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The measurement of two-phase flow parameters has never been an easy task in the experimental thermal-hydraulics and the need of such measurements in the SPES3 facility has led to investigation of different possibilities and evaluation methods to determine mass flows and energies. This paper deals with the theoretical prediction of the two-phase mass flow rate by the development of a mathematical model for a spool piece, consisting of a drag disk, a turbine and a void fraction detector. Data obtained by simulation of DBAs in the SPES3 facility, with the RELAP5 thermal-hydraulic code, have provided the reference conditions for defining the main thermal-hydraulic parameter ranges and selecting a set of instruments potentially suitable to measure and derive the required quantities. The governing equation and the instrumentation output are defined for each device. Three different turbine models (Aya, Rouhani and volumetric) have been studied to understand which one better adapts to two-phase flow conditions and to investigate the best instrument combination. The mathematical model has been tested versus the RELAP5 results with a reverse process where calculated variables, like void fraction, quality and slip ratio, are given as input to a specifically developed program to get back the mass flow rate. The analytical results, verified versus the DVI break transient, well agree with the RELAP5 mass flow rate. Specific tests on proper experimental loops are required to verify the analytical studies.

A preliminary evaluation of gaseous radiocarbon (14C) behavior under geological repository conditions for Italian radioactive high level waste-long-lived and intermediate level waste disposal has been performed. Although in Italy there is still no defined project for a geological disposal facility, current work may support future safety assessment studies for a hypothetical future repository in deep salt rock. In the Italian context of radioactive waste, the percentage of 14C bearing waste to be disposed in a possible geological repository is low; irradiated graphite is the most important radiological source. Data about the radiological inventory has been collected to simulate production and migration of gaseous 14C in a hypothetical geological repository. Three different conceptual models have been developed and simulated. The first model has considered a preliminary evaluation of the radiological impact referred to the whole inventory; the second and third model have evaluated the impact only due to the irradiated graphite. A preliminary sensitivity analysis has been carried out, highlighting the importance of geometry and of distribution coefficients (Kd) in materials used to seal the disposal underground facility. Results show the possibility to correlate the Kd values, the volume and the location of the sealing materials to the amount of 14C migrating toward the surface.

Disposal facilities are designed to provide long-term isolation of the wastes from the human environment by means of a system of barriers, both engineered and natural. In this regard, it is required that the containment function of the multi-barrier system be evaluated by a quantitative risk analysis procedure, also called performance assessment. This entails the implementation of a Probabilistic Safety Assessment model for representing the stochastic failure behavior of the repository barriers, estimating the release rates of radionuclides into the groundwater stream and the doses from subsequent ingestion of drinking water. To this aim, a Monte Carlo simulation scheme is provided within a reliability modelling approach for the safety performance analysis of a repository, accounting also for barrier degradation processes. An application to a case study of the literature is presented to validate the approach and demonstrate its simplicity and flexibility.

Radiocarbon (14C) is one of the key radionuclides for the performance and safety assessment of a radioactive waste disposal, due to its high activity concentration in waste materials from the nuclear cycle and to its mobility. The measurement of the 14C content in spent ion exchange resins from nuclear reactors is important for the safety assessment of the disposal concept and for the choice of the appropriate treatment/disposal method. Ion exchange resins are commonly used in nuclear reactors as filters for the purification of process liquids or wastes streams and they retain molecules containing radioactive isotopes, among which is 14C. Their efficiency, both as filters and as waste containers, is strictly connected with the morphology. The preservation of spherical shape upon aging is one of the key parameters for their quality assessment and for the evaluation of the potential release of 14C during storage conditions. In this study, the change in IERs morphology during storage periods has been investigated in order to verify correlation with 14C release. Both brand new and aged specimens have been studied in order to assess the quality of the resins after 10 yr of storage and to contribute to the understanding of 14C release mechanisms.

Prediction of radionuclides release is a central issue in the performance assessment of nuclear waste repositories. To this aim a model of radionuclides migration through the repository barriers must be set up, accounting for the uncertainties affecting the process. In this context, the present paper presents the application of Monte Carlo simulation to a Markovian modeling framework proposed in the literature; two cases are presented to highlight the value added by the flexibility of the Monte Carlo simulation approach.

The SPES3 facility is being built at the SIET laboratories, in the frame of an R&D program on Nuclear Fission, led by ENEA and funded by the Italian Ministry of Economic Development. The facility is based on the IRIS reactor design, an advanced medium size, integral layout, pressurized water reactor, based on the proven technology of PWR with an innovative configuration and safety features suitable to cope with Loss of Coolant Accidents through a dynamic coupling of the primary and containment systems. SPES3 is suitable to test the plant response to postulated Design and Beyond Design Basis Events, providing experimental data for code validation and plant safety analysis. It reproduces the primary, secondary and containment systems of the reactor with 1:100 volume scale, full elevation, prototypical fluid and thermal-hydraulic conditions. A design-calculation feedback process, based on the comparison between IRIS and SPES3 simulations, performed respectively by FER, with GOTHIC and RELAP5 coupled codes, and by SIET, with RELAP5 code, led to reduce the differences in the two plants behaviour, versus a 2-inch equivalent DVI line DEG break, considered the most challenging LOCA for the IRIS plant. Once available the final design of SPES3, further calculations were performed to investigate Beyond Design Basis Events, where the intervention of the Passive Containment Condenser is fundamental for the accident recovery. Sensitivity analyses showed the importance of the PCC actuation time, to limit the containment pressure, to reach an early pressure equalization between the primary and containment systems and to allow passive water transfer from the containment to the RPV, enhanced by the ADS Stage-II opening.