The editors would like to extend thanks to B. W. Burnham and M. A. Morgan for their invaluable assistance in maintaining the MATPRO library. Mrs. Morgan's contributions in making computer plots and analyzing data are also gratefully acknowledged. We also extend thanks to S. B. Letson for his technical editorial expertise. Blank Page4.3 Fuel Thermal Expansion Subcode FTHEXP Listing 4.4 References VI X A-1.4. Specific heat capacity as a function of temperature and oxygen to metal ratio for UO2 A-1.5. Specific heat capacity as a function of temperature and oxygen to metal ratio for (UQ g,PuQ 2)02+x ^^ A-2.1. Comparison of measured and predicted values of the thermal conductivity of UO2 for materials corrected to 95% TD and standard deviation of data from theoretical curve A-2.2. Comparison of measured and predicted values of the thermal conductivity of (U,Pu)02 for materials corrected to 96% TD and standard deviation of data from the theoretical curve A-2.3. Calculated curves showing comparison between UO2 and (U,Pu)02 thermal conductivity A-2.4. The effect of varying the assumed value for the electronic contribution, Kp, on the calculated thermal conductivity of 95% TD UO2 with/kdt = 96 A-2.5. The effect of varying the assumed value for/QQ kdt on the calculated thermal conductivity of 95% TD UO2 with Kg held constant at 0.002 W/cm-K A-2.6. The standard deviation of the calculated UO2 thermal conductivity from the data base as a function of the assumed value of the conductivity integral A-3.1. Comparison of fitting polynomials with emissivity data of Claudson A-3.2. FEMISS representation of UO2 emissivity A-4.1. Comparison of UO2 thermal expansion data with those calculated from FTHEXP subcode A-4.2 Comparison of PUO2 thermal expansion data with those calculated from FTHEXP subcode A-5.1. Young's modulus for stoichiometric UO2 fuel at several temperatures and fractions of theoretical density A-5.2. Young's modulus for (U,Pu)02 with various oxygen to metal ratios XI A-5.3. Young's modulus data and least-squares hnear fit for stoichiometric UO2 fuel at room temperature and several different densities A-5.4. Ratio of Young's modulus for stoichiometric and nonstoichiometric fuels measured at room temperature compared to values predicted by de Novion's correlation A-5.5. Poisson's ratio as a function of temperature A-6.1.

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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...