• Energy Research

The paper presents an investigation of the potential for the recriticality phenomenon occurrence during core reflooding at the time of early in-vessel phase of the Fukushima-Daiichi Unit 3 (1F3) accident. A Monte Carlo criticality analysis was performed using the SERPENT code for STEP-3 BWR fuel type. The analysis was conducted for a representative core unit cell composed of four assemblies in three dimensions. The MELCOR computer code was used as the source of the accident progression and thermal-hydraulic input. A data exchange framework between MELCOR and SERPENT was developed and employed. The calculations reveal that for the applied MELCOR model, recriticality occurs due to simultaneous control blades loss, fuel rods intactness and non-borated water injection. It suggests that during reflooding even a relatively small fraction of the core covered with water and without control blades is sufficient to lead to the critical condition. Additionally, a sensitivity analysis for various control blade survival fractions was performed.

A nuclear reactor based on MIT BEAVRS benchmark was used as a typical power generating Pressurized Water Reactor (PWR). The PARCS v3.2 nodal-diffusion core simulator was used as a full-core reactor physics solver to emulate the operation of a reactor and to generate training, and validation data for the ANN. The ANN was implemented with dedicated Python 3.8 code with Google's TensorFlow 2.0 library. The effort was based to a large extent on the process of appropriate automatic transformation of data generated by PARCS simulator, which was later used in the process of the ANN development. Various methods that allow obtaining better accuracy of the ANN predicted results were studied, such as trying different ANN architectures to find the optimal number of neurons in the hidden layers of the network. Results were later compared with the architectures proposed in the literature. For the selected best architecture predictions were made for different core parameters and their dependence on core loading patterns. In this study, a special focus was put on the prediction of the fuel cycle length for a given core loading pattern, as it can be considered one of the targets for plant economic operation. For instance, the length of a single fuel cycle depending on the initial core loading pattern was predicted with very good accuracy (>99%). This work contributes to the exploration of the usefulness of neural networks in solving nuclear reactor design problems. Thanks to the application of ANN, designers can avoid using an excessive amount of core simulator runs and more rapidly explore the space of possible solutions before performing more detailed design considerations.

Abstract The report presents the novel genetic algorithm (GA) developed to improve fuel cycle performance and in-core fuel management process. The primary purpose was to develop GA dedicated to solving core loading pattern (LP) problem with predefined core operation constraints. Particular focus was put on maximizing the length of the fuel cycle, but presented solutions allow to limit or control the magnitude of excess reactivity, population and type of fuel assemblies, fissile material mass and inventory of burnable absorbers. GA was implemented in a new computational framework coupled with PARCS3.2 core simulator. The Pressurized Water Reactor (PWR) based on the MIT BEAVRS Benchmark was used as the demonstration case, and developed tools were applied to improve the first fuel cycle. Several test simulations were performed, including population size, a number of generations, mutation level and other aspects of GA. New variance control method, was proposed and tested. It was compared with the standard roulette and rank method, showing improved convergence. Four test cases were studied with different constraints put on the fuel assemblies population, mass of fissile materials, burnable absorbers and limiting excess reactivity. The obtained results show that the new algorithm works properly and can allow designing efficient fuel loading pattern with increased fuel cycle length.

