In the transportation of radioactive waste, the package is designed as the major engineered system capable to ensure the containment and provide safety functions, such as radiation shielding, structural integrity against external mechanical and thermal loads, dissipation of the decay heat, etc. Packaging systems are designed in accordance to rigorous acceptance requirements, like the International Atomic Energy Agency (IAEA) ones, so to provide protection to human being and environment against radiation exposure and contamination, particularly in reference accident scenarios including, as it is widely known in literature, drop, puncture, fire and submersion tests. The scope of the present study is to evaluate the structural response and performance in a free drop test condition of a new Italian packaging system that should be used for the transportation of low and intermediate level radioactive wastes. For this purpose the carried out numerical analyses are presented and discussed. The numerical analyses, performed by the finite element MARCO code, simulate the behaviour of the packaging system components: the overpack, gasket, cover lid, bolts and a concrete matrix representative of the radioactive content.The obtained results for 1.2 m horizontal drop, on a flat and unyielding surface, were critically analysed and also compared to the experimental ones obtained from the experimental test campaign performed at the Unipi test facility on the new Italian packaging system considered.The stress and acceleration values indicate that the package, although rather local deformations in correspondence of bolts and secondary lid, is capable to withstand the dynamic loading generated during the drop test without any unacceptable loss of the safety features.

Safety and security aspects are of meaningful importance in the design of nuclear facilities.In this study, the attention is so focused on the potential damaging effects that a large civilian airplane impact could bring in safety relevant structures, like a superficial repository similar to El Cabril one.Safety performances of such a type of superficial disposal facility, subjected to the aircraft impact and fuel burning, have been analysed and discussed. Conservative assumptions have been made: normal impact on the lateral repository surface, fire scenario based on the amount of fuel burnt.Load functions (calculated with the Riera approach) and the maximum temperature reached by fuel during its combustion were used as input (boundary condition) in the numerical simulations as well as the damaging phenomena occurring in the concrete structure.Numerical analyses, by MARCO code, allowed to simulate the thermo-structural performances of the superficial repository.The obtained results showed that a repository wall thickness, ranging from 0.6 to 0.9 m, is not sufficient to prevent the penetration of wall itself. Despite the ongoing concrete degradation phenomena, the global strength of the repository seemed to be guaranteed.

For the long term nuclear energy sustainability, the transition to a fast reactor (FRs) based fleet and the adoption of closed fuel cycles is envisaged for answering both for resources optimization and waste reduction. The fuel cycle actually implemented is a once-through cycle based on uranium consuming reactors with thermal spectrum. In this cycle, only ~1% of the uranium extracted is used and the spent fuel is sent to disposal without reusing the fissile material (e.g. Pu239). This leads to a high demand on disposal capacities in terms of masses, radiotoxicity and heat load. Due to specific features of fast spectrum reactors, fuels containing a fraction of Minor Actinides (MAs) can be loaded into their cores and closed (or partially closed) fuel cycles can be implemented thus providing an option for MAs transmutation. However, advantages in sustainability should not lead to a lower safety level. Within the Collaborative Project on European Sodium Fast Reactor, CP-ESFR, reference designs for cores with oxide and carbide fuels were proposed by CEA (France). Both systems show a positive Sodium Void Reactivity Effect (SVRE) at BOL. In the present paper, the optimization process followed for reducing the BOL SVRE is described. Only the oxide core has been considered. The main modifications are related to the axial structure by introducing a larger Na plenum (60 cm) close to the core, an absorber layer above and a lower fertile blanket. All together these attempts allow improving neutron leakages under voided conditions and therefore reducing the SVRE. In the paper, we refer to the optimized configuration as CONF2. The low SVRE in CONF2 (about 2$ less than the reference configuration) offers an opportunity for introduction of MAs in core. Heterogeneous and homogeneous strategies have been considered within the CP-ESFR project. In the present paper, we focus only on homogenous loadings considering the introduction of 2 and 4%wt. AmO2 in core plus lower blanket. The results in terms of safety and fuel cycle characteristics have been compared. The introduction of MAs implies the deterioration of safety parameters (void, Doppler, etc.) but allows the core burning its own waste. The advantages of Am loading in terms of fuel cycle have been evaluated considering the impact on fuel cycle facilities, fuel composition evolution, radiotoxicity and heat load. SESSION IV: TRANSMUTATION SYSTEMS: DESIGN, PERFORMANCE AND SAFETY 29 In order to assess these parameters, a simplified reference scenario representative of a country that wants to develop nuclear energy in isolation has been selected. A constant nuclear energy production in the period 2020-2200 has been defined to better underline the parameters affecting the dynamic of the transition from Light Water Reactors (LWRs) to FRs. Only Pu and MAs actinides produced in the cycle have been considered focusing on the characteristics of the reactor in transmuting its own waste by reaching an equilibrium composition. Neutronic analyses have been performed by means of the deterministic ERANOS code systems using JEFF3.1 data library and scenarios mentioned above have been modelled with the COSI6 code, the dynamic scenario code developed at CEA (France), widely used in Europe.

