[en] The computer program SAFE has been used to size and analyze the performance of a steam generator which has two types of heat transfer regions in Korean Standard Nuclear Power Plants (KSNP) and Korean Next Generation Reactor (KNGR) design. The SAFE code calculates the analytical boiling heat transfer area using the modified form of the saturated nucleate pool boiling correlation suggested by Rohsenow. The predicted heat transfer area in the boiling region is multiplied by a constant to obtain a final analytical heat transfer area. The inclusion of the multiplier in the analytical calculation has some disadvantage of loss of complete correlation by the governing heat transfer equation. Several comparative analyses have been performed quantitatively to evaluate the possibility of removing the multiplier in the analytical calculation in the SAFE code. The evaluation shows that the boiling correlation and multiplier used in predicting the boiling region heat transfer area can be replaced with other correlations predicting nearly the same heat transfer area. The removal of multiplier included in the analytical calculation will facilitate a direct use of a set of concerned analytical sizing values that can be exactly correlated by the governing heat transfer equation. In addition this will provide more reasonable basis for the steam generator thermal sizing calculation and enhance the code usability without loss of any validity of the current sizing procedure. (author)

[en] Most of nuclear power plants have a spent fuel pool (SFP) for the wet storage of the spent fuel assemblies (FAs). The methods used to provide cooling for the removal of decay heat from the stored spent FAs vary from plant to plant depending upon the plant specific designs. But the basic safety function remains the same in principle; that is, the spent FAs must be cooled and must remain covered with water during the storage conditions. NUREG-0800 SRP suggests/recommends guidelines for the acceptance criteria on the design of the spent fuel cooling system. According to the criteria recommended in NUREG-0800 SRP, the design specifications/requirements are prepared for individual plants. The following thermal-hydraulics requirements in terms of discharge scenarios and their cooling acceptance criteria are frequently used to demonstrate the adequacy of the cooling system design

[en] A valve opening and centrifugal pump speed control was investigated analytically in a simple pumping system where the differential pressure across the control valve is maintained constant over the required flow range. The valve control program was derived analytically only as a function of the required flow rate to maintain the constant differential pressure across the valve. The centrifugal pump speed control program was also derived analytically for the required flow rate for the constant differential pressure across the control valve. These derivations theoretically show that the independent control is possible between the valve and pump speed in a system with a constant valve pressure drop. In addition, it was shown that a linear pump speed control is impossible in maintaining the constant valve pressure drop