SCDAP/RELAP5/MOD2 code manual

Abstract

ABSTRACl The SCDAP/REtAPS code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols In the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offslte power, loss of feedwater, and loss of flow. A generic modeling approach Is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are Included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented In Volumes I and II to acquaint the user with the modeling base and thus aid In effective use of the code. Volume III contains detailed instructions for code application and Input data preparation. In addition. Volume III contains user guidelines that have evolved over the past several years from application of the REtAP5 and SCDAP codes at the Idaho National Engineering taboratory, at other national laboratories, and by users throughout the world. The light water reactor (LWR) severe accident transient analysis code, SC0AP/RELAP5, has been developed at the Idaho National Engineering Laboratory (INEL) for the U. S. Nuclear Regulatory Commission (NRC) to provide an advanced best-estimate predictive capability for use In severe accident applications In support of the regulatory process. Code uses Include analysis required to support rulemaking, licensing audit calculations, evaluation of accident mitigation strategies, and experiment planning and analysis. Specific applications of this capability have Included analytical support for the loss-of-fluld test (LOFT), Power Burst Facility (PBF), ACRR, MISl, ROSA IV, and NRU experimental programs, as well as simulations of transients that lead to severe accidents, such as loss of coolant, anticipated transients without scram (ATWS), and operational transients In LWR systems. SCDAP/RELAP5 Is a highly generic code that. In addition to calculating the behavior of a reactor coolant system (RCS) during a severe accident transient, can be used for simulation of a wide variety of hydraulic and thermal transients In both nuclear and nonnuclear systems Involving steam-water noncondenslble solute fluid mixtures. SC0AP/RELAP5 was developed by Integrating three separate codes, RELAP5/M0D2, SCDAP, and TRAP-MELT. These codes were combined to model the coupled Interactions that occur between the core, the RCS, and the fission products during a severe accident. For example, blockage In the core, caused by fuel rod ballooning and meltdown, can have a significant effect on RCS flows. Fission products released from the core can have a significant effect on the RCS because of the heat produced during decay. These and many other coupled effects can have a sig...

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

“…Swelling as a result of both solid and gaseous fission products is included using empirical relations from MATPRO [18]. Solid fission product swelling is expressed as a simple linear function of burnup:…”

“…2 shows a plot of the gaseous and total fission product swelling as a function of temperature and burnup. The MATPRO [18] correlations indicate that gaseous swelling does not become significant until above 1500 K and is saturated by a burnup of 20 MWd/kgU. Fuel densification is computed using the ESCORE empirical model [4] given by:…”

“…The thermal conductivity of the gas mixture (k g ) is computed as a function of the individual gas conductivities and their respective mole fractions, using the gas mixture rule described in detail in MATPRO [18]. The temperature-dependent conductivity of each gas component is expressed as a simple power law (k = A T B ) and the respective mole fractions are computed based on the initial fill gas and all fission gas released from the fuel.…”

Advanced multiphysics coupling for LWR fuel performance analysis

“…The Young's modulus, Poisson's ratio, and thermal expansion coefficient for the fuel were assumed constant at 219 GPa, 0.345, and 10 Â 10 À6 1=K respectively, as given by Olander (Olander, 1976). The thermal conductivity, Young's modulus, Poisson's ratio, and thermal expansion coefficient of Zr-4 were assumed constant at 16 W/m/K, 75 GPa, 0.3, and 5:0 Â 10 À6 1=K respectively, using typical values from MATPRO (Allison et al, 1993). Mechanical contact between the fuel and clad was assumed to be frictionless.…”

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

“…In addition, material properties such as thermal expansion, Young's modulus, Poisson's ratio, yield stress and fracture stress are adopted from MATPRO package [42].…”