September, 23Th -27Th San Diego - CA L.E.N.A. APPLIED NUCLEAR ENERGY LABORATORY, UNIVERSITY OF PAVIA - ITALY Evaluation of the TRIGA reactor fuel burn-up by means of Monte Carlo codes and measurements , D. Alloni1,2, A. Borio di Tigliole1,2, J. Bruni2, M. Cagnazzo1, R. Cremonesi2, G. Magrotti1 S. Manera1, F. Panza2, M. Prata1, A. Salvini1 1 Applied Nuclear Energy Laboratory - University of Pavia 2 Department of Physics - University of Pavia TRTR Annual Conference September, 23Th -27Th San Diego - CA

LENA Research Center facilities 250 kW TRIGA Mk II Research Reactor X ray industrial generator 250kV, 12mA dose 15,6 Gy/min 350kV, 6mA dose 17,5 Gy/min Gamma source of Co-60 (0.86KGy/h) Radiochemistry Laboratory Cyclotron IBA 18/9 18 MeV protons (Imax = 80 µA) 9 MeV deuterons (Imax = 40 µA)

TRIGA Mk II Research Reactor of L.E.N.A. Research thermal reactor Power: 250 kW Fuel: U-ZrH (enrichment: 20%) n° of fuel elements: 83 Moderator: demineralized water Reflector: graphite Control rods: 3 Various irradiation channels Thermal and thermalizing column

Abstract Knowledge of fuel composition as a function of irradiation time is of paramount importance both for reactor operation and fuel management. During reactor operation, the evaluation of each fuel element composition, allows to accomplish the constraints of the plant operation license while optimizing the core management. In this work we present a methodology of analysis based on Monte Carlo codes MCNP (vers. 4C) and MCB that was developed to evaluate the fuel burn-up and poisoning of the TRIGA Mark II nuclear research reactor of the Applied Nuclear Energy Laboratory (LENA) of the University of Pavia. This methodology, extendable to other TRIGA reactors, has been validated by comparing the results of the Monte Carlo simulation with the reactivity measurements performed at the reactