Chapter 1 INTRODUCTION
The DIMNP of University of Pisa has been engaged in the assessment and application of system codes in the area of safety evaluation of the Light Water Reactor (LWR) in the last three decades [1]. Proposal for nodalization qualification criteria and for uncertainty method to evaluate the envisaged error of any code prediction [2] and tool to quantify the accuracy of calculation [3], constitute examples of the activities that have been completed.
The great advance in the computer’s power and in numerical methods during the last years, has allowed to couple system codes with 3D neutron kinetics codes, in order to obtain best estimate (BE) analysis. The international research community as well as the nuclear industry and the regulatory agencies have welcomed this new approach for calculation.
The DIMNP has readily followed this new trend in the safety analyses, and it has engaged several activities (e.g. international benchmark participation, Advanced and East-Europe reactor analyses) to acquire and qualify this new technology. From what has been written until now, it is clear the contest for the development of this thesis.
Furthermore, this thesis response to the request of the nuclear industry to improve the analysis tools for the high burnup fuel and also to the DIMNP request to acquire complete knowledge about the all processes involved in the coupled codes analyses, like cross section library generation.
So, the main thesis objectives can be resumed briefly as:
• acquire knowledge about the nuclear fuel behavior, and especially about the high burnup fuel behavior, during Reactivity Initiated Accident (RIA)
• acquire the capability to perform cross section library calculations
• acquire the capability to perform 3D neutron kinetics / thermal-hydraulic codes calculations
• analyze the safety of the WWER-1000 core during a Rod Ejection Accident (REA) • develop a methodology for the Energy released to the fuel calculation during REA Starting from a previously nodalization developed by the DIMNP for the participation to the NEA/OECD ‘WWER-1000 Coolant Transient Benchmark’ [4], it was acquired in Pisa the coupling codes methodology. Then, during a period of 6 months spent at the Pennsylvania State University (PSU) in the USA, it was acquired the cross section library calculation capability and it was developed the methodology for the Energy Release evaluation.
Three main parts compose this thesis.
In the first (Chapters 2 and 3) is reported an introduction to the fuel behavior during RIA, the state-of-the-art of the main international experiments performed and an analytical approach to this accident description. It is also reported a detailed description of the reference nuclear power plant (NPP) Kozloduy-6, and a description of the particular RIA studied, the REA.
In the second part (Chapters 4 and 5) is reported the modeling utilized and developed for the cross section library generation and for the transient calculation. It is also reported the description of the all the transient types modeled.
In the third part (Chapters 6 and 7) are reported the results of the calculations, their interpretation and finally the conclusions of the all work.
In Appendix A are instead reported those data that are not of immediate usefulness for the transient description, like the steady state calculations, but that can be show the correctness of the modeling adopted.
In Appendix B are reported, for some transient, the main parameters describing the behavior of the FA that were neat the FA that experienced the rod ejection during the transient.
In Appendix C is reported the description of the software modified and developed for the data processing, for the cross section interpolating process, the mapping process and the cross section formatting process.
In Appendix D is reported an analytical calculation for the Hot Zero Power / End Of Cycle REA transient in order to demonstrate the conservativeness of the zero dimensional (0-D) analysis compared to the result of the coupled codes analysis. A description of the software developed for the Steady State reactor code parameter evaluation is also reported.