Analysis of feasibility of a new core catcher for the in-vessel core melt retention strategy

This study deals with the feasibility study of a new in-vessel core melt retention (IVCMR) strategy capable to extend the coping period in the event of adverse situations, involving the melting of the core. Since Fukushima accident, many studies have been carried out to resolve the severe accident mitigation issues related to the corium stabilization inside and outside the reactor vessel. This is in fact one of the most relevant safety issues to secure LWRs from the point of view of severe accident mitigation and containment integrity. As for the corium stabilization inside the reactor vessel, in this study it is proposed a new IVCMR concept, developed at the University of Pisa, based on the adoption of an original core catcher design made of batches of ceramic material. By profiting of its low thermal conductivity, this core catcher is capable to retard the heat-up of the lower head of the vessel during the phase of relocation of the corium. To support the feasibility of its design analytical and numerical analyses have been performed assuming homogeneous pool condition. Results show that the adoption of the proposed core catcher solution extends the severe accident coping period: after 1 h from the initiating event, the maximum temperature of the vessel wall is below the limit for which localized failure may appear

Preliminary analysis of an aged RPV subjected to station blackout

Today, 46% of operating Nuclear Power Plants (NPP) have a lifetime between 31 and 40 years, while 19% have been in operation for more than 40 years. Long Term Operation (LTO) is an urgent requirement for all of the nuclear industry. The aim of this study is to assess the performance of a reactor pressure vessel (RPV) subjected to a station blackout (SBO) event. Alterations suffered by the material properties and creep at elevated temperatures are considered. In this study, coupling between MELCOR and Finite Element Method (FEM) codes is carried out. In the Finite Element (FE) model, the combined effects of ageing and creep are implemented through degraded material properties and a viscoplastic model. The reliability of the model is validated by comparing the FOREVER/C1 experimental results. The results show that the RPV lower head bends downwards with a maximum radial expansion of about 260 mm and RPV thermomechanical properties are reduced by more than 50% at high temperatures. The effects of ageing, creep and long heat-up strongly affect the resistance of the RPV system until the point of compromising it in the absence of/delayed emergency intervention. Aged RPV at end-of-life may collapse earlier, and in less time, with the same accidental conditions

Detailed neutronic study of the power evolution for the European Sodium Fast Reactor during a positive insertion of reactivity

Abstract The new reactor concepts proposed in the Generation IV International Forum require the development and validation of new components and new materials. Inside the Collaborative Project on the European Sodium Fast Reactor, several accidental scenario have been studied. Nevertheless, none of them coped with mechanical safety assessment of the fuel cladding under accidental conditions. Among the accidental conditions considered, there is the unprotected transient of overpower (UTOP), due to the insertion, at the end of the first fuel cycle, of a positive reactivity into the reactor core as a consequence of the unexpected runaway of one control rod. The goal of the study was the search for a detailed distribution of the fission power, in the radial and axial directions, within the power peaked fuel pin under the above accidental conditions. Results show that after the control rod ejection an increase from 658 W/cm 3 to 894 W/cm 3 , i.e. of some 36%, is expected for the power peaked fuel pin. This information will represent the base to investigate, in a future work, the fuel cladding safety margin

RELAP5/SIMMER-III code coupling development for PbLi-water interaction

A major safety issue in the Water-Cooled Lead-Lithium Breeding Blanket (WCLL-BB) system foreseen for fusion reactor is the interaction concerning the primary coolant (water) and the neutron multiplier (PbLi), due to a hypothetical tube rupture in the coolant circuit. This scenario involves an exothermic chemical reaction between PbLi and water with the production of hydrogen, in addition to critical interactions in a complex multiphase system in non-thermal equilibrium. In recent years the PbLi/water reaction was successfully implemented in the SIMMER-III code and validated against data from the LIFUS5/Mod3 experimental campaign. However, due to limitations of SIMMER-III, this work was restricted to the prediction of the phenomena inside the vessel, neglecting the simulation of the injection line. Nevertheless, since the injection line may actually have an important effect on the development of the transient, the simulation of the whole facility would be highly desirable. Indeed, the University of Pisa recently developed a coupling methodology between the SIMMER-III and RELAP5/Mod3.3 codes and applied it to simple single-phase cases. In this paper the complete simulation of the LIFUS5/Mod3 facility is presented, with the injection line modelled through RELAP5. Furthermore, all the complex aspects of the phenomena inside the reaction tank were included: the multiphase system and the interaction between water and PbLi with the chemical reaction and the production of hydrogen were modelled by SIMMER. Preliminary results are presented, showing that the coupling methodology can be effectively employed for the prediction of the chemical and thermal-hydraulic behaviour of complex loop experimental facilities

