G. M. Galassi

Outline of the Uncertainty Methodology Based on Accuracy Extrapolation

AbstractUncertainty methodology based on accuracy extrapolation (UMAE) is outlined. This methodology is suitable for evaluating the uncertainty in the prediction of transient scenarios in nuclear reactors when carried out by thermal-hydraulic system codes. It is based on the extrapolation of the accuracy resulting from a comparison between code results and relevant experimental data obtained in small scale facilities. A simplified diagram of the UMAE is compared with a similar one derived for the code scaling, applicability and uncertainty evaluation (CSAU) previously proposed by the U.S. Nuclear Regulatory Commission. A few results related to the full application of the UMAE to a small-break loss-of-coolant accident in a pressurized water reactor, including core uncovery, are also reported.

Code validation and uncertainties in system thermalhydraulics

Abstract We describe in this paper the state of the art in the area of thermalhydraulic system codes assessment and uncertainty evaluation. System codes have been used in the past three decades in the areas of design, operation, licensing and safety of Nuclear Power Plants (NPP). Such applications require preliminary comprehensive code-user-nodalization qualification activities that we discuss in the paper. Although huge amounts of financial and human resources have been invested for the development and improvement of codes, including the user interface, the calculation results are still affected by erros. In the sophisticated nuclear technology, design and safety of NPP, these errors must be quantified. Therefore we propose uncertainty methodologies to achieve this goal. The reported analysis is based on the activities performed at University of Pisa and the experience we have acquired from the participation in a number of international projects.

Scaling of natural circulation in PWR systems

Abstract This paper deals with the problem of scaling complex thermalhydraulic scenarios measured in experimental facilities which simulate Pressurized Water Reactor systems. Phenomena occurring during different phases of natural circulation between the core and the steam generators are considered. The experimental data obtained in some integral test facilities were analyzed with a large system code and for this purpose a simple model was built. The code predicted scenario and the extrapolated one for the actual plant are compared.

Analysis of the Peach Bottom turbine trip 2 experiment by coupled RELAP5-PARCS three-dimensional codes

Abstract Thanks to continuous progress in computer technology, it is now possible to perform best-estimate simulations of complex scenarios in nuclear power plants. This method is carried out through the coupling of three-dimensional (3-D) neutron modeling of a reactor core into system codes. It is particularly appropriate for transients that involve strong interactions between core neutronics and reactor loop thermal hydraulics. For this purpose, the Peach Bottom boiling water reactor turbine trip test was selected to challenge the capability of such coupled codes. The test is characterized by a power excursion induced by rapid core pressurization and a self-limiting course behavior. In order to perform the closest simulation, the coupled thermal-hydraulic system code RELAP5 and 3-D neutron kinetic code PARCS were used. The obtained results are compared to those available from experimental data. Overall, the coupled code calculations globally predict the most significant observed aspects of the transient, such as the pressure wave amplitude across the core and the power course, with an acceptable agreement. However, sensitivity studies revealed that more-accurate code models should be considered in order to better match the void dynamic and the cross-section variations during transient conditions.

Scaling of complex phenomena in system thermal hydraulics

A methodology is described that can be used for the extrapolation of thermal-hydraulic phenomena measured in differently scaled integral test facilities to nuclear reactor plant conditions. The use of a system code in this context is confirmed to be of fundamental importance, provided that the codes scaling capability has been demonstrated. The starting data base for the proposed study consists of the measured quantities and corresponding RELAP5/MOD2 code calculation results related to a boiling water reactor small-break loss-of-coolant accident (SBLOCA) counterpart test activity, a pressurized water reactor (PWR) natural-circulation type test activity, and a PWR SBLOCA counterpart test activity. The proof that this methodology can be used for evaluating uncertainties in predicting transient behavior in nuclear power plants is the main result of this study. Data have been obtained that give a value of the foreseeable error ranges in the provision of plant behavior in the three cases considered.

Flowrate and Density Oscillations During Two-Phase Natural Circulation in PWR Typical Conditions

Abstract Strong oscillations in fluid velocities and densities have been measured in the LOBI PWR simulator during natural circulation tests. A sort of ‘syphon condensation’ occurs in the loop when the primary side mass inventory roughly equals 70% of the initial value. RELAP5/MOD2 code predicts oscillations of the same quantities with similar frequency. This report deals with the analysis of the above physical phenomena even from the code calculation point of view.

Characterization of instabilities during two-phase natural circulation in typical PWR conditions

Abstract Strong oscillations in fluid velocities and densities were measured in the LOBI test facility during natural circulation experiments. A sort of “siphon condensation” occurs in the U-tubes of steam generators when primary-side mass inventory reaches roughly 64% of the initial value. The present paper deals with the characterization of the phenomenon considering flooding and condensation dynamics of U-tubes; RELAP5/MOD2 calculations made it possible to select system parameters affecting the oscillation characteristics. An attempt was made to evaluate the possibility of instability occurring in real plant situations.

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).

Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

Abstract A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock & Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark. Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference: 1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling 2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling) 3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code. The influence of PVM and of direct coupling is also discussed. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper. The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some “recriticality” or “return to power” whose magnitude is largely affected by boundary and initial conditions.

Analyses of PACTEL passive safety injection experiments with APROS, CATHARE and RELAP5 codes

Abstract The European Commission fourth framework programme project ‘Assessment of passive safety injection systems of advanced light water reactors’ involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of passive safety injection systems (PSIS) of advanced light water reactors (ALWRs) in small break loss-of-coolant accident (SBLOCA) conditions. The PSIS consisted of a core make-up tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the phenomena in the PSIS.