The present thesis deals with a 3D analysis of core behaviour during a fast control rod ejection accident in a Pressurized Water Reactor.

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ABSTRACT

The present thesis deals with a 3D analysis of core behaviour during a fast control rod ejection accident in a Pressurized Water Reactor.

The Nuclear Power Plant that has been chosen in order to simulate this accident was the Three Mile Island 1 Unit, a PWR with nominal power of 2772 MW

(850 MW

), built by Babcock &

Wilcox, still in operation in Pennsylvania (USA).

The rod ejection accident (REA) is a design basis accident for a PWR and is characterised by a rapid reactivity insertion together with a power burst and an adverse core power distribution. If this event were to happen, a fuel rod thermal transient which could cause DNB may occur together with limited fuel damage.

In order to study such a particular accident, the coupled 3D Neutron Kinetic and Thermal- Hydraulic (3D-NK-TH) codes technique was implemented to simulate a multi-dimensional core behaviour. Coupled RELAP5/3.3-PARCS and RELAP5/3D-NESTLE codes were used in the development of this work.

The 3D method for TH and NK calculations of a reactivity-initiated accident is at present time worldwide spread and tends to replace the former 2D or 1D evaluations. The motivation for its extensive use is the possibility of gaining margins by predicting a more realistic core behaviour during the accident than the former calculation tools. Recent computer developments, resulting in the availability of powerful computation capabilities at reasonable costs, made it possible to perform detailed dynamic TH system analysis together with coupled 3D-NK core simulation even on a standard commercial PC system.

The main objectives of the present work were the following:

• Study of the coupled codes methodology, identification of its potentialities and fields of application;

• Appraisal of computational tools and acquisition of capability to perform 3D-NK-TH coupling studies;

• Comparison between predictions of different codes in best-estimate transient simulations;

• Verification of the capability of these system codes to analyse complex transients with coupled core-plant interactions; and

• Evaluation of safety margins for the reference plant.

The study confirms the consistency between the methods used for 3D transient calculations

and their strong capability in predicting core response during a REA in a PWR.