Davide Rozzia

OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout) occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes) can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters). One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT) conducted by the Nuclear Power Engineering Corporation (NUPEC) in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

Modeling Large Break-LOCA: In Reactor Fuel Bundle Materials Test MT-4 and MT-6A

When creating power from nuclear fission, the fuel matrix and its cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well.Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. In this connection, OECD/NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation (IFPE)”. This database includes the data set of the projects MT-4 and MT-6A analyzed in the current paper.The MT-4 test bundle simulated a 6×6 section of a 17×17 3% enriched, full-length non-irradiated PWR fuel assembly. There were 20 non-pressurized guard fuel rods to isolate the 12 central, pressurized tests rods; the four corner rods were deleted. In the MT-6A test, the 20 guard rods used in the previous tests were replaced with 9 pressurized thus, a total of 21 test rods were in MT-6A. Only limited destructive post irradiation examination was performed on these two tests.The objective of the activity is the validation of TRANSURANUS “v1m1j09” code in predicting fuel and cladding behavior under LOCA conditions using the experimental databases MT-4 and MT-6A. It is pursued assessing the capabilities of the code models in simulating the phenomena and parameters involved, such as: pressure trend in the fuel rod, cladding creep, α to β-phase phase transformation, oxidation, geometry changes and finally failure prediction. The analysis is aimed at having a comprehensive understanding of the applicability and limitations of the code in the conditions of the experiments. Finally, probabilistic calculations are performed to complete the analysis.The objective of the activity is fulfilled addressing the behavior of two equivalent full lengths fuel rods, one for each test., suitable for the assessment of TU code versions “v1m1j09”.Copyright

Preliminary Discussion on LFR Fuel Pin Design: Current Status, Fuel Modeling and Open Issues

The present paper deals with the investigation of the LFR fuel rod conceptual design proposed in the European Lead cooled System project (ELSY) and in the Advanced Lead Fast Reactor European Demonstrator (ALFRED). Two main objectives are pursued. The first one is to provide a general overview of the status of the LFR fuel rod conceptual design. The main focus is to point out comments related to the pellet pre-design and the methodologies adopted for the preliminary thermo-mechanical assessment of the single fuel rod by means of fuel pin mechanic codes.The second objective is connected to the analysis of the cladding component. In the long term, LFR fuel rod cladding is expected to be based on Ferritic-Martensitic steels. Since irradiation data on these class of materials in high temperature fluent lead are not available, sodium technology cladding materials are considered and briefly discussed as candidate materials for LFR application in the near to medium term.Copyright