The report presents the simulations of the medium size 1000 MWth Sodium-Cooled Fast Reactor (SFR) core with oxide and metallic fuel. The core design is based on the OECD/NEA “Benchmark for Neutronic Analysis of Sodium- cooled Fast Reactor Cores with Various Fuel Types and Core Sizes”. The full-core 3D model and simulations were prepared with SERPENT 2.1.28 Monte Carlo neutron transport computer code. Static k-eigenvalue criticality and isotopic depletion calculations were performed with ENDF\B-VII.0 and JEFF 3.1.1 nuclear data libraries. The Beginning of Equilibrium Cycle (BOEC) state, the core depletion process and End of Equilibrium Cycle (EOEC) state were simulated. The obtained results confirm that the prepared models are appropriate, as the majority of the obtained results is in agreement with the Benchmark data single standard deviation.The report presents the simulations of the medium size 1000 MWth Sodium-Cooled Fast Reactor (SFR) core with oxide and metallic fuel. The core design is based on the OECD/NEA “Benchmark for Neutronic Analysis of Sodium- cooled Fast Reactor Cores with Various Fuel Types and Core Sizes”. The full-core 3D model and simulations were prepared with SERPENT 2.1.28 Monte Carlo neutron transport computer code. Static k-eigenvalue criticality and isotopic depletion calculations were performed with ENDF\B-VII.0 and JEFF 3.1.1 nuclear data libraries. The Beginning of Equilibrium Cycle (BOEC) state, the core depletion process and End of Equilibrium Cycle (EOEC) state were simulated. The obtained results confirm that the prepared models are appropriate, as the majority of the obtained results is in agreement with the Benchmark data single standard deviation.

The paper presents a study of the critical flow phenomena modeling with MELCOR 2.2.15254 severe accident computer code. The Marviken Critical Flow Test number 21 (CFT-21) experiment was selected as a representative critical flow-focused Separate Effect Test (SET). Various modeling aspects were investigated, including the nodalization, model setup, parameters, and sensitivity coefficients. A local-type sensitivity study was performed to analytically identify the significant parameters and assess their impact on the modeling. A dedicated regression-based approach, using standard deviation, was developed to find the best-fit MELCOR modeling parameters. The primary purpose of this work was to determine the appropriate approach to model critical flow with MELCOR 2.2, investigate the model performance, assess the influence of nodalization choices, identify significant sensitivity parameters, and prepare recommendations with an emphasis on best-estimate modeling. An additional outcome was the benchmark of the recent code version with the Marviken test.

Abstract Study of the Phebus FPT-1 experiment with MELCOR 2.2.11932 severe accident integral computer code is presented. The developed model and simulations cover the fuel bundle phase up to 20,000 s after the test initiation. The paper is focused on the assessment of material released from the bundle with emphasis on radioactive releases. The core degradation, thermal-hydraulics, hydrogen production and structural material releases are also studied as coupled phenomena. Simulations are compared with experimental results and selected solutions available in the literature. Parametric sensitivity calculations were performed for selected modelling parameters influencing releases. Especially, the CORSOR-Booth model revisions are compared and assessed. Radioactive releases for MELCOR 2.1 and 2.2 are compared. New modelling parameters for molybdenum and Ag-In-Cd released as pseudo fission products are presented. A satisfactory agreement was obtained, confirming the validity of the developed model and efficiency of the new MELCOR2.2 code for the FPT-1 test.

Artykuł przedstawia wyniki symulacji reaktora jądrowego EPR w przypadku cięzkiej awarii z całkowitą utratą zasilania zewnętrznego oraz całkowtym uszkodzeniem awaryjnych generatorów Diesla (Całkowita utrata zasilania prądem przemiennym). Obliczenia wykonano kodem komputerowym MELCOR 2.1 dla fazy awarii wewnątrz zbiornika reaktora. W rozpatrywanym scenariuszu całkowita utrata źródeł zasilania zewnętrznego oraz wewnątrznego prowadzi bezpośrednio do utraty chłodzenia, degradacji rdzenia, relokacji stopionego materiału do dolnej komory mieszania i ostatecznie do rozerwania zbiornika reaktora. Wyniki otrzymane kodem MELCOR porównano jakościowo oraz ilościowo z wynikami orzymanymi kodem MAAP4, wyniki są ze sobą zgodne w zadowalającym stopniu. In this paper the results of severe accident simulations for the EPR reactor in the case of loss of offsite power combined with total failure of all diesel generators (total loss of AC power) are presented. Calculations were performed with MELCOR 2.1 computer code for in-vessel phase of the accident. In this scenario, the unavailability of all offsite and onsite power sources and the lack of cooling leads directly to core degradation, material relocation to the lower plenum and rupture of the reactor pressure vessel. MELCOR results were compared qualitatively and quantitatively with MAAP4 code results and show a good agreement.