Giuseppe Forasassi

Preliminary Evaluation of a Nuclear Scenario Involving Innovative Gas Cooled Reactors

In order to guarantee a sustainable supply of future energy demand without compromising the environment, some actions for a substantial reduction of CO2 emissions are nowadays deeply analysed. One of them is the improvement of the nuclear energy use. In this framework, innovative gas-cooled reactors (both thermal and fast) seem to be very attractive from the electricity production point of view and for the potential industrial use along the high temperature processes (e.g., H2 production by steam reforming or I-S process). This work focuses on a preliminary (and conservative) evaluation of possible advantages that a symbiotic cycle (EPR-PBMR-GCFR) could entail, with special regard to the reduction of the HLW inventory and the optimization of the exploitation of the fuel resources. The comparison between the symbiotic cycle chosen and the reference one (once-through scenario, i.e., EPR-SNF directly disposed) shows a reduction of the time needed to reach a fixed reference level from ∼170000 years to ∼1550 years (comparable with typical human times and for this reason more acceptable by the public opinion). In addition, this cycle enables to have a more efficient use of resources involved: the total electric energy produced becomes equal to ∼630u2009TWh/year (instead of only ∼530u2009TWh/year using only EPR) without consuming additional raw materials.

SEISMIC ISOLATION OF THE IRIS NUCLEAR PLANT

The safety-by-design™ approach adopted for the design of the International Reactor Innovative and Secure (IRIS) resulted in the elimination by design of some of the main accident scenarios classically applicable to Pressurized Water Reactors (PWR) and to the reduction of either consequences or frequency of the remaining classical at-power accident initiators. As a result of such strategy the Core Damage Frequency (CDF) from at-power internal initiating events was reduced to the 10−8 /ry order of magnitude, thus elevating CDF from external events (seismic above all) to an even more significant contributor than what currently experienced in the existing PWR fleet. The same safety-by-design™ approach was then exported from the design of the IRIS reactor and of its safety systems to the design of the IRIS Nuclear Steam Supply System (NSSS) building, with the goal of reducing the impact of seismically induced scenarios. The small footprint of the IRIS NSSS building, which includes all Engineered Safety Features (ESF), all the emergency heat sink and all the required support systems makes the idea of seismic isolation of the entire nuclear island a relatively easy and economically competitive solution. The seismically isolated IRIS NSSS building dramatically reduces the seismic excitation perceived by the reactor vessel, the containment structure and all the main IRIS ESF components, thus virtually eliminating the seismic-induced CDF. This solution is also contributing to the standardization of the IRIS plant, with a single design compatible with a variety of sites covering a wide spectrum of seismic conditions. The conceptual IRIS seismic isolation system is herein presented, along with a selection of the preliminary seismic analyses confirming the drastic reduction of the seismic excitation to the IRIS NSSS building. Along with the adoption of the seismic isolation system, a more refined approach to the computation of the fragility analysis of the components is also being developed, in order to reduce the undue conservatism historically affecting seismic analysis. The new fragility analysis methodology will be particularly focused on the analysis of the isolators themselves, which will now be the limiting components in the evaluation of the overall seismic induced CDF.Copyright

The Capabilities of HTRs to Burn Actinides and to Optimize Plutonium Exploitation

At present, the 125 GWe of nuclear power in the European Union produce about 3000 tons of spent fuel annually, containing about 25 tons of plutonium, 2.5 tons of minor actinides (MA) and about 100 tons of fission products, of which 3.1 tons are long-lived fission products. Actual reprocessing of LWR fuel and a first recycling as mixed plutonium and depleted uranium oxide fuel (MOX) in LWR already contribute to a significant reduction of waste volumes and radiotoxicity. However HTRs have some characteristics which make them particularly attractive: intrinsic safety, cost-effectiveness, reduced thermal pollution, capability of increasing energy availability (with the use of Pu-Th cycle) and of minimizing actinides radiotoxicity and volume of actinides. In this paper particularly the last item is investigated. Symbiotic fuel cycles of LWR and HTR can reach much better waste minimization performances. It happens because of the specific features of HTRs cores that leads to an ultra-high burnup and, last but not least, the ability to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This property is due to a decoupling between the parameters of cooling geometry and of neutronic optimization. In our calculations we considered a pebble-bed HTR using a Pu-based fuel (deriving from reprocessing of classical LWR fuel and/or weapons grade plutonium) at the maximum technological discharge burnup. As results, we find, at EOL (End Of Life), a relatively small amounts of residual Pu and MA produced in terms of quantities and of radiotoxicities. Furthermore we used in our calculations a different type of fuel based on a mixture of Pu and Th to try to optimize the previous results and to increase energy availability. Calculations have been done using MCNP-based burnup codes, capable of treating 3-D complex geometry and ultra-high burnup.Copyright

Buckling of Imperfect Thin Cylindrical Shell under Lateral Pressure

The strength of thin shells, under external pressure, is highly ndependent by the nature of imperfection. This paper investigates nbuckling behaviour of imperfect thin cylindrical shells with nanalytical, numerical, and experimental methods in conditions for nwhich, at present, a complete theoretical analysis was not found nin literature. In general, collapse is initiated by yielding, but ninteraction with geometrical instabilities is meaningful, in that nimperfections reduce the load bearing capacity by an amount of nengineering significance also when thickness is considerable. The naim of this study was to conduct experiments that are nrepresentative of buckling, in the context of NPP applications as, nfor instance, the IRIS (international reactor innovative and nsecure) and LWR steam generator (SG) tubes. At Pisa University, a nresearch activity is being carried out on the buckling of thin nwalled metal specimen, with a test equipment (and the necessary ndata acquisition facility) as well as numerical models were set up nby means FEM code. The experiments were conducted on A-316 test nspecimens, tubes with and without longitudinal welding. The nnumerical and experimental results comparison highlighted the ninfluence of different types of imperfections on the buckling nloads with a good agreement between the finite-element predictions nand the experimental data.

