Studsvik R2 Materials Test Reactor Ad Hoc Depletion Strategy for the Derivation of the Fuel Isotopic Composition of the MPCMIV Benchmark

L. Giaccardi ^a, S. Di Pasquale ^a, S. Dulla ^b, M. Cherubini ^a and A. Petruzzi ^a

^aNuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

^bPolitecnico di Torino, Dipartimento Energia, NEMO group, Corso Duca degli Abruzzi 24, 10129 Torino, Italy

International Conference on Physics of Reactors 2022 (PHYSOR 2022)
Pittsburgh (PA), USA, May 15–20, 2022

Abstract — The Ad Hoc Depletion Strategy elaborated by the NINE company, developed in support of the organization of the MPCMIV (Multi-physics Pellet Cladding Mechanical Interaction Validation) benchmark input and output specifications, is presented. This work aims at illustrating the strategy itself and then showing the results obtained with its application over the Studsvik R2 Testing Reactor, which is analyzed in the benchmark. The objective of the application of the strategy is to compute the fuel elements isotopic compositions at the beginning of some core loadings of interest for the benchmark. To this objective, it is necessary to implement first the simulation model of the three single assembly types and perform the infinite lattice depletions, then, to build the full core model and to perform the simulation of the core cycle. All the models and simulations were carried out with the use of the Monte Carlo particle transport code Serpent 2. Finally, the simulations results are assessed against Studsvik isotopic compositions of the fuel elements discharged from the R2 Reactor at the end of the core loading. Several assumptions were necessary during all the steps of the strategy, to overcome the lack of information regarding the core management. For this reason, the solution found at the end of the current analysis may not be completely optimized and further improvements regarding the model assumptions will be tested in a future work.

KEYWORDS: MPCMIV, Serpent 2, infinite lattice depletion, Core cycle, R2 Testing Reactor

Analysis of the Reactivity Effects Exercises of the Neutronics Benchmark of the CEFR Start-Up Tests

S. Di Pasquale ^a, M. Cherubini ^a, A. Petruzzi ^b and V. Giusti ^c

^aNuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
^bDipartimento di Ingegneria Civile ed Industriale (DICI), Università di Pisa, Italy, Largo Lucio Lazzarino (accanto all’edificio C), 56122 Pisa, Italy

International Conference on Physics of Reactors 2022 (PHYSOR 2022)
Pittsburgh (PA), USA, May 15–20, 2022

Abstract — The “Neutronics Benchmark of the CEFR Start-Up Tests” is an IAEA coordinated research project based on the simulation of the CEFR start-up tests. The main goals of the project are to improve the participant capabilities in SFR analysis and to perform an international validation of codes for Sodium Fast Reactor simulation. NINE-UNIPI work together on the creation of the Serpent 2 model and on the simulation of all the start-up tests proposed in the benchmark. In this work the three experiments related to the reactivity measurements are discussed. The geometry model is briefly described and the simulation set-up is presented. In particular, the geometry has been modeled considering the thermal expansion at the experimental temperatures. The nuclear data libraries used are the ENDF/B-VIII.0, pre-processed at the experimental temperatures and provided to the benchmark participants from SCK-CEN. The obtained results show a good agreement with the experimental data, except for the assembly-swap reactivity effect, which shows a small shift for all the considered cases. The results presented in this work could contribute to the validation of Serpent 2 for SFR criticality calculations.

KEYWORDS: CEFR. Serpent 2, SFR, Start-Up Tests, Validation

Thermal-Hydraulics Analysis of the IAEA CRP FFTF LOFWOS Test #13

Domenico De Luca, Kaiyue Zeng, Marco Cherubini, Alessandro Petruzzi

Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022

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. Within this framework, the IAEA organized a Coordinated Research Projects (CRP) on FFTF Loss of Flow Without Scram (LOFWOS) Test #13, aimed at improving Member States’ fast reactor analytical simulation capabilities, international validation, and qualification of codes currently employed in the field of fast reactor. The Fast Flux Test Facility (FFTF) was a 400 MW thermal powered, oxide-fueled, liquid sodium cooled test reactor built to assist development and testing of advanced fuels and materials for fast breeder reactors. The present paper shows the work performed by NINE for the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the FFTF. In particular, a detailed nodalization was developed following the NEMM (NINE Evaluation Model Methodology) already applied for LWR safety analysis. After achievement of acceptable steady-state results, transient analysis was performed. In addition, the NINE validation procedure was adopted in order to validate the Simulation Model (SM) against the experimental data. Two system thermal-hydraulic codes, namely RELAP5 and TRACE, were used to analyse the selected test and the comparison between the two SM results is also presented in this paper. The final goal of the activity is to present the main outcomes achieved through the use of codes currently employed in the field of fast reactor, and how the application of the NEMM procedures allows to develop and qualify the SM results and validate the computer codes against experimental data.

MELCOR-To-MELCOR Coupling Method in Severe Accident Analysis Involving Core and Pent Fuel Pool

Hector Lopez, Alessandro Petruzzi, Walter Giannotti, Domenico De Luca

Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022

Abstract — A lot of effort has been spent to prevent the occurrence of SA in nuclear plant and to develop Severe Accidents (SA) Management to mitigate the consequences of a SA. Those consequences are mainly related to limit the release of fission product to the environment. The core in the vessel is not the only source of fission products as the Spent Fuel Pool (SFP) hosting the fuel removed by the core is, in some NPP, inside the containment and SA conditions can also occur. This is especially important in reactors having proximity between the RPV and SFP such as the VVER-1200. This close proximity implies that any SA occurring in the SFP potentially affects the RPV and vice-versa. This potential combination might cause unexpected evolution in the SA progression to whom the safety systems are not able to contain. MELCOR code is a widely used, flexible powerful SA code but it is incapable (due to the uniqueness of the COR package use inside the same input) to reproduce a situation in which both the fuel in vessel core and the fuel in the SFP, inside the same containment, are going to experience a severe accident scenario. The current study presents a MELCOR-to-MELCOR coupling method to simulate simultaneously scenarios with both, core and SFP, as sources capable of H2 generation, fuel damage and FP release in a VVER-1200 NPP. The coupling is performed by running two simulations in parallel and with the data exchange supervised and managed by a dedicated Python coupling supervising script developed at NINE.