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CAP Code Version-up to 3.0 and Its Application to Pressure and Temperature Analysis
Abstract
This paper introduces the CAP version-up to 3.0 and the PT analysis method in order to apply to SMR containment.
In order to experimentally investigate the thermal hydraulic interaction between the reactor coolant system (RCS) and the containment, the integral effect tests were conducted in a test facility, ATLAS-CUBE. Three tests for an intermediate-break loss-of-coolant accident (IBLOCA) were conducted for simulating the break of a cold leg or a direct vessel injection (DVI) nozzle. The effect of the mass/energy (M/E) supply from the break and the asymmetric phenomena inside the containment according to the break simulation position were investigated with considering its impact on the pressure/temperature (P/T) behavior inside the containment. As the test results, both of the RCS and the containment were effectively cooled down during the transient, by activation of the safety injection system and the containment spray system, respectively. The M/E supply from the RCS increased a pressure and a steam-gas mixture temperature of the containment, while a wall condensation on the passive heat sink and a spray injection contributed to cool down the steam-gas mixture. The two-phase flow characteristics of the break flow such as the quality or the subcooling of the liquid phase could highly affect the P/T build-up of the steam-gas mixture in the containment. The ATLAS-CUBE test data in this study will contribute to evaluate the thermal hydraulic safety analysis codes for the RCS and the containment.
A condensation experiment in the presence of non-condensable gas in a vertical tube of the passive containment cooling system of the CP-1300 is performed. The experimental results show that the heat transfer coefficients (HTCs) increase as the inlet air mass fraction decreases and the inlet saturated steam temperature decreases. However, the dependence of the inlet mixture Reynolds number on the HTC is small for the operating range. An empirical correlation is developed, and its predictions are compared with experimental data to show good agreement with the standard deviation of 22.3%. The experimental HTCs are also compared with the predictions from the default and the alternative models used in RELAP5/MOD3.2. The experimental apparatus is modeled with two wall-film condensation models in RELAP5/MOD3.2 and the present model, and simulations are performed for several subtests to be compared with the experimental results. Overall, the simulation results show that the default model of RELAP5/MOD3.2 underpredicts the HTCs, and the alternative model over-predicts them, while the present model predicts them well throughout the condensing tube.
The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc.
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1992. Includes bibliographical references (leaves 124-128).
Two-Phase Pressure Drop in Straight Pipes and Channels; Water -Steam Mixtures at 600 to 1400 psia
- Janssen
Janssen et al., Two-Phase Pressure Drop in Straight Pipes
and Channels; Water -Steam Mixtures at 600 to 1400 psia,
GEAP 4616, General Electric Co. Atomic Power Equipment
Dept., San Jose, Calif. May 1964