Nuclear Energy for Sustainable Economic Development

The discovery and the application of nuclear energy constitute the most important technological achievement of the past century. However, the development and the exploitation of this technology have been remarkably smaller than foreseeable. An overview of the significant features of the nuclear technology including the comparison with competitive energy sources is made. The “embedded” safety engineering and the pollution are discussed and the main features are mentioned. Indeed, nuclear technology can be applied for the sustainable society development by producing substantial amount of clean water from the ocean. The idea is to build up nuclear power plant sites that produce desalinated water and pump it several tens of kilometers away to form a lake into a desert region. This could help to establish the conditions for an agriculture-based civilization

User effect on code application and qualification needs

A training Workshop /Seminar was organized (F. D’Auria was lead scientist) by Organization for Economic Cooperation and Development / Nuclear Energy Agency / Committee on the Safety of Nuclear Institutions (OECD/NEA/CSNI) to preserve the knowledge and the expertise generated with the CSNI in relation to thermal-hydraulics and related applications to Nuclear Power Plant (NPP) safety and design analyses. The Workshop Seminar was titled THICKET ‘Seminar on Transfer of Competence, Knowledge and Experience gained through CSNI activities in the field of Thermal-hydraulics’. This was a first of Series (other followed in 2008, 2012 and 2016). A group f about ten invited specialists (F. D’Auria part of the group) was presenting lectures at the Seminar. The lectures are supported by ‘.ppt’ presentation/file and by ‘.pdf’ paper-format material issued in thick books-reports and in electronic format. The material was distributed to participants. The present lecture deals with user effect upon the results of calculations performed by system thermal-hydraulic codes in application to accident analyses of NPP

Accident analysis in research reactors

The incomplete understanding of the complex mechanisms connected with the interaction between thermal-hydraulic and neutron kinetics stili chailenges the design and the operation of nuclear reactors and imposes the adoption of conservatism in the evaluation of safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience suggests the revisiting of those areas and the identification of designloperation requirements that can be relaxed. So far, almost all of the safety analyses of research reactors have been performed using conservative computational tools such as channel codes but, nowadays, the appiication of Best-Estimate (BE) methods constitutes a rea! necessity. The global aim of the current work is an attempt to apply the best-estimate system thermalhydraulic code Relap5. For this purpose, the generic IAEA research reactor Benchmark problem is re-considered for proving the adequacy of the available computationai tools. Within the same framework, one of the most severe accident categories that may occur during a research reactor iifetime is aiso considered. This is related to a total and partiai blockage of the cooling charme! of a singie Fuel Assembly. Such event constitutes a stern scenario for this type of reactor since it may lead to local dryout and eventually to the bss of the fuei assembly integrity. The study constitutes the first step of a !arger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. To demonstrate the suitabi!ity of the technique, the bss of Shutdown Heat Removal accident in a MTR poo! type research reactor is ana!ysed. The accident occurs when the passive shutdown naturai convection coo!ing system is fai!ing for instance due to the rupture of an experimenta! beam tube. The accident will !ead to a partia! core uncovering. Although most of the research investigations in the world were performed for the analysis of the natura! air coo!ed research reactor core, it is demonstrated that there is a power density may exist above which partial submergence causes higher temperature than no submergence at all

The BWR stability issue

A training Workshop /Seminar was organized (F. D’Auria was lead scientist) by Organization for Economic Cooperation and Development / Nuclear Energy Agency / Committee on the Safety of Nuclear Institutions (OECD/NEA/CSNI) to preserve the knowledge and the expertise generated with the CSNI in relation to thermal-hydraulics and related applications to Nuclear Power Plant (NPP) safety and design analyses. The Workshop Seminar was titled THICKET ‘Seminar on Transfer of Competence, Knowledge and Experience gained through CSNI activities in the field of Thermal-hydraulics’. This was a first of Series (other followed in 2008, 2012 and 2016). A group f about ten invited specialists (F. D’Auria part of the group) was presenting lectures at the Seminar. The lectures are supported by ‘.ppt’ presentation/file and by ‘.pdf’ paper-format material issued in thick books-reports and in electronic format. The material was distributed to participants. The present lecture deals with Boiling Water Reactor (BWR) stability and the occurred/monitored instability events in operating reactor

