Teresa Barrachina

New methodology for analysing and increasing the cost-efficiency of environmental monitoring networks.

This study focuses on the coastal monitoring network established in the scope of WFD implementation. The objective of this network was to provide an ecological assessment of Valencian coastal waters. After three years, sufficient data had been collected to enable us to analyse and explore ways to increase the networks efficiency. A methodology was developed to select the best subset of sampling stations to be surveyed. This method was approached from the perspective of an inter-observer variability problem. In order to compare the concordance between the k-observers and the reference observer, two measures were considered: euclidean distance, and interclass correlation coefficient. The obtained results confirm that the current network can be reduced by over 50% and still guarantee accurate results. This methodology (not limited by indicators, geographically, or by type of water body) could be applied to different environmental monitoring networks and could significantly decrease the efforts and costs required by the WFD.

Improvements in the decay heat model in the thermalhydraulic code TRAC-BF1

The implemented models in TRAC-BF1 for the decay heat calculations are based on the 1971 American Nuclear Society (ANS) Standard or the 1979 ANS Standard if selected by the user. With the entry into force of the 1994 ANS Standard and the subsequent review, the TRAC-BF1 models were made completely obsolete, so a revision of the older models and the implementation of ANSI/ANS-5.1-2005 Standard in the code are required. In this paper, a novel numerical study is presented based on the analytical resolution of the decay heat equation which takes into account the number of reactor operation histories, the number of fissionable nuclides, and the number of groups per fissile. Moreover, this paper describes the influence of the short-term power histories on the total decay heat power calculation due to the high interest in the simulation of severe transients, such as Anticipated Transients without SCRAM, which cannot be ignored in nuclear safety analysis.

Methodology to resolve the transport equation with the discrete ordinates code TORT into the IPEN/MB-01 reactor

Resolution of the steady-state Neutron Transport Equation in a nuclear pool reactor is usually achieved by means of two different numerical methods: Monte Carlo (stochastic) and Discrete Ordinates (deterministic). The Discrete Ordinates method solves the Neutron Transport Equation for a set of selected directions, obtaining a set of directional equations and solutions for each equation which are the angular flux. In order to deal with the energy dependence, an energy multi-group approximation is commonly performed, obtaining a set of equations depending on the number of energy groups. In addition, spatial discretization is also required and the problem is solved by sweeping the geometry mesh. However, special cross-sections are required due to the energy and directional discretization, thus a methodology based on NJOY99 code capabilities has been used. Finally, in order to demonstrate the capability of this method, the 3D discrete ordinates code TORT has been applied to resolve the IPEN/MB-01 reactor.

Validation of 3D neutronic-thermalhydraulic coupled codes RELAP5/PARCSv2.7 and TRACEv5.0P3/PARCSv3.0 against a PWR control rod drop transient

ABSTRACT In nuclear safety field, neutronic and thermalhydraulic codes performance is an important issue. New capabilities implementation, as well as models and tools improvements are a significant part of the community effort in looking for better nuclear power plants (NPP) designs. A procedure to analyze the PWR response to local deviations on neutronic or thermalhydraulic parameters is being developed. This procedure includes the simulation of Incore and Excore neutron flux detectors signals. A control rod drop real plant transient is used to validate the used codes and their new capabilities. Cross-section data are obtained by means of the SIMTAB methodology. Detailed thermalhydraulic models were developed: RELAP5 and TRACE models simulate three different azimuthal zones. Besides, TRACE model is performed with a fully three-dimensional core, thus, the cross-flow can be obtained. A Cartesian vessel represents the fuel assemblies and a cylindrical vessel the bypass and downcomer. Simulated detectors signals are obtained and compared with the real data collected during a control rod drop trial at a PWR NPP and also with data obtained with SIMULATE-3K code.

EXPERIENCES IN DEVELOPING AND APPLYING A NEW METHODOLOGY IN MASTER’S DEGREE

In the subject Advanced Energy and Thermal Machines of the Master’s Degree in Industrial Engineering a great effort was made to apply a new methodology based on the development of activities directly related to the real or professional applications together with the use of computer codes to solve these real problems. The simulation based learning has been the basis of the application of this new methodology. The goal is that students can achieve a high level of theory knowledge together with a high capacity of autonomy in solving problems. For that, a plan has been stablished to be implemented gradually. According to this plan, the subject is clearly defined in each of the sections: objectives, contents, methodology, activities and assessment looking for the constructive alignment. Besides, new materials have been created. In this paper, the experiences and results obtained in the process of improving the way of teaching this subject are shown and analyzed from the point of view of the teacher and from the point of view of the student.

Rod Ejection Accident 3D-dynamic analysis in trillo NPP with RELAP5/PARCS V2.7

The Rod Ejection Accident (REA) belongs to the Reactivity-Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In the present work, we have analyzed this transient in Trillo NPP at different power conditions at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5MOD3.3/PARCSv2.7. The transient departs from an initially critical core, being the ejection speed of the control rod a typical bounding value. The simulation includes the SCRAM signal in order to compare our best-estimate results with the results from the conservative calculations provided by the NPP. These analyses will allow knowing more accurately characterize the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods.

Experimental orthogonal functions for the qualification of BWR stability events. Application to Peach Bottom NPP

In this work, BWR stability analysis was performed on an operating point (PT_UPV) of Peach Bottom NPP which is inside the exclusion region. The simulation was made with the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved departing from test point 3 by a control rod movement as it is usually performed in Nuclear Power Plants. The transient starts with this control rod movement. The purpose of this study is to qualify this coupled code against this kind of 3D complex accidents that take place inside the core. The calculated results show that point PT_UPV is an unstable point and the obtained relative axial power distribution shows a bottom-peaked profile, which is characteristic of unstable cores. After control rod movement, a limit cycle in-phase oscillation on the total reactor power evolution is obtained, together with a coupled out-of-phase oscillation in the whole 3D power evolution. In order to complete the stability analysis, the local power range monitors (LPRMs) simulated signals obtained from the neutronic code has been analyzed using the singular system analysis. The results confirm the out-of-phase power oscillation in the core.

A New Methodology to Obtain the 1D Cross-Sections for TRAC-BF1 Code: Application to Peach Bottom NPP

In TRAC-BF1 the cross-sections are specified in the input deck in a polynomial form. Therefore, it is necessary to obtain the coefficients of this polynomial expansion. One of the methods proposed in the literature is the KINPAR methodology developed in the UPV. This methodology uses the results from different perturbations of the original state to obtain the coefficients of the polynomial expansion. The simulations are performed using the SIMULATE3 code. In this work, a new methodology called SIMTAB-1D to obtain the cross-sections sets in 1D is presented. The first step consists of the application of the SIMTAB methodology, developed in the UPV, to obtain the 3D cross-sections sets from CASMO4/SIMULATE3. These 3D cross-sections sets are collapsed to 1D, using as a weighting factor, the 3D thermal and rapid neutron fluxes obtained from SIMULATE3. This new methodology will be applied to the simulation of the turbine trip transient in Peach Bottom NPP using the TRAC-BF1 code. The results of the steady state in TRAC-BF1 using the KINPA R methodology and the new methodology are compared with the reference SIMULATE3 results.Copyright