David L. Aumiller

A generic semi-implicit coupling methodology for use in RELAP5-3D©

A generic semi-implicit coupling methodology has been developed and implemented in the RELAP5-3D{copyright} computer program. This methodology allows RELAP5-3D{copyright} to be used with other computer programs to perform integrated analyses of nuclear power reactor systems and related experimental facilities. The coupling methodology potentially allows different programs to be used to model different portions of the system. The programs are chosen based on their capability to model the phenomena that are important in the simulation in the various portions of the system being considered. The methodology was demonstrated using a test case in which the test geometry was divided into two parts each of which was solved as a RELAP5-3D{copyright} simulation. This test problem exercised all of the semi-implicit coupling features which were installed in RELAP5-3D0. The results of this verification test case show that the semi-implicit coupling methodology produces the same answer as the simulation of the test system as a single process.

An integrated relap5-3d and multiphase cfd code system utilizing a semi-implicit coupling technique

An integrated code system consisting of RELAP5-3D and a multiphase CFD program has been created through the use of a generic semi-implicit coupling algorithm. Unlike previous CFD coupling work, this coupling scheme is numerically stable provided the material Courant limit is not violated in RELAP5-3D or at the coupling locations. The basis for the coupling scheme and details regarding the unique features associated with the application of this technique to a four-field CFD program are presented. Finally, the results of a verification problem are presented. The coupled code system is shown to yield accurate and numerically stable results.

A PVM Executive Program for Use With RELAP5-3D©

A PVM executive program has been developed for use with the RELAP5-3D{sup C} computer program. The PVM executive allows RELAP5-3D{sup C} to be coupled with any number of other computer programs to perform integrated analyses of nuclear power reactor systems and related experimental facilities. The executive program manages all phases of a coupled computation. It starts up and configures a virtual machine, spawns all of the coupled processes, coordinates the time step size between the coupled codes, manages the production of printed and plotable output, and shuts the virtual machine down at the end of the computation. The executive program also monitors that status of the coupled computation, repeating time steps as needed and terminating a coupled computation gracefully if one of the coupled processes is terminated by the computational node on which it is executing. (authors)

Numerical Methods in Coupled Monte Carlo and Thermal-Hydraulic Calculations

Large-scale reactor calculations with Monte Carlo (MC), including nonlinear feedback effects, have become a reality in the course of the last decade. In particular, implementations of coupled MC and thermal-hydraulic (T-H) calculations have been separately developed by many different groups. Numerous MC codes have been coupled to a variety of T-H codes (system level, subchannel, and computational fluid dynamics). In this work we review the numerical methods that have been used to solve the coupled MC–T-H problem with a particular focus on the formulation of the nonlinear problem, convergence criteria, and relaxation schemes used to ensure stability of the iterative process. We use a simple pressurized water reactor pin cell problem to numerically investigate the stability of commonly used schemes and which problem parameters influence the stability—or lack thereof. We also examine the role that the running strategy used in the MC calculation plays in the convergence of the coupled calculation. Results indicate that the instability in fixed-point iterations is driven by the Doppler feedback effect and that underrelaxation can be used to restore stability. We also observed that a form of underrelaxation could be achieved by performing the coupled iterations without converging the MC fission source each iteration. By performing many iterations of few histories, we observed rapid convergence to the coupled MC–T-H solution in a relatively small number of batches. Numerical results also showed that the presence of instability in the fixed-point iteration is independent of the stochastic noise in the MC simulation.

Critical Heat Flux During Reflood Transients in Small-Hydraulic-Diameter Geometries

Abstract Critical heat flux (CHF) at a low-flow condition in a small-hydraulic-diameter duct is an important phenomenon for a Materials Test Reactor/Advanced Test Reactor (MTR/ATR) design under a number of accident conditions, including reflood transients. Current CHF models in the literature, such as the Mishima/Nishihara and Oh/Englert CHF models, are based on macroscopic system parameters and not local thermal-hydraulic conditions. These macroscopic parameter–based models cannot be readily used for analysis in transient best-estimate thermal-hydraulic codes. The present work focuses on developing a low-flow-rate CHF correlation, based on local conditions, that is amenable to implementation into a best-estimate transient thermal-hydraulic code for a small-hydraulic-diameter duct. The model development proceeds with a means of correlating CHF data to local conditions parameters and then applying a correction factor to the resulting correlation, subsequently permitting accurate predictions over a range of pressures. An evaluation of the proposed local conditions-based CHF model is conducted by predicting independent sets of CHF experimental results over a range of flow rate, pressure, and subcooling conditions. Conclusions on the viability of the proposed CHF model and suggestions for future efforts in improving the reflood heat transfer CHF models for small-hydraulic-diameter ducts are provided with an evaluation of the model results.

