Fuel Performance Codes

The severe environment and the interaction with the neutron flux during the in-pile irradiation lead to a relevant variation of the nuclear fuel structure and characteristics. In order to properly estimate the status of the thermo-mechanical properties and dimensional variations of the fuel pins, several fuel performance codes have been developed and validated in the last decades. The processing of this information allows the user to initialize the data for irradiated fuel pins for further analysis in transient conditions with the SIMMER code.



TRANSURANUS is a fuel-performance code developed by JRC-Karlsruhe for thermal and mechanical analysis of fuel pins behavior during irradiation in nuclear reactors. Thermal and mechanical properties for different type of fuel, clad materials and coolants are available, with the possibility to simulate normal, off normal and accident conditions. The code counts several users across Europe and is used and developed by various research centers, universities, nuclear safety authorities and industrial partners. [1]



Developed by the Japan Atomic Energy Agency (JAEA), the Light Water Reactor (LWR) fuel analysis code FEMAXI-7 is capable to simulate high-burnup fuel behavior in normal and transient conditions. Mechanistic and empirical models allow the user to simulate thermal and mechanical behavior of fuel pins under different irradiation conditions in LWR reactors. [2]



Developed by the Nuclear Regulation Authority (NRA) as part of the ASTERIA-FBR code system for the analysis of Core Disruptive Accidents (CDAs) for Fast Breeder Reactors (FBR), the FEMAXI-FBR code is a fuel behavior calculation module which performs thermal and mechanical analysis of fast reactor nuclear fuels. Based on the light water reactor fuel behavior code FEMAXI-6, the FEMAXI-FBR code implements modified material properties and dedicated models for fast reactor, allowing the code to evaluate FBR fuel pin behavior from normal to accidental conditions. [3]


Application of the codes in:

  • V. Kriventsev, A. Rineiski, W. Pfrang, S. Perez-Martin, K. Mikityuk, G. Khvostov, Benchmark on behavior of MOX fuel pin under irradiation at nominal power in Sodium Fast Reactor, TopFuel Conference Proceeding, Zurich (2015);



Fig. 1. Initially full FBR pellet after irradiation: cross-section and fuel temperature profile at gap closure




1.  K. Lassmann, TRANSURANUS: a fuel rod analysis code ready for use, J. Nucl. Mater. 188 (1992) 295–302.

2.  M. Suzuki, H. Saitou, Y. Udagawa, F. Nagase, Light Water Reactor Fuel Analysis Code FEMAXI-7 ; Model and Structure, JAEA-Data/Code. (2013).

3.  T. Okawa, I. Tatewaki, T. Ishizu, H. Endo, Y. Tsuboi, H. Saitou, Fuel behavior analysis code FEMAXI-FBR development and validation for core disruptive accident,   Prog. Nucl. Energy. 82 (2015) 80–85.