full textThe renewal of interest in Fast Reactors (FR) deployment as possible answer to the future perspective of natural uranium shortage has led to reconsider the plutonium as an energy source with a consequent radical change in its management. The transition to a closed fuel cycle is estimated to occur in a period ranging from 2030 to 2050. In the near term, Inert Matrix Fuels (IMF) are currently viable options, capable to contribute in reducing the proliferation risk associated to nowadays separated plutonium stockpiles in a business-as-usual nuclear energy scenario where a prominent role is still played by Light Water Reactors (LWR) in a once-through fuel cycle strategy. Double-strata strategy with Accelerator Driven Systems (ADS) machines to transmute minor actinides also support the hypothesis of inert matrix heterogeneous fuels. These reasons justify the current interest in IMF. A high burning efficiency achieved by preventing new plutonium build-up under irradiation (U-free fuel), a proved high radiation damage and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Amongst other options, Calcia-Stabilised Zirconia (CSZ) and Yttria-Stabilised Zirconia (YSZ) fulfil these criteria standing as the most promising matrices to host plutonium. ENEA has conceived an in-pile testing (IFA-652 experiment) of viable inert matrices concepts: CSZ and thoria. This experiment was performed in the Halden Heavy Boiling Water Reactor (HBWR), a joint project of the OECD Nuclear Energy Agency. The test-rig is a six-rod bundle loaded with IM, IMT and T innovative fuels. IM and T fuels have, respectively, CSZ and thoria matrix, the fissile phase is High Enriched Uranium (HEU) oxide (UO2 93% 235U enriched). IMT is a ternary fuel composed of CSZ + thoria matrix and HEU oxide as fissile phase. Thoria is added in IMT fuel mainly to improve the low IM reactivity feedback coefficients. The discharge burnup, under typical LWR conditions, was, for all pins, 90-95% of the planned 45 MWd/kgU. Pins are instrumented providing fuel centreline temperature, pin inner pressure and fuel stack elongation measurements. On the basis of the experimental dataset under power ramping, a possible correlation of the FGR onset of CSZ-based fuels with the Halden Vitanza curve apparently exists. The TRANSURANUS (TU) fuel performance code predictions proved to be in nice agreement with the experimental findings concerning the thermal performance of IMF whilst, regarding FGR and densification, a satisfactory modelling is still needed
Inert Matrix Innovative Fuels for Plutonium Transmutation in LWRs
No full textAbout 100 MT of excess plutonium are going to be originated from warheads dismantling under the START I and II agreements (50 MT by each side US and former Soviet Union), and another about 200 MT are already stockpiled from commercial spent fuel reprocessing. The civilian or reactor grade plutonium (RG-Pu), amounted at about 1000 MT total world inventory at the end of 1995, and is forecasted to attain some of 1600-1700 MT by year 2000. Therefore, there is a pressing need in finding novel and ever more safe methods to deal with all kind of excess plutonium in the aim of rendering it ultimately unusable for proliferation purposes.As to weapons Pu, one leading option (US and Russia) is to burn it in LWRs after having converted it to MOX fuel. However, among the possible types of fuel which can be envisaged to burn plutonium in LWRs, inert matrix fuel is a completely novel concept that finds justification since, especially for weapons Pu transmutation, the precisely required fuel is not immediately available and even for the LWR weapon-MOX, processes and facilities have to be set-up before reaching the operational stage.Inert matrix fuel is an non-fertile oxide fuel consisting of PuO2, either weapon-grade or reactor-grade, diluted in a mixed compound of inert oxides such as stabilized ZrO2, Al2O3, MgO or MgAl2O4, and its primary advantage is the no-production of new plutonium during irradiation, because it does not contain uranium (U-free fuel) whose U-238 isotope is the departure nuclide for breeding Pu-239. This new fuel will have a plutonium relative content comparable to that one used in standard MOX fuel for PWRs. After discharge from reactor and adequate cooling time, the spent fuel is outlooked to be sent, as a HLW, directly to the final disposal in deep geological formations without requiring any further reprocessing treatment (once-through solution)
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