The Advanced Multilevel Predictor-Corrector Quasi-Static Method for Pin-Resolved Neutron Kinetics Simulation

The Advanced Multilevel Predictor-Corrector Quasi-static Method (AML-PCQM) is proposed in this work. The four computational levels, including transport, Multi-Group (MG) Coarse Mesh Finite Difference (CMFD), One-Group (1G) CMFD, and Exact Point-Kinetics Equation (EPKE), are coupled with a new dynamic iteration strategy. In each coupling algorithm, the original Transient Fixed Source Problem (TFSP) is solved in the predictor process using coarse time step, and then the flux distribution is factorized to the functions of amplitude and shape in the next corrector process. Finally, multiple fine time steps are used to adjust the predicted solution. Two heterogeneous single assembly problems with the prompt control rod withdrawal event are used to verify the AML-PCQM scheme’s accuracy and efficiency. The numerical results obtained by different cases are compared and analyzed. The final results indicate that the AML-PCQM performs the remarkable advantages of efficiency and accuracy with the reference cases.

SPX Benchmark Part II: Transient Results

The paper presents a transient simulation phase of the new benchmark on a large sodium fast reactor (SFR). This phase of the benchmark is devoted to the modelling of selected operational transients performed during start-up tests of the French SFR Superphénix. Six operational transients were selected for the analysis. The specifications of a simplified thermal hydraulic model equipped with point kinetics reactivity data and boundary conditions for the selected transients are given in the paper. The developed model contains necessary thermal hydraulic description of the primary system components, assumptions to account for thermal expansion reactivity feedbacks from out-of-core structures, neutron kinetics parameters, power distribution, and reactivity coefficients. The neutronic input parameters were obtained with the help of the Monte Carlo code Serpent during the first phase of the benchmark related to static neutronic characterization of the core. In this study, the solution of the transient benchmark was obtained with three thermal hydraulic system codes, namely TRACE, SIM-SFR, and ATHLET. The numerical results, compared to the available experimental data, exhibit a reasonable mutual agreement. Particular discrepancies between calculations and experiments could not be fully resolved. Therefore, a set of recommendations for achieving an improved agreement was proposed. In general, the proposed transient benchmark can be seen as an effective tool for validation and cross comparisons of system codes applied for safety analyses of SFRs, including approbation and comparison of different modelling features for thermal expansion of the out-of-core structures.

Modeling of Reactivity Effects and Transient Behaviour of Large Sodium Fast Reactor

In the paper the reactivity characteristics of the core of the large sodium fast reactor Superphenix (SPX) were evaluated and compared with available experimental data. The analysis was performed using the TRACE system code modified for the fast reactor applications. The simplified core model was developed aiming to overcome the lack of detailed information on design and realistic core conditions. Point Kinetics neutronic model with all relevant reactivity feedbacks was used to calculate transient power. The paper focuses on challenging issue of modelling of the transient thermal responses of primary system structural elements resulting in reactivity feedbacks specific to such large fast reactor which cannot be neglected. For these effects, the model was equipped with dedicated heat structures to reproduce important feedbacks due to vessel wall, diagrid, strongback, control rod drive lines thermal expansion. Peculiarly, application of the model was considered for a whole range of core conditions from zero power to 100% nominal. The developed core model allowed reproducing satisfactorily the core reactivity balance between zero power at 180?C and full power conditions. Additionally, the reactivity coefficients k, g, h at three power levels were calculated and satisfactory agreement with experimental measurements was also observed. The study demonstrated feasibility of application of relatively simple model with adjusted parameters for analysis of different conditions of very complex system.

THE APPLICABILITY RANGE OF THE MODIFIED SOURCE MULTIPLICATION (MSM) METHOD TESTED IN THE FAST VENUS-F REACTOR

The MSM method is an experimental technique for determination of reactivity of a sub-critical reactor. It consists of one dynamic measurement followed by two static measurements, which use an extraneous neutron source. For the data analysis, the core averaged kinetic parameters need to be calculated as well as a spatially-dependent correction factor that corrects for the point kinetics approximation. In order to test the range for which the method is valid and to demonstrate the reliability of the correction factor calculations in a fast reactor, a dedicated experimental campaign was performed in the fast lead-bismuth VENUS-F reactor. The reactivity of a dozen of sub-critical configurations was measured with the MSM method using ten 235U fission chambers. The detectors were located at various distances from the active zone and from the extraneous neutron source, leading to a large range of values of a correction factor (calculated with the Monte Carlo MCNP5 code) used in the data analysis.

PREDICTOR-CORRECTOR QUASI-STATIC METHOD FOR TIGHTLY-COUPLED REACTOR MULTIPHYSICS TRANSIENT CALCULATIONS

Several methods have recently been developed to solve multiphysics transients using the Improved Quasi-Static method and its derivatives. In order to address some perceived drawbacks of these methods, we have developed a new method for solving multiphysics transient calculations using the Predictor-Corrector Quasi-Static method. Our method involves computing reactivity feedback parameters during each transport timestep in order to enable reactivity feedback on the small timescale used by the point kinetics phase of the calculation. The advantage of this approach is that the transport solver does not need to store local flux information between time steps, potentially making it more appropriate for use with a Monte Carlo solver. We have demonstrated the accuracy of our method by solving a simple model problem that exhibits difficult multiphysics behavior. Additionally, we have compared our results against another recently published multiphysics coupling scheme, confirming that our approach does not negatively affect the accuracy of the transient solution.