Analysis of the Peach Bottom BWR turbine trip pressure wave propagation and its effect on the kinetic core dynamics

The modeling of complex transients in Nuclear Power Plants remain a challenging topics for Best Estimate computational tools. Nowadays, such calculations are performed through the so called coupled code method, which consists in incorporating three-dimensional (3D) neutron modeling of reactor core into system codes. This technique is extensively used for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. In this framework the Peach Bottom BWR 2 Turbine Trip test is considered since it involves a rapid (water hammer transient) pressure wave induced positive reactivity addition into the core. To perform a numerical simulation of such phenomenon a reference code case was run and an overall data comparison was carried out. The results show good agreements between the calculations and most of the significant global aspects observed experimentally. However, the test was revealed very sensitive to the feedback modeling and requires a tightly accurate simulation of the thermal-hydraulic and the cross sections parameters. For this purpose, sensitivity studies have been carried out in order to identify the most influent parameters that govern the transient behavior

Analysis of extreme scenarios and sensitive cases of the Peach Bottom 2 BWR Turbine Trip by coupled Relap5/Parcs

A research activity started at the end of 80’s was substantiated within the context of the participation of the TMI-1 (Three Mile –Unit 1) Benchmark in 1998. A coupled thermal-hydraulic neutron physics code was developed. This was used for the Boiling Water Reactor (BWR) Turbine Trip (TT) experiment planned and executed in the Peach Bottom Nuclear Power Plant (NPP) in the USA. The test was originated by the quick closing of the turbine inlet valve, followed by the quick opening of the condenser bypass line and, in the core region, by the arrival of a pressure wave. The pressure wave cause void collapse and increase of fission core power for a factor 5 in a few seconds after the turbine trip. The overall scenario was successfully reproduced by the coupled Relap5/Parcs code. This document discusses the results of additional accident scenarios in the ‘now-qualified’ Peach Bottom NPP scenario. The analysis demonstrated, among the other things, the role and the importance of SRV (Steam Relief Valves) opening in preventing code damage

NEA (Nuclear Energy Agency) Benchmarks

A training Workshop /Seminar was organized (F. D’Auria was lead scientist) by Organization for Economic Cooperation and Development / Nuclear Energy Agency / Committee on the Safety of Nuclear Institutions (OECD/NEA/CSNI) to preserve the knowledge and the expertise generated with the CSNI in relation to thermal-hydraulics and related applications to Nuclear Power Plant (NPP) safety and design analyses. The Workshop Seminar was titled THICKET ‘Seminar on Transfer of Competence, Knowledge and Experience gained through CSNI activities in the field of Thermal-hydraulics’. This was a first of Series (other followed in 2008, 2012 and 2016). A group f about ten invited specialists (F. D’Auria part of the group) was presenting lectures at the Seminar. The lectures are supported by ‘.ppt’ presentation/file and by ‘.pdf’ paper-format material issued in thick books-reports and in electronic format. The material was distributed to participants. The present lecture deals with description of NEA Benchmarks (NPP applications) involving the coupling between thermal-hydraulics and neutron physics codes

Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code

Nowadays, the coupled codes technique, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for carrying out best estimate (BE) simulation of complex transient in nuclear power plants (NPP). This technique is particularly suitable for transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Such complex interactions are encountered under normal and abnormal operating conditions of a boiling water reactors (BWR). In such reactors Oscillations may take place owing to the dynamic behavior of the liquid- steam mixture used for removing the thermal power. Therefore, it is necessary to be able to detect in a reliable way these oscillations. The purpose of this work is to characterize one aspect of these unstable behaviors using the coupled codes technique. The evaluation is per- formed against Peach Bottom-2 low-flow stability tests number 3 using the coupled RELAP5/PARCS code. In this transient dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein assessed and compared against the available experimental data