Modeling moving systems with RELAP5-3D

Abstract The RELAP5-3D code is typically used to model stationary, land-based, thermal-hydraulic systems and contains specialized physics for the modeling of nuclear power plants. It can also model thermal-hydraulic systems in other inertial and accelerating frames of reference. By changing the magnitude of the gravitational vector through user input, RELAP5-3D can model thermal-hydraulic systems on planets, moons, and space stations. Additionally, the field equations were modified to model thermal-hydraulic systems in a noninertial frame, such as occur onboard moving craft or during earthquakes for land-based systems. Transient body forces affect fluid flow in thermal-fluid machinery aboard accelerating crafts during rotational and translational accelerations. It is useful to express the equations of fluid motion in the accelerating frame of reference attached to the moving craft. However, careful treatment of the rotational and translational kinematics is required to accurately capture the physics of fluid motion. Correlations for flow at angles between horizontal and vertical are generated via interpolation because limited experimental data exist. Equations for three-dimensional fluid motion in a noninertial frame of reference are developed. Two different systems for describing rotational motion are presented, user input is discussed, and examples of a modeled simple thermal-hydraulic system undergoing both rotational and translational motion are provided.

Extremely accurate sequential verification of RELAP5-3D

Abstract Large computer programs like RELAP5-3D solve complex systems of governing, closure, and special process equations to model the underlying physics of thermal-hydraulic systems and include specialized physics for the modeling of nuclear power plants. Further, these programs incorporate other mechanisms for selecting optional code physics, input, output, data management, user interaction, and post-processing. Before being released to users, software quality assurance requires verification and validation. RELAP5-3D verification and validation are focused toward nuclear power plant applications. Verification ensures that the program is built right by checking that it meets its design specifications, comparing coding algorithms to equations, comparing calculations against analytical solutions, and the method of manufactured solutions. Sequential verification performs these comparisons initially, but thereafter only compares code calculations between consecutive code versions to demonstrate that no unintended changes have been introduced. An automated, highly accurate sequential verification method, based on previous work by Aumiller, has been developed for RELAP5-3D. It provides the ability to test that no unintended consequences result from code development. Moreover, it provides the means to test the following code capabilities: repeated time-step advancement, runs continued from a restart file, and performance of coupled analyses using the R5EXEC executive program. Analyses of the adequacy of the checks used in these comparisons are provided.

Development of an Integrated Code System Using R5EXEC and RELAP5-3D

Abstract The RELAP5-3D system analysis code is a generic and flexible analysis program that is used for the analysis of steady-state and transient scenarios of nuclear reactors. While RELAP5-3D has been successfully applied to a very large range of problems, other analysis tools exist that provide either additional accuracy or capability. This paper describes an extensible integrated code system that utilizes the RELAP5-3D analysis code in conjunction with the R5EXEC executive. A description of the different types of time-step algorithms and data transfer methods is provided. Discussions of the various strengths and weaknesses of the coupled code system are provided. Finally, examples of how the coupled code system has been exercised are included.

AUTOMATED, HIGHLY ACCURATE VERIFICATION OF RELAP5-3D

Computer programs that analyze light water reactor safety solve complex systems of governing, closure and special process equations to model the underlying physics. In addition, these programs incorporate many other features and are quite large. RELAP5-3D[1] has over 300,000 lines of coding for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. Verification ensures that a program is built right by checking that it meets its design specifications. Recently, there has been an increased importance on the development of automated verification processes that compare coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions[2]. For the first time, the ability exists to ensure that the data transfer operations associated with timestep advancement/repeating and writing/reading a solution to a file have no unintended consequences. To ensure that the code performs as intended over its extensive list of applications, an automated and highly accurate verification method has been modified and applied to RELAP5-3D. Furthermore, mathematical analysis of the adequacy of the checks used in the comparisons is provided.

A Mechanistic Model for Droplet Deposition Heat Transfer in Dispersed Flow Film Boiling

Abstract A mechanistic droplet deposition model has been developed to quantify the direct-contact heat transfer present in dispersed flow film boiling. Lagrangian subscale trajectory calculations utilizing realistic velocity and temperature distributions in the momentum boundary layer are used to determine the number of dispersed droplets able to achieve contact with the heated wall. Coupling the droplet deposition model with a physical direct-contact heat transfer coefficient model allows the total direct-contact heat transfer to be determined based upon the local vapor mass flux, wall superheat, and vapor superheat. Comparisons to the existing models highlight the more mechanistic nature of the proposed model.