Wastes Management Through Transmutation in an ADS Reactor

The main challenge in nuclear fuel cycle closure is the reduction of the potential radiotoxicity, or of the time in which that possible hazard really exists. Probably, the transmutation of minor actinides with fast fission processes is the most effective answer. This work, performed in SCK⋅CEN (Belgium) and DIMNP Pisa University, is focused on preliminary evaluation of industrial scale ADS (400u2009MWth, 2.5u2009mA) burning capability. An inert matrix fuel of minor actinides, 50%u2009vol. MgO and 50%u2009vol. (Pu,Np,Am,Cm)O1.88, core content, with 150 GWd/ton discharge burn up, is used. The calculations were performed using ALEPH-1.1.2, MCNPX-2.5.0, and ORIGEN2.2. codes.

Evaluation of safety performances of a superficial disposal facility subject to an aircraft impact and fuel burning

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 MARC© 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.Copyright

Horizontal drop test evaluation of a packaging system

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 MARC® 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.Copyright

Preliminary evaluation of core compaction phenomenon: Methodological approach

Nuclear reactors have to be maintained in a critical state so as to keep the chain fission process stationary and under control. Nuclear stability considerations dictate that the geometry of the core be closely controlled at all times: therefore any modification of it must be predictable, compatible with the requirements of the interfacing reactor systems and safely manageable by (intrinsic and engineered) control mechanisms.This study deals with the evaluation of the deformation of core (and restraint system) geometry due to dynamic perturbations. This deformation may determine, at large or small extent, an assembly compaction, that is generally characterized by a radial inward displacement and, eventually, results in an insertion of reactivity.To the aim it is of meaningful importance to set up and develop an overall and reliable methodological approach to be used in designing the core system (all structures and components characterizing the core region) and evaluating its performance under operation and accident condition. In particular a LMR reactor configuration similar to the Advanced Lead-cooled Fast Reactor European Demonstrator - ALFRED (300 MWth) has been considered.The assessment of the dynamic behaviour of a LMR core is particularly needed for seismic design purposes: these solicitations could deform the core system and fuel assembly. A preliminary finite element model, in which all the core sub-assemblies were represented as masses distributed on the supporting plate, was carried out in order to investigate the dynamic response of the structures once confidence was established by sensitivity analyses of size and type of the adopted elements.The preliminary results indicate that the core region is undergoing local deformations (of about 3 cm) that could influence the normal reactor operation. Although any deformation influences the normal reactor operation, it is expected that the reactivity specifically related to this deformation will not pose concerns to the safe manageability of the associated abnormal operation.Copyright

Global Structural Response of an SMR Reactor Subjected to an Aircraft Impact

Although a 11 September 2001-style vicious attack on a nuclear power plant is considered as ‘beyond design basis’ event, an appropriate design of the nuclear facilities, with features and functional capabilities, is of meaningful importance to demonstrate that the reactor containment could safely withstand the impact of a large commercial aircraft without any radioactive release. This paper deals with the evaluation of the global structural response and of the vulnerability of a reactor building subjected to a deliberate commercial aircraft impact in the assumption of an attack from multiple entry directions. In this framework, separately from penetration and fire, the ‘shock’ loadings due to the progressive aircraft crashing on the power plant buildings were evaluated, taking into account that even if such penetration occurred, together with some concrete crushing and bent steel rebars, it very probably would not reach the reactor vessel. To the purpose a rather refined numerical methodology was employed and three-dimensional models (FEM approach) of a SMR reactor building and possible realistic, even if simplified, aircraft structures were set up and used in the performed analyses, taking also into account a suitable materials behaviour and constitutive laws. The analysis was performed increasing the severity of the crash scenario, assuming that the during the impact the aircraft transfers the full impact energy of the crash to the structure being struck. The obtained results were analysed to check the additional safety margin of the reactor containment and the fuel pool.Copyright

Evaluation of Damaging Effects on a NPP Integrity Subjected to Beyond Design Earthquake

The aim of this study is to evaluate the influence of the damaging or failure effects induced by seismic loadings on the loading bearing capability of the overall nuclear structures, systems and components (SSCs). In particular it was analyzed the behaviour of the SSCs in terms of their required safety functions and capacity assuming the occurrence of an earthquake having magnitude beyond the design basis value (in agreement with the “stress tests” suggestions, presently foreseen by the European and International Associations).To the purpose, quite refined FEM models were set up and implemented considering suitable materials behaviour and constitutive laws for the reactor materials, particularly for the concrete material behaviour (which could suffer failure and damage mechanisms during the ground shaking). Several type of damaging mechanics were considered.The obtained results seemed to confirm the overall containment reliability even if with minor upgrading actions in operating procedures and design characteristics. Moreover they allowed to appropriately check the NPP containment strength reserve.Copyright