MARIA Reactor Irradiation Technology Capabilities towards Advanced Applications

The MARIA research reactor is designed and operated as a multipurpose nuclear installation, combining material testing, neutron beam experiments, and medical and industrial radionuclide production, including molybdenum-99 (99Mo). Recently, after fuel conversion to LEU and rejuvenation of the staff while maintaining their experience, MARIA has been used to respond to the increased interest of the scientific community in advanced nuclear power studies, both fission and fusion. In this work, we would like to introduce MARIA’ s capabilities in the irradiation technology field and how it can serve future nuclear research worldwide.

Pitting corrosion prediction from cathodic data: application of machine learning

Purpose To constitute input data, the authors carried out electrochemical experiments. The authors performed voltammetric scans in a very cathodic potential region. The authors constituted an experimental table where for each experiment we note the current values recorded at a low polarization range and the pitting potential observed in the anodic region. This study aims to concern carbon steel used in a nuclear installation. The properties of the chemical solutions are close to that of the cooling fluid used in the circuit. Design/methodology/approach In a previous study, this paper demonstrated the effectiveness of machine learning in predicting the localized corrosion resistance of a material by considering as input data the physicochemical properties of its environment (Boucherit et al., 2019). With the present study, the authors improve the results by considering as input data, cathodic currents. The reason of such an approach is to have input data that integrate both the surface state of the material and the physicochemical properties of its environment. Findings The experimental table was submitted to two neural networks, namely, a recurrent network and a convolution network. The convolution network gives better pitting potential predictions. Results also prove that the prediction by observing cathodic currents is better than that obtained by considering the physicochemical properties of the solution. Originality/value The originality of the study lies in the use of cathodic currents as input data. These data contain implicit information on both the chemical environment of the material and its surface condition. This approach appears to be more efficient than considering the chemical composition of the solution as input data. The objective of this study remains, at the same time, to seek the optimal neuronal architectures and the best input data.

Overview of the seismic probabilistic safety assessment applied to a nuclear installation located in a low seismicity zone

Deterministic and probabilistic nuclear safety analysis methodologies have been developed and updated based on operational experience, investigation of past incidents or accidents, and analysis of postulated initiating events in order to maintain the protection of workers, the public and the environment. The evaluation of accident sequences and the total radiological risk resulting from off-site releases are general objectives addressed by these methodologies. There are hazards that continually challenge the safety of a nuclear facility or its nearby area. In particular, seismic events represent a major contributor to the risk of a nuclear facility. Different levels of ground motion induced by earthquakes may be experienced by the structures, systems and components (SSCs) of the installation. In this context, a seismic hazard analysis, seismic demand analysis and seismic fragility analysis must be carried out in order to characterize the local seismic hazard and what are the seismic demands on SSCs, allowing an adequate seismic classification of SSCs, even in installations located in sites with low seismicity. In this article, a general description of the Seismic Probabilistic Safety Assessment (Seismic PSA) methodology is presented, with emphasis on their support studies, aiming at applying the methodology described in this article to an experimental nuclear installation containing a PWR reactor designed for naval propulsion to be installed in a low seismicity zone in Brazil.

Results of Analysis of Aging Related Events at Ukrainian NPPs

The design lifetime (30 years) of Ukrainian NPPs has already been expired or will be expired soon. One of the priority areas for the development of nuclear energy is long-term operation (LTO) based on the results of periodic safety review. At present, eleven Ukrainian NPP units are operated under LTO conditions, which in turn causes certain features of operation related to wear, fatigue and general aging of components and structures. An issue of aging management is considered one of the main factors to ensure the safety of a nuclear installation, both in the design lifetime and in the long-term operation. The primary purpose of aging management is to ensure the safety and maximum efficiency of operation through the implementation of technically and economically feasible measures aimed at timely detection and keeping of degradation of NPP components caused by aging within the limits. One way to determine the efficiency of implemented measures and aging management programs is to analyze operational experience and, in particular, analyze the number of NPP operational occurrences during the design and long-term operation period and that are directly related to the aging effects. This paper presents the results of the analysis of the events at Ukrainian NPPs caused by aging processes, compares the results of national and international experience.