The aircraft impact accident has become very significant in the design of a nuclear facilities, particularly, after the tragic September 2001 event, that raised the public concern about the potential damaging effects that the impact of a large civilian airplane could bring in safety relevant structures. The aim of this study is therefore to preliminarily evaluate the global response and the structural effects induced by the impact of a military or commercial airplane (actually considered as a “beyond design basis” event) into a near surface radioactive waste (RWs) disposal facility. The safety evaluation was carried out according to the International safety and design guidelines and in agreement with the stress tests requirements for the security track. To achieve the purpose, a lay out and a scheme of a possible near surface repository, like for example those of the El Cabril one, were taken into account. In order to preliminarily perform a reliable analysis of such a large-scale structure and to determine the structural effects induced by such a types of impulsive loads, a realistic, but still operable, numerical model with suitable materials characteristics was implemented by means of FEM codes. In the carried out structural analyses, the RWs repository was considered a “robust” target, due to its thicker walls and main constitutive materials (steel and reinforced concrete). In addition to adequately represent the dynamic response of repository under crashing, relevant physical phenomena (i.e. penetration, spalling, etc.) were simulated and analysed. The preliminary assessment of the effects induced by the dynamic/impulsive loads allowed generally to verify the residual strength capability of the repository considered. The obtained preliminary results highlighted a remarkable potential to withstand the impact of military/large commercial aircrafts, even in presence of ongoing concrete progressive failure (some penetration and spalling of the concrete wall) of the impacted area.

The paper is aimed at preliminary evaluating the load bearing capability of a Generation IV metal cooled reactor (LFR), like the European Lead-cooled System (ELSY), with the intent to support the viability of lead technology for use in a future commercial power plant with waste transmutation capability.Two relevant safety aspects were investigated: the fluid-structure interaction, induced by the propagation of seismic loading (in the event of safe shutdown earthquake), and the high thermal loads due to the residual decay heat (influenced also by the burn up level). The first issue depends on the presence of a liquid free surface that, especially, in SSE condition allows for fluid motion ("sloshing"); the forming and the impact of lead waves could impair the integrity of reactor structures. The second aspect deals with the thermal effects induced by the irradiated core heat source after the reactor shutdown.In order to evaluate the ELSY reactor structural response, induced by both high temperature and hydrodynamic loads, appropriate 3-D finite element models were set up and implemented in MARC code. In addition the input decay power after shutdown has been evaluated by means of MCNPX and Origen2.2 codes.The obtained preliminary results provide information to make possible an upgrading of the reactor design, however with no significant modification of their functionality. (C) 2012 Elsevier B.V. All rights reserved.

In order to investigate the impact of nuclear energy introduction in a country with a fossil fuel based energy mix, several scenarios have been compared in terms of fuel cycle needs (resources and infrastructure) and wastes produced.As case study, the Italian situation (represented by ca. 300 TWhe-y of electricity needs in 2007 and by no nuclear energy production at present) has been selected. However, the obtained results could he extrapolated to other countries by means of scale factors.For the reference scenario, the introduction of Gen.III+ Light Water Reactors and once-through fuel cycle has been considered. Under the hypothesis that only the plutonium produced in the country will be available and used for a possible transition to a fast fleet, the introduction of different types of fast reactors (a 600 MWe lead-cooled and two 1500 MWe sodium-cooled systems with different breeding characteristics) and of a more sustainable fuel cycle (closed or partially closed) have been compared. The adoption of fast systems enables to reduce of 50% the uranium consumption and to favorably impact the cycle back-end by reducing the Pu inventory in the cycle, and by reducing the long term waste radiotoxicity and heat load in a repository.A parametric study has been carried out in order to deal with the systematic uncertainties connected to scenario investigations.

