Mathieu Hursin

Analysis of the Core Power Response During a PWR Rod Ejection Transient Using the PARCS Nodal Code and the DeCART MOC Code

Abstract The current state of the art in the analysis of a control rod ejection event in a pressurized water reactor (PWR) relies on homogenization methods in which the assembly-averaged power from a whole-core nodal neutronics simulator is used with some type of flux reconstruction to estimate the individual fuel rod power. Recently, there has been interest in taking advantage of methods that do not require homogenization, such as the DeCART code, to perform time-dependent neutron transport calculations. These calculations could provide not only more accurate pin power results but also intrapin power information during the transient. The work described in this paper is the analysis of a PWR control rod ejection transient using the nodal core simulator PARCS, which employs homogenization methods, and the method of characteristics (MOC) code DeCART, which treats the explicit geometry. Higher-fidelity methods such as those used by DeCART have the potential to quantify the homogenization and modeling errors inherent in the lower-order methods. The methods used in PARCS and DeCART are briefly described as well as the approach to generate the temperature feedback for the rod ejection event. The results are compared and discussed. For the considered transient scenario, PARCS and DeCART are in generally good agreement for the predicted global and local powers as well as for the temperature.

The Development and Implementation of a One-Dimensional Sn Method in the 2D-1D Integral Transport Solution

Abstract During the last several years, a class of algorithms has been developed based on two-dimensional–one-dimensional (2D-1D) decomposition of the reactor transport problem. The current 2D-1D algorithm implemented in the DeCART (Deterministic Core Analysis based on Ray Tracing) code solves a set of coupled 2D planar transport and 1D axial diffusion equations. This method has been successfully applied to several light water reactor analysis problems. However, applications with strong axial heterogeneities have exposed the limitations of the current diffusion solvers used for the axial solution. The work reported in this paper is the implementation of a discrete ordinates (Sn)-based axial solver in DeCART. An Sn solver is chosen to preserve the consistency of the angular discretization between the radial method of characteristics and axial solvers. This paper presents the derivation of the nodal expansion method (NEM)-Sn equations and its implementation in DeCART. The subplane spatial refinement method is introduced to reduce the computational cost and improve the accuracy of the calculations. The NEM-Sn axial solver is tested using the C5G7 benchmark. The DeCART results with the axial diffusion solver shows keff errors of approximately −95, −74, and −110 pcm for the unrodded configuration, rodded configuration A, and rodded configuration B, respectively. These errors decrease to approximately −40, −11, and −12 pcm by using the NEM-Sn solver. In terms of pin power distribution, the use of the NEM-Sn solver has a small effect, except for the heavily rodded configuration. The implementation of the subplane scheme makes it possible to maintain a coarse axial mesh and therefore to reduce the computational cost of the three-dimensional calculations without reducing the accuracy of the solution.

FUTURE EXPERIMENTAL PROGRAMMES IN THE CROCUS REACTOR

CROCUS is a teaching and research zero-power reactor operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology (EPFL). Three new experimental programmes are scheduled for the forthcoming years. The first programme consists in an experimental investigation of mechanical noise induced by fuel rods vibrations. An in-core device has been designed for allowing the displacement of up to 18 uranium metal fuel rods in the core periphery. The vibration amplitude will be 6 mm in the radial direction (±3 mm around the central position), while the frequency can be tuned between 0.1 and 5 Hz. The experiments will be used to validate computational dynamic tools currently under development, which are based on DORT-TD and CASMO/S3K code systems. The second programme concerns the measurement of in-core neutron noise for axial void profile reconstruction. Simulations performed at Chalmers University have shown how the void fraction and velocity profiles can be reconstructed from noise measurements. The motivation of these experiments is to develop an experimental setup to validate in-core the method in partnership with Chalmers University. The third experimental programme aims at continuing the validation effort on the nuclear data required in the calculation of GEN-III PWR reactors with heavy steel reflectors. This is a collaboration with CEA Cadarache that extends the results of the PERLE experiments carried out in the EOLE reactor at CEA. Scattering cross sections at around 1 MeV will be studied separately by replacing successively the water reflector by sheets of stainless steel alloy and pure metals – iron, nickel, and chromium. Data will be extracted from the measured flux attenuation using foils in the metal reflector and from the criticality effects of these reflectors. In parallel to the three reactor experiments, we develop in-core detectors and measurement systems. Following the last development of a neutron noise measurement station in pulse mode, a second neutron noise station in current mode is being designed. In current mode the reactor can be used at higher power without dead time effects. It allows faster measurement time or lower results uncertainties. Finally, a joint development of a full new detection system based on chemical vapour deposited (sCVD) diamond has been started with the CIVIDEC instrumentation start-up company. A first prototype has been tested in November 2015 in CROCUS. One of the main purposes is to work on the discrimination of gammas, thermal and fast neutrons for demonstrating the interest of this detector type in a mixed neutron-gamma field.

Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.