Analysis of the thermal stress state of the flange connections elements of the emergency cooling heat exchangers during accidents of the WWER-1000 nuclear installation

During the assessment of the static strength of the flange connections elements Dn2130 and Dn2080 of the emergency cooling heat exchangers 08.8111.335 SB (TOAR), it was found that there is an excess of the allowable stress values. These calculations of static strength performed using the finite element method (FEM). The analysis of the static strength of the flange joints was performed taking into account the design values of the tightening of the studs, equal to 22,527 kgf and 8,836 kgf, accordingly. At the same time, one of the main purposes of heat exchangers TOAR nuclear installation (NI) WWER-1000 is the work until accidents. The analysis of accidents of NI WWER-1000 showed that the largest values of change of parameters of environments in heat exchangers of TOAR correspond to accident “LOCA: Bilateral rupture of MCT”. Based on this, we considered the thermal stress state of heat exchangers for this accident. To determine the thermal stress state of the TOAR heat exchanger elements, during accidents of the nuclear installation, strength calculations were performed in the non-stationary formulation of the problem. One of the boundary conditions for these strength calculations is the distribution of temperatures along the thickness and length of the walls of the elements of the heat exchanger, which changes over time. Numerical thermohydraulic calculations were performed to determine these boundary conditions. In the article for the first time the results of calculations of thermal stress state of separate elements of heat exchangers TOAR, for work of heat exchangers during accidents of nuclear installation are received. It is established that the elements of the flange connection Dn2130 are one of the most critical elements of TOAR heat exchangers. To determine the thermal stress state of the heat exchanger elements, analytical thermal calculations, numerical thermohydraulic and strength calculations were performed using the FEM method. As a result of the analysis of the performed strength calculations, it was concluded that it is necessary to reduce the tightening value of the flanges of the flange connection Dn2130 to 14600 kgf.

Control Experiences Regarding Clearable Materials from Nuclear Power Plants and Nuclear Installations: Scaling Factors Determination and Measurements’ Acceptance Criteria Definition

Arpa Piemonte has been carrying out, for a long time, controls on clearable materials from nuclear power plants to verify compliance with clearance levels set by ISIN (Ispettorato Nazionale per la Sicurezza Nucleare e la Radioprotezione - National Inspectorate for Nuclear Safety and Radiation Protection) in the technical prescriptions attached to the Ministerial Decree decommissioning authorization or into category A source authorization (higher level of associated risk, according to the categorization defined in the Italian Legislative Decree No. 230/95). After the experience undertaken at the “FN” (Fabbricazioni Nucleari) Bosco Marengo nuclear installation, some controls have been conducted at the Trino nuclear power plant “E. Fermi,” “LivaNova” nuclear installation based in Saluggia, and “EUREX” (Enriched Uranium Extraction) nuclear installation, also based in Saluggia, according to modalities that envisage, as a final control, the determination of γ-emitting radionuclides through in situ gamma spectrometry measurements. Clearance levels’ compliance verification should be performed for all radionuclides potentially present, including those that are not easily measurable (DTM, Difficult To Measure). It is therefore necessary to carry out upstream, based on a representative number of samples, those radionuclides’ determination in order to estimate scaling factors (SF), defined through the logarithmic average of the ratios between the i-th DTM radionuclide concentration and the related key nuclide. Specific radiochemistry is used for defining DTMs’ concentrations, such as Fe-55, Ni-59, Ni-63, Sr-90, Pu-238, and Pu-239/Pu-240. As a key nuclide, Co-60 was chosen for the activation products (Fe-55, Ni-59, Ni-63) and Cs-137 for fission products (Sr-90) and plutonium (Pu- 238, Pu-239/Pu-240, and Pu-241). The presence of very low radioactivity concentrations, often below the detection limits, can make it difficult to determine the related scaling factors. In this work, the results obtained and measurements’ acceptability criteria are presented, defined with ISIN, that can be used for confirming or excluding a radionuclide presence in the process of verifying clearance levels’ compliance. They are also exposed to evaluations regarding samples’ representativeness chosen for scaling factors’ assessment.