Pulse Width Variation in a Rod Ejection Accident 1. Objectives

The work completed had the following objectives: i) To compute pulse width (full width at half maximum-FWHM) of the power curves for the Three Mile Island Unit 1 Pressurized Water Reactor (TMI-1 PWR) core model in the event of a super-prompt-critical rod ejection accident (REA) at Hot Zero Power (HZP) in the central fuel assembly (Rod 7A) at both EOC and BOC for various control rod worths and delayed neutron fractions. ii) To correlate the pulse width and maximum fuel pellet enthalpy change with control rod worth, and delayed neutron fractions. iii) To compare enthalpy and pulse width results with those obtained by Russian colleagues at the Russian Research Center – Kurchatov Institute (RRC-KI) in Moscow. iv) To make recommendations for future computational analyses of the REA in PWR's, and also for future fuel testing experiments. 2. Methodology 2.1 PARCS Reactor Dynamics Simulation Code The PARCS (Purdue Advanced Reactor Core Simulator) code (Version v1.05) was used to simulate both the steady-state and transient reactor dynamics behavior of the TMI-1 PWR core model. The PARCS code is a three-dimensional, two-group diffusion model using nodal methods [1]. PARCS can be coupled with a thermal-hydraulics code such as RELAP5 to get a complete self-consistent simulation of the reactor core, or a simplified thermal-hydraulics model that is incorporated in the PARCS code can be used in a stand-alone mode. Although the stand-alone version of PARCS has a limited range of applicability and accuracy, it can be run more quickly than the PARCS code coupled with RELAP5, and it is used to obtain the simulation results shown in this memorandum.