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MEKIN: MIT-EPRI nuclear reactor core kinetics code. Final report. [BWR and PWR]
R.W. Bowring, J.W. Stewart, R.A. Shober, R.N. Sims
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Abstract
The computer program MEKIN provides a solution in three dimensions for the time-dependent, two-group, neutron diffusion equations and corresponding thermal-hydraulic equations that model the transient behavior of a light water moderated power reactor. The code accepts initial steady state power level, inlet coolant temperature and flow rate. It first computes the initial steady state flux and power distributions throughout the core and then the response to a user-specified transient (change in inlet flow, temperature or control rod pattern). For PWR's the code represents transient cross flow between thermal-hydraulic channels. This document, in three parts, describes the code capabilities and overall calculational strategies, the instructions for setting up a problem, and the detailed structure of the program for those who may wish to make changes in it. (GRA)