George L. Mesina

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.

Reformulation RELAP5-3D in FORTRAN 95 and Results

RELAP5-3D is a nuclear power plant code used worldwide for safety analysis, design, and operator training. In keeping with ongoing developments in the computing industry, we have re-architected the code in the FORTRAN 95 language [2], the current, fully-available, ANSI standard FORTRAN language. These changes include a complete reworking of the database and conversion of the source code to take advantage of new constructs. The improvements and impacts to the code are manifold. It is a completely machine-independent code that produces machine independent fluid property and plot files and expands to the exact size needed to accommodate the user’s input. Runtime is generally better for larger input models, many prior user-reported problems have been resolved, and the program is better tested. Other impacts of code reformulation are improved code readability, reduced maintenance and development time, increased adaptability to new computing platforms, and increased code longevity. Comparison between the pre- and post-conversion code are made on the basis of programming metrics and code performance.Copyright

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 Fuel Cycle and Core Design Analysis Method for New Cladding Acceptance Criteria Using PHISICS, RAVEN and RELAP5-3D

The Nuclear Regulatory Commission (NRC) has considered revising the 10 CFR 50.46C rule [1] for analyzing reactor accident scenarios to take the effects of burn-up rate into account. Both maximum temperature and oxidation of the cladding must be cast as functions of fuel exposure in order to find limiting conditions, making safety margins dynamic limits that evolve with the operation and reloading of the reactor. In order to perform such new analysis in a reasonable computational time with good accuracy, INL (Idaho National Laboratory) has developed new multi-physics tools by combining existing codes and adding new capabilities. The PHISICS (Parallel Highly Innovative Simulation INL Code System) toolkit [2,3] for neutronic and reactor physics is coupled with RELAP5-3D [4] (Reactor Excursion and Leak Analysis Program) for the LOCA (Loss of Coolant Accident) analysis and RAVEN [5] for the PRA (Probabilistic Risk Assessment) and margin characterization analysis. In order to perform this analysis, the sequence of RELAP53D input models had to get executed in a sequence of multiple input decks, each of them had to restart and slightly modify the previous model (in this case, on the neutronic side only) This new RELAP5-3D multi-deck processing capability has application to parameter studies and uncertainty quantification. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical PWR (Pressurized Water Reactor). INTRODUCTION The nuclear power industry is continually improving its designs, safety equipment, processes, and analysis methods. The NRC is considering a revision of the requirements in 10 CFR 50.46C rule, focused on the operation of the ECCS (Emergency Core Coolant System) in LOCA scenarios [1]. Novel analysis strategies will be required to account for the effects of fuel burn-up rate. It is necessary to cast the maximum temperature and oxidation of the cladding as functions of the fuel exposure in order to find the limiting conditions of the reactor, with its different design and different reloading patterns. This revision requires the development of new tools and capabilities to calculate the dynamic phenomena of the multiphysics system to the required accuracy in a reasonable amount of time. To perform such analysis, a rigorous Probabilistic Risk Assessment (PRA) strategy must be employed. The PHISICS code toolkit [2,3] is being developed at INL to provide state of the art analysis tools to nuclear engineers. It implements many choices of algorithms and meshing schemes for optimizing accuracy needs on available computational resources. Analysis tools currently available in the PHISICS package are a nodal and semi-structured transport core solver, INSTANT, a depletion module, MRTAU, a time-dependent solver, TimeIntegrator, a cross section interpolation and manipulation framework, MIXER, a criticality search module CRITICALITY, and a fuel management and shuffling tool SHUFFLE. The tools are developed as independent modules in a pluggable fashion in order to simplify maintenance and development. PHISICS can be run in parallel to takes advantage of multiple computer cores (workstations and highperformance computing systems). The package is directly coupled with the system safety analysis code RELAP5-3D [4] through a Fortran 95 interfacing module that contains communication subroutines that translate

Nuclear Hybrid Energy System Modeling: RELAP5 Dynamic Coupling Capabilities

The nuclear hybrid energy systems (NHES) research team is currently developing a dynamic simulation of an integrated hybrid energy system. A detailed simulation of proposed NHES architectures will allow initial computational demonstration of a tightly coupled NHES to identify key reactor subsystem requirements, identify candidate reactor technologies for a hybrid system, and identify key challenges to operation of the coupled system. This work will provide a baseline for later coupling of design-specific reactor models through industry collaboration. The modeling capability addressed in this report focuses on the reactor subsystem simulation.