Comprehensive Calculations of IRR1 Fuel Burnup

INTRODUCTION
The Israeli Research Reactor 1 (IRR1) reached first criticality in 1960 and is in operation now for about 55 years. The current fuel stockpile of IRR1 consists of HEU MTR plate-type fuel elements (FEs), which are intermittently irradiated in the core for the past 40 years.
Considering the high burnup values of IRR1 fuel (up to 70%), reducing the uncertainty of burnup values should enable longer and safer utilization of IRR1 fuel. Hence, a joint experimental and calculation effort was launched by SNRC and Ben-Gurion University of the Negev, aiming at reducing the uncertainty associated with IRR1 fuel burnup data to below 5%.

METHODS
A series of detailed three-dimensional full core deterministic burnup calculations of IRR1 fuel stockpile is performed based on a comprehensive description of the core elements and the irradiation history of the core. The available data details every core configuration assembled and irradiated since 1985, including its total burnup. Unit cell calculations and few-group cross section generation are performed for each element in the reactor core using the Monte Carlo code Serpent. Three-dimensional full core burnup calculations are subsequently performed sequentially on all 164 different core configurations irradiated in IRR1 since 1985 using the multi-group nodal diffusion code DYN3D.
The few-group cross section calculations of core elements containing fissile material are straightforward. Other core elements, either FEs with strong absorbers or FEs without fissile material, need a different treatment for generating a correct neutron energy spectrum in the unit cell. In these cases, a mini-core structure is employed, with standard (fresh) FEs surrounding the element, providing source neutrons in the appropriate spectrum.

RESULTS
Out of total of 50 FEs, of which 45 are standard and 5 are special: all 5 special FEs are depleted between 61-67%, 19 standard FEs are depleted between 50-60%, and 10 standard FEs are depleted between 30-50%.

CONCLUSIONS
Three-dimensional full core burnup calculations were performed on all 164 different core configurations irradiated in IRR1 since 1985. The utilized codes include the continuous-energy Monte Carlo neutron transport code Serpent and the multi-group nodal diffusion code DYN3D.
The results shown here do not include the impurities in the aluminum, graphite, and other structural materials. The burnup calculations shown here are performed assuming power level of 5 MW. Similar calculations, carried out using the number of operation days with an adapted (lower) power level, gave similar results and validated this assumption. The burnup calculations were performed with the control blades inserted to height 75-80%.