Multicentre investigation of neutron contamination at cardiac implantable electronic device (CIED) location due to high-energy photon beams using passive detectors and Monte Carlo simulations

Radiotherapy treatments involving LINACs operating at accelerating potentials >10 MV generate (photo)neutrons which deliver dose to patients also outside the target volume. This effect is particularly relevant for patients with cardiac implantable electronic devices (CIEDs), which can be damaged by the therapeutic irradiation. In the last few years, there has been a rising interest in this issue, and it seems that damage to CIEDs is primarily associated with the thermal component of the photoneutron field. In particular, a recent study led by Politecnico di Milano considered CIEDs from various manufacturers and showed that some of these devices can be damaged after an irradiation with a thermal neutron fluence of about 10^9 cm^-2. The present work results from a collaboration among Politecnico di Milano, the University of Pisa, the University of Trieste and three Italian hospitals located in Lucca, Trieste and Varese, respectively, and it is primarily aimed at evaluating the thermal neutron fluence in CIED region for some high-energy treatments delivered at 15 and 18 MV and to determine whether it is comparable to the critical value given above, which has been experimentally determined to be potentially harmful for CIEDs. Thermal neutron fluence was measured through CR-39 detectors and TLDs, which were housed inside a BOMAB-like phantom mimicking the patient’s trunk. The experimental sessions involved two models of LINAC, Varian Clinac DHX (Varese hospital) and Elekta Synergy (Lucca and Trieste hospitals). The experimental results show that the treatments considered in this study can lead to a thermal neutron fluence in the cardiac region comparable to the critical value. Furthermore, detailed Monte Carlo geometries for the facilities involved in this project were developed with the MCNP code (v. 6.2), and they were tested by comparing simulation results to measurements considering some benchmark irradiation plans. Bubble detectors were also employed for fast neutron fluence measurements to be compared to simulation outputs. These computational models stand out as promising tools for the investigations required in this work, and they can be used for further studies also extending their use to analogous facilities hosting the same models of LINACs

Calibration of PADC-based neutron area dosemeters in the neutron field produced in the treatment room of a medical LINAC

a b s t r a c t PADC-based nuclear track detectors have been widely used as convenient ambient dosemeters in many working places. However, due to the large energy dependence of their response in terms of ambient dose equivalent (H * (10)) and to the diversity of workplace fields in terms of energy distribution, the appropriate calibration of these dosemeters is a delicate task. These are among the reasons why ISO has introduced the 12789 Series of Standards, where the simulated workplace neutron fields are introduced and their use to calibrate neutron dosemeters is recommended. This approach was applied in the present work to the UAB PADC-based nuclear track detectors. As a suitable workplace, the treatment room of a 15 MV Varian CLINAC DHX medical accelerator, located in the Ospedale S. Chiara (Pisa), was chosen. Here the neutron spectra in two points of tests (1.5 m and 2 m from the isocenter) were determined with the INFN-LNF Bonner Sphere Spectrometer equipped with Dysprosium activation foils (Dy-BSS), and the values of H * (10) were derived on this basis. The PADC dosemeters were exposed in these points. Their workplace specific H*(10) responses were determined and compared with those previously obtained in different simulated workplace or reference (ISO 8529) neutron fields

Cutting-edge R&D activities of CIRTEN in support of the Technology Park annexed to the Italian National Repository of radioactive waste

R&D activities taking place at the institutions belonging to Consorzio Interuniversitario per la Ricerca TEcnologica Nucleare are here presented and discussed. A special focus is on Technology Park annexed to the Italian National Repository of radioactive waste

Relazione sulle attività di monitoraggio finali presso l’impianto RTS-1 (Fase 2 del decommissioning)

Nel presente documento sono descritte le attività di monitoraggio radiologico finale effettuate dal Dipartimento di Ingegneria Civile e Industriale (DICI) dell’Università di Pisa, in qualità di Ente Terzo Certificatorio, nell’ambito della Fase 2 del decommissioning del reattore “RTS-1 Galileo Galilei” del Ministero della Difesa, situato presso il sito del Centro Interforze Studi per le Applicazioni Militari (CISAM) di San Piero a Grado (Pisa)