Francesco Saverio D'Auria

University of Pisa

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G. M. Galassi

University of Pisa

A. Petruzzi

Engineering (Italy)

O. Mazzantini

Comisión Nacional de Energía Atómica

A Bousbia Salah

Federal Agency for Nuclear Control

H. Glaeser

Gesellschaft für Anlagen und Reaktorsicherheit

N. Muellner

BOKU University

F. Moretti

University of Pisa

Carlo Parisi

Idaho National Laboratory

P. Vigni

University of Pisa

W. Giannotti

IMT School for Advanced Studies Lucca

Cooperative Institutions

University of Pisa

101

Commissariat à l'Énergie Atomique et aux Énergies Alternatives

Korea Atomic Energy Research Institute

Lappeenranta-Lahti University of Technology

National Agency for New Technologies, Energy and Sustainable Economic Development

Gesellschaft für Anlagen und Reaktorsicherheit

Helmholtz-Zentrum Dresden-Rossendorf

CEA Grenoble

National Nuclear Energy Commission

Direction des énergies

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Analysis of the Peach Bottom Turbine Trip 2 Experiment by Coupled RELAP5-PARCS Three-Dimensional Codes

Nuclear Science and Engineering (2004)

A Bousbia Salah J. Vedovi Francesco Saverio D'Auria Kostadin Ivanov G. M. Galassi

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.

Neutron Transport

Thermal hydraulics

Boiling water reactor

Transient (computer programming)

Hydraulics

10.13182/nse04-a2462

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Citations (25)

Thermo-Hydraulic Calculation and Safety Analysis of the Tehran Research Reactor

Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance (2009)

Jalil Jafari M. K. Firuzjaee Samad Khakshournia Kamran Sepanloo Francesco Saverio D'Auria

The aim of this study is to evaluate the thermal hydraulic parameters such as, Minimum DNBR, fuel, clad and coolant temperatures of the Tehran Research Reactor (TRR) core in order to assure that the thermal limitations are not exceeded at full power and steady-state and the loss-of-flow accident (LOFA) operating conditions and the existing design is adequate to assure that the consequences from this anticipated occurrence does not lead to a severe accident. The PARET and TRANS computer codes were used to analysis thermal-hydraulic and safety parameters under study steady-state conditions. Furthermore the LOFA as a consequence of two type of transients are analyzed; one related to main cooling pump failure accident (slow LOFA), and the other related to accidentally opening the flapper valve whereas the main cooling pump is in working (fast LOFA). The steady-state analysis involves the determination of the departure from nucleate boiling ratio (DNBR), the temperatures profile, heat flux across the hottest channel and etc. The LOFA analysis involves the power and coolant flow rate, the fuel centerline, clad surface and coolant outlet temperatures, and the occurring of coolant flow reversal.

Thermal hydraulics

Decay heat

10.1115/icone17-75484

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Citations (0)

Basic Scaling Principles Adopted for the Simulation of Behaviour during Small LOCAs

Francesco Saverio D'Auria M. Mazzini P. Vigni

Source

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Numerical effects in the prediction of single phase natural circulation stability

Walter Ambrosini Francesco Saverio D'Auria A Pennati J. C. Ferreri

The paper discusses the effects of numerical discretisation in the prediction of single-phase natural circulation as observed in a simple experimental thermosyphon loop. Stability of steady-state conditions is particularly addressed, making use of a calculation model purposely set up for studying natural circulation in loops having general shape and distribution of heat sources and sinks. The model is equipped with both a classical upwind explicit first order numerical scheme and a low diffusion scheme devised on the basis of the Warming-Beam second order upwind explicit method. The results obtained by the model making use of different degree of detail in spatial discretisation, different time steps and numerical methods are compared with the predictions of a thermal-hydraulic system code making use of similar nodalisations. The main emphasis of the discussion concerns the effects of first order numerical discretisation as pointed out by both calculation models, which make the prediction of the dynamic behaviour of the considered loop a rather challenging problem. The need and the feasibility of adopting higher order numerical methods for studying unsteady single-phase natural circulation is one of the outcomes of the analysis.

Natural circulation

Numerical diffusion

Upwind scheme

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Citations (6)

The role of nuclear thermal-hydraulics in the licensing of Atucha-II: The LBLOCA

Nuclear Engineering and Design (2019)

O. Mazzantini G. M. Galassi Francesco Saverio D'Auria

Pressurized water reactor

Cabin pressurization

Thermal hydraulics

Loss-of-coolant accident

Natural circulation

10.1016/j.nucengdes.2019.05.025

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Citations (3)

SCALING ANALYSIS IN BEPU LICENSING OF LWR

Nuclear Engineering and Technology (2012)

Francesco Saverio D'Auria M. Lanfredini N. Muellner

Thermal hydraulics

Neutron Transport

10.5516/net.02.2012.711

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Citations (2)

Sensitivity analysis of the Peach Bottom Turbine Trip 2 experiment

Kerntechnik (2004)

A Bousbia Salah Francesco Saverio D'Auria M. Bambara

Abstract The recent availability of powerful computer and computational techniques has enlarged the possibilities to perform best estimate simulations of complex scenarios in nuclear power plants. Nowadays, the coupled codes method, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used. It is particularly suited for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Within this framework the Peach Bottom BWR 2 Turbine Trip test was selected since it involves a rapid pressure induced positive reactivity addition into the core. It is also characterized by a self-limiting power course due to compensated inherent reactivity mechanisms. To perform a numerical simulation of the turbine trip a reference case was run for the coupled thermal-hydraulic system code RELAP5/mod3.3 and 3D neutron kinetic PARCS/2.3 code. The overall data comparison shows good agreements between the calculations and most of the significant global aspects observed experimentally. However, the test is 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 influential parameters that govern the transient behavior. The considered cases showed that the self-limiting power amplitude as predicted by the coupled code calculation is mainly due to delayed feedback mechanisms whereas the experimental data shows that the power quenching before the Scram is governed by prompt feedback effects.

Thermal hydraulics

Neutron Transport

Scram

Core model

Transient (computer programming)

Hydraulics

10.3139/124.100182

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Citations (3)

UNIPI Blind Calculation Results of SP-2 and SP-3 by RELAP5-3D Code

A. Kovtonyuk Francesco Saverio D'Auria

Code (set theory)