The main goal of the present study is the preliminary evaluation of the seismic demand of a LFR with reference to European Lead System project (ELSY) considered one of the most promising innovative Generation IV reactor. The safety aspects of the ELSY reactor in the event of a Safe Shutdown Earthquake, taking into account also the effects of the possible fluid-structure interaction, have been analyzed.To the purpose to determine the seismic demand, in according with the international rules, a non-linear dynamic analysis method was used with rather refined 3-D model of LFR for the foreseen structural analyses and simulations of the plant and of the reactor internals behaviour. In this report numerical results are presented and discussed highlighting the relevance of the fluid-structure interaction in terms of structural integrity as well as the isolation technique effectiveness, which is expected to increase the safety margin of the reactor structures during a seismic event, if the isolators frequency is far from that of the reactor.The present work has been performed within the 6th European Framework Project. (C) 2011 Elsevier B.V. All rights reserved.

For the assessment of the safety and durability of a nuclear power plant (NPP), the containment building behaviour shall be evaluated, under various service and extreme conditions, both natural or produced by natural accident or vicious man activities, like September 2001 jet aircraft crashes.The aim of this paper is to preliminary evaluate the effects and consequences of the energy transmitted to the outer containment walls (according to the international safety and design code guidelines, as NRC or IAEA ones) due to a military or civil aircraft impact into a nuclear plant, considered as a 'beyond design basis' event.To perform reliable analysis of such a large-scale structure and determine the structural effects of the propagation of this types of impulsive loads (response of containment structure), a realistic but still feasible numerical model with suitable materials characteristics were used by means of which relevant physical phenomena are reflected. Moreover a sensitivity analysis has also been carried out considering the effects of different containment wall thickness and reinforced/prestressed concrete features. The obtained results were analysed to check the NPP containment strength margins. (C) 2011 Elsevier B.V. All rights reserved.

Like all industries, the generation of electricity from Nuclear Power Plant produces wastes to be managed. Spent fuel element casks used for transport of nuclear materials must be designed according to rigorous acceptance criteria and standards requirements, e.g. International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure.The aim of this work was the evaluation of the integrity of spent fuel cask under both normal and accident transport conditions, such as impact (9 m drop impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation) and rigorous fire events (full exposure to an engulfing fire for 30 min (fire test) or to an environment at 800 degrees C temperature for a numerical simulation or for a furnace test).Using the finite element code ANSYS both steady-state and transient thermal analyses were carried Out to determine the maximum fuel temperature and the temperatures behaviour into the cask, considering all the heat transfer modes between the cask and the external environment as well as inside the cask itself. Moreover, both wet and dry fuel storage inside the cavity of the body were analyzed. The obtained results, used for the new licensing approval by the Italian competent Authority of the cask for PWR spent fuel transport, are discussed. (C) 2010 Elsevier Ltd. All rights reserved.

In recent times there is a renewed worldwide interest in the development and application of advanced nuclear power plants (NPPs). Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR in Finland and France, AP1000 in China, etc.) and more are under consideration for licensing in several countries.Innovative NPPs are designed to be built with very broad siting conditions; therefore the safety aspects related to the external events might follow new scenarios and failure modes, different from those well known for the currently operated reactors.In this paper, the intent is evaluating the structural integrity of a nuclear containment system subjected to dynamic loadings due to a Design Base Earthquake and an aircraft impact (large size civilian jets or military aircrafts impact), which represent the two most relevant external accidents that should be considered and investigated as part of the basic design of a NPP in particular a III+ and IV Gens.In fact a suitable safety design of the NPP containment system (according to the international safety and design code guidelines, as NRC or IAEA ones), even if designed to meet other design goal, may represent a "built-in protection" to avoid or mitigate the effects of mentioned dynamic loadings.To the purpose a rather sophisticated numerical methodology, adopting finite element (FEM) approach, is employed for studying the overall dynamic behaviour of nuclear reactor and to determine the structural effects of the propagation of dynamic seismic as well as impulsive loads (containment structure response) up to the relevant nuclear components. Therefore representative three-dimensional FEM models of mentioned NPP containment and aircraft structures were set up, and used, in the performed analyses taking also into account the suitable materials behaviour and their related constitutive laws as well as the seismic excitation (determined according to the NRC rules).Moreover the performed analyses and the carried out response analyses of internal components, to both the ground motion and impact loads, were studied to check the considered NPP containment strength reserve in the case of the considered events. The obtained results seem to confirm the possibility to achieve an optimization of the NPP internal components.

This paper deals with the numerical and experimental analyses of a shell type shock absorber for a nuclear spent fuel cask. Nine-meter free drop tests performed on reduced scale models are described. The results are compared with numerical simulations performed with FEM computer codes, considering reduced scale models as well as the prototype. The paper shows the results of a similitude analysis, with which the data obtained by means of the reduced scale models can be extrapolated to the prototype. Small discrepancies were obtained using large-scale models (1:2 and 1:6), while small-scale models (1:12) did not give reliable results. A 1:9 scale model provided useful information with a less than 20% error.