Accident Analysis in Research Reactors

The first Research Reactor reaches its criticality on 2nd December 1942. Nowadays, more than 300 research reactors are currently in operation. Over the past fifty years, research reactors have progressed through a variety of tasks. These have included materials research using neutron scattering and diffraction, materials characterization by activation analysis and radiography, isotope production, irradiation testing, as well as training, and service as centres of excellence in science and technology. With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible a few years ago can now be performed. The purpose of the present thesis is to provide an overview of the accident analysis technology applied to the research reactor, with emphasis given to the capabilities of computational tools

Methodology for the Analysis of Fuel Behavior during LOCA and RIA

Loss of coolant accidents (LOCA) mean those postulated accidents that result from the loss of reactor coolant at a rate in excess of the capability of the reactor coolant makeup system from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the reactor coolant system (DEGB LB-LOCA). A peculiarity of the Atucha-2 design is the positive void reactivity coefficient. This is a characteristic in common to other heavy water moderated reactors that utilize natural uranium as fuel. This implies that after a LB-LOCA event, the fission power peak at the very beginning of the transient is controlled by the void formation in the core channels, and then it is determined by the pressure wave propagation from the break. Indeed, the moderator is still liquid and flashes delayed with respect to the coolant, thus the LOCA event is also a RIA (reactivity insertion accident) event. Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and in general imply the use of complex system thermal hydraulic computer. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code such as TRANSURANUS, which can be coupled with thermal-hydraulic system codes to be used for the safety analysis. This thesis consists in the development and application of a methodology for the analysis of the 2A LB-LOCA scenario in Atucha-2 nuclear power plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis of Atucha-II NPP (Chapter 15 FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions (BIC) (e.g. pin power axial profiles) are provided by core physics and three dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method. Validation activities are performed to enhance the TRANSURANUS code capabilities and to improve the reliability of the code results. They relies on the two main sources of data, namely, specific data on Atucha-1 and/or 2, experiments or independent calculations and other data which are representative of the Atucha-2 fuel, in particular for the analysis of the normal operation and power ramp during normal operation and severe transients like LOCA and RIA

The Fukushima Event: The Outline and the Technological Background

The paper deals with the evaluation of the Fukushima-Daiichi Nuclear Power Plant (NPP) accident in Units 1 to 4: an attempt is made to discuss the scenario within a technological framework, considering precursory documented regulations and predictable system performance. An outline is given at first of the NPP layout and of the sequence of major events. Then, plausible time evolutions of relevant quantities in the different Units, is inferred based on results from the application of numerical codes. Scenarios happening in the primary circuit and containment (three Units involved) are distinguished from scenarios in spent fuel pool (four Units involved). Radiological releases to the environment and doses are approximately estimated. The event is originated by a natural catastrophe with almost simultaneous occurrence of earthquake and tsunami. These caused heavy destruction in a region in Japan much wider than the land around the NPP which was affected by the nuclear contamination. Key outcome from the work is the demonstration of strength for nuclear technology; looking at the past, misleading Probabilistic Safety Assessment (PSA) data and inadequacy in licensing processes have been found. Looking into the future keywords are Emergency Rescue Team (ERT), Enhanced Human Performance (EHP), and Robotics in Nuclear Safety and Security (RNSS)

Capabilities of TRANSURANUS Code in Simulating BWR Super-Ramp Project

After one-two years of normal operation in a LWR, the fuel–cladding gap may close, as a result of as a result of several phenomena and processes, including the different thermal expansion and swelling of both the fuel and the cladding (Pellet Cladding Interaction). In this equilibrium state, a significant increase of local power (like a transient power ramp, i.e. power increase in the order of 100kW/m-h), induces circumferential stresses in the cladding. In presence of corrosive fission products (i.e. iodine) and beyond specific stress threshold, material dependent, cracks typical of stress corrosion may appear and grow-up: this phenomenon is called stress corrosion cracking (SCC). The cracks of the cladding may spread out from the internal surface, causing the fuel failure. The objective of the activity (performed in the framework of the IAEA CRP FUMEX III), is to validate the TRANSURANUS models relevant in predicting the fuel failures due to PCI/SCC during power ramps. Focus is given on the main phenomena, which are involved or may influence the cladding failure behavior. The database selected is the Studsvik BWR Super-Ramp Project, which belongs to the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments (IFPE) database” by OECD/NEA. It comprises the data of sixteen BWR fuel rods, that have been modeled and simulated with suitable input decks. The burn-up values range between 28 and 37 MWd/kgU. Eight rods, of KWU standard type, are subjected to fast ramps, the remaining rods experience slow ramps and are of standard GE type

Nuclear Fuel Modelling During Power Ramp

Fuel rods operating for several years in a LWR can experience fuel-cladding gap closure as a result of the phenomena due to temperature and irradiation. Local power increase induces circumferential stresses in the cladding as a result of the different expansion in the cladding and the pellet. In presence of corrosive fission products (i.e. Iodine) and beyond specific stress threshold and level of burnup, cracks may grow-up from the internal to the external cladding surface, causing fuel rod failure. The phenomenon, known as pellet cladding interaction-stress corrosion cracking PCI/SCC, or PCI, has been identified as a problem since the 70\u27s. The PWR Super-Ramp experiment (part of OECD/NEA “International Fuel Performance Experiments (IFPE) database”) twenty eight fuel rods behaviour has been simulated using TRANSURANUS code version “v1m1j11”. Two sets (“Reference” and “Improved”) of suitable input decks modelling the fuel rods, based on the available literature are used to run the simulations. Focus is given to the main phenomena which are involved or may influence the cladding failure. Systematic comparison of the code results with the experimental data are performed for the parameters relevant for the PCI phenomenon. Sensitivity calculations on fission gas release models implemented in TRANSURANUS code are also performed in order to address the impact on the results. The results show the ability of TRANSURANUS version “v1m1j11” in conservatively predicting the rods failure due to PCI in PWR fuel and Zircaloy-4 cladding. Increased availability of experimental data would help to perform a deeper analysis

Proceedings of the Fifth Italian Conference on Computational Linguistics CLiC-it 2018

On behalf of the Program Committee, a very warm welcome to the Fifth Italian Conference on Computational Linguistics (CLiC-­‐it 2018). This edition of the conference is held in Torino. The conference is locally organised by the University of Torino and hosted into its prestigious main lecture hall “Cavallerizza Reale”. The CLiC-­‐it conference series is an initiative of the Italian Association for Computational Linguistics (AILC) which, after five years of activity, has clearly established itself as the premier national forum for research and development in the fields of Computational Linguistics and Natural Language Processing, where leading researchers and practitioners from academia and industry meet to share their research results, experiences, and challenges

The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios

The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature

Application of best-estimate thermal-hydraulic codes for the safety evaluation of research reactors

The Best Estimate approach in nuclear thermal-hydraulics has been developed within the framework of system codes application to large Nuclear Power Plants for electricity production. To this aim a wide range code validation program has been undertaken by the international scientific community. More recently the validation range and the applicability of those codes has been extended to cover the conditions, namely geometry, operating pressure and other fluid related parameters of Research Reactors (RR). This paper deals with an overview of needs in research reactor technology and with the description of results from selected applications of the codes to identified RR