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International Benchmark Activity in the Field of Sodium Fast Reactors Domenico De Luca, Simone Di Pasquale, Marco Cherubini, Alessandro Petruzzi and Gianni Bruna Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy |
Abstract - Global interest in fast reactors has been growing since their inception in 1960 because they can provide efficient, safe, and sustainable energy. Their closed fuel cycle can support long-term nuclear power development as part of the world’s future energy mix and decrease the burden of nuclear waste. In addition to current fast reactors construction projects, several countries are engaged in intense R&D and innovation programs for the development of innovative, or Generation IV, fast reactor concepts. Within this framework, NINE is very actively participating in various Coordinated Research Projects (CRPs) organized by the IAEA, aimed at improving Member States’ fast reactor analytical simulation capabilities and international qualification through code-to-code comparison, as well as experimental validation on mock-up experiment results of codes currently employed in the field of fast reactors. The first CRP was focused on the benchmark analysis of Experimental Breeder Reactor II (EBR-II) Shutdown Heat Removal Test (SHRT-17), protected loss-of-flow transient, which ended in the 2017 with the publication of the IAEA-TECDOC-1819. In the framework of this project, the NINE Validation Process– developed in the framework of NEMM (NINE Evaluation Model Methodology) – has been proposed and adopted by most of the organizations to support the interpretation of the results calculated by the CRP participants and the understanding of the reasons for differences between the participants’ simulation results and the experimental data. A second project regards the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the Fast Flux Test Facility (FFTF), the Loss of Flow Without Scram (LOFWOS) Test #13, started in 2018. A detailed nodalization has been developed by NINE following its nodalization techniques and the NINE validation procedure has been adopted to validate the Simulation Model (SM) against the experimental data of the selected test. The third activity deals with the neutronics benchmark of China Experimental Fast Reactor (CEFR) Start-Up Tests, a CRP proposed by the China Institute of Atomic Energy (CIAE) launched in 2018 the main objective of which is to improve the understanding of the start-up of a SFR and to validate the fast reactor analysis computer codes against CEFR experimental data. A series of start-up tests have been analyzed in this benchmark and NINE also proposed and organized a further work package focused on the sensitivity and uncertainty analysis of the first criticality test. The present chapter intends to summarize the results achieved using the codes currently employed in the field of fast reactor in the framework of international projects and benchmarks in which NINE was involved and emphasize how the application of developed procedures allows to validate the SM results and validate the computer codes against experimental data.
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External function for GOTHIC code to estimate critical heat flux conditions for in-vessel retention assessment A. Pop a b , A. Petruzzia,W. Giannottia a Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy b Università di Pisa, Largo Lucio Lazzarino 2, Pisa, PI, Italy Nuclear Engineering and Design, Volume 380, 15 August 2021, 111301 |
Abstract - GOTHIC is an integrated, general purpose thermal hydraulic software package for design, licensing, safety and operating analysis of Nuclear Power Plant containments, confinement buildings and system components. It bridges the gap between the lumped parameter codes frequently used for containment analysis (such as MELCOR, MAAP, COCOSYS, ASTEC codes) and Computational Fluid Dynamics codes. Within a single model, GOTHIC can include regions treated in conventional lumped parameter mode and regions with three-dimensional flows in complex geometries. The heat transfer correlations built into GOTHIC cover the portion of the boiling curve which spans single phase heat transfer up to pre-Critical Heat Flux (CHF) heat transfer. The implemented boiling curve is truncated to exclude post-CHF heat transfer as it has not been adequately verified and was considered by the developers to have little application in general containment analysis. As such, one area that the code is not currently qualified for is post-CHF heat transfer, which could occur for example in the case of In-Vessel Corium Retention, where cooling water enters in contact with the high temperature of the Reactor Pressure Vessel wall. The presented research focuses on creating an external subroutine that solves this limitation, enabling the GOTHIC code to account for CHF phenomena. The modeling of CHF would be very useful in order to enable the code to simulate the external Reactor Pressure Vessel (RPV) Cooling , as well as other types of severe accidents or analyses where post-CHF simulation is required. Several subroutine function switches were implemented in order to facilitate its usage for different types of heat structures and correlations. The subroutine determines the CHF values based on either: the 2006 Groeneveld Look-up Tables, Lookup Tables for Large Diameter Vertical Tubes, Look-up Tables for Large Diameter Horizontal Tubes, or the correlation used by the MELCOR code for critical heat flux situations. It shall be noted that the developed subroutine and its implementation were performed without the need to have access to the GOTHIC source code. In a previous paper, the GOTHIC code was used to perform a containment safety analysis for the Atucha-I NPP (CNA-I) for an in-vessel retention type of analysis. Highly conservative vapour generating boundary conditions were used in order to simulate the boiling between the cavity water and the RPV surface, and to bypass the GOTHIC limitation. The newly developed subroutine was used for the analysis of two postulated Atucha-I in-vessel retention scenarios, a Large Break Loss Of Coolant Accident (LBLOCA) and a Station Black-Out (SBO), with the simulation of heat transfer between RPV and cavity water. Specifically for in-vessel retention situations, a separate user selectable option for the subroutine was developed, in which the Critical Heat Flux is determined based on experiments performed at the ULPU facility from the University of California, Santa Barbara, USA, which were combined with the 2006 Groeneveld Look-up Tables primarily in order to have a pressure dependence. This was performed because for the Atucha-I analysis, the pressure was higher than in the ULPU facility.