SCALING ANALYSIS IN BEPU LICENSING OF LWR

FRANCESCO D’AURIA,MARCO LANFREDINI,NIKOLAUS MUELLNER 한국원자력학회 2012 Nuclear Engineering and Technology Vol.44 No.6

“Scaling” plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a “roadmap” to scaling. Key elements are the “scaling-pyramid”, related “scaling bridges” and a logical path across scaling achievements (which constitute the “scaling puzzle”). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the “key-to-scaling”, in the licensing process of a nuclear power plant. The proposed “road map to scaling” aims at solving the “scaling puzzle”, by introducing a unified approach to the problem.

STATE OF THE ART IN USING BEST ESTIMATECALCULATION TOOLS IN NUCLEAR TECHNOLOGY

FRANCESCO D'AURIA, ANIS BOUSBIA-SALAH, ALESSANDRO PETRUZZI, ALESSANDRO DEL NEVO 한국원자력학회 2006 Nuclear Engineering and Technology Vol.38 No.1

SCALING ANALYSIS IN BEPU LICENSING OF LWR

D'auria, Francesco,Lanfredini, Marco,Muellner, Nikolaus Korean Nuclear Society 2012 Nuclear Engineering and Technology Vol.44 No.6

"Scaling" plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a "roadmap" to scaling. Key elements are the "scaling-pyramid", related "scaling bridges" and a logical path across scaling achievements (which constitute the "scaling puzzle"). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the "key-to-scaling", in the licensing process of a nuclear power plant. The proposed "road map to scaling" aims at solving the "scaling puzzle", by introducing a unified approach to the problem.

STATE OF THE ART IN USING BEST ESTIMATE CALCULATION TOOLS IN NUCLEAR TECHNOLOGY

D'AURIA FRANCESCO,ANIS BOUSBIA-SALAH,PETRUZZI ALESSANDRO,NEVO ALESSANDRO DEL Korean Nuclear Society 2006 Nuclear Engineering and Technology Vol.38 No.1

System thermal-hydraulic codes have been used in the past decades in the areas of design, operation, licensing and safety of Nuclear Power Plants (NPPs). The development and validation of these codes have reached a high degree of maturity, through the consideration of huge experiments and advanced numerical models. Nowadays, the analyses are based upon realistic approaches rather than the conservative evaluation models. However the applications of these computational tools require preliminary qualification issues. Although huge amounts of financial and human resources have been invested for the development and improvement of codes, the calculation results are still affected by errors. In the sophisticated nuclear technology, design and safety of NPP, these errors must be quantified. An overview of the state of the art of the current thermal-hydraulic system code is developed and the need of uncertainty analysis in code calculations is emphasized. Several sources of uncertainty have been classified and commented, and typical applications of such methods are shown.

INTERNATIONAL STANDARD PROBLEM 50: THE UNIVERSITY OF PISA CONTRIBUTION

Cherubini, Marco,Lazzerini, Davide,Giannotti, Walter,D'auria, Francesco Korean Nuclear Society 2012 Nuclear Engineering and Technology Vol.44 No.6

The present paper deals with the participation of the University of Pisa in the last International Standard Problem (ISP) focused on system thermal hydraulic, which was led by the Korean Atomic Energy Research Institution (KAERI). The selected test was a Direct Vessel Injection (DVI) line break carried out at the ATLAS facility. University of Pisa participated, together with other eighteen institutions, in both blind and open phase of the analytical exercise pursuing its methodology for developing and qualifying a nodalization. Qualitative and quantitative analysis of the code results have been performed for both ISP-50 phases, the latter adopting the Fast Fourier Transfer Based Method (FFTBM). The experiment has been characterized by three-dimensional behavior in downcomer and core region. Even though an attempt to reproduce these phenomena, by developing a fictitious three-dimensional nodalization has been realized, the obtained results were generally acceptable but not fully satisfactory in replicating 3D behavior.

Steady-State and Stability behavior of a Single-Phase Natural Circulation Loop

Mario Misale,Monica Frogheri,Paolo Ruffino,Francesco D`Auria 대한기계학회 1998 Heat transter conference Vol.3 No.1

INTERNATIONAL STANDARD PROBLEM 50: THE UNIVERSITY OF PISA CONTRIBUTION

MARCO CHERUBINI,DAVIDE LAZZERINI,WALTER GIANNOTTI,FRANCESCO D’AURIA 한국원자력학회 2012 Nuclear Engineering and Technology Vol.44 No.6

The present paper deals with the participation of the University of Pisa in the last International Standard Problem (ISP)focused on system thermal hydraulic, which was led by the Korean Atomic Energy Research Institution (KAERI). The selected test was a Direct Vessel Injection (DVI) line break carried out at the ATLAS facility. University of Pisa participated, together with other eighteen institutions, in both blind and open phase of the analytical exercise pursuing its methodology for developing and qualifying a nodalization. Qualitative and quantitative analysis of the code results have been performed for both ISP-50phases, the latter adopting the Fast Fourier Transfer Based Method (FFTBM). The experiment has been characterized by threedimensional behavior in downcomer and core region. Even though an attempt to reproduce these phenomena, by developing a fictitious three-dimensional nodalization has been realized, the obtained results were generally acceptable but not fully satisfactory in replicating 3D behavior.

Analysis of loss of cooling accident in VVER-1000/V446 spent fuel pool using RELAP5 and MELCOR codes

Mousavian Seyed Khalil,Shirani Amir Saeed,Francesco D'Auria 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.8

Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the high content of long-lived radionuclides and lack of protection by the pressure vessel. In this study, the loss-of-cooling accident (LOFA) for the VVER-1000/V446 spent fuel pool is simulated by employing RELAP5 and MELCOR 1.8.6 as the best estimate and severe accident analysis codes, respectively. For two cases with different total power levels, decay heat of spent fuels is calculated by ORIGEN-II code. For modeling SFP of a VVER-1000, a qualified nodalizations are considered in both codes. During LOFA in SFP, the key sequences such as heating up of the pool water, boiling and reducing the water level, uncovering the spent fuels, increasing the temperature of the spent fuels, starting oxidation process (generating Hydrogen and extra power), the onset of fuel melting, and finally releasing radionuclides are studied for both cases. The obtained results show a reasonable consistency between the RELAP5 and MELCOR codes, especially before starting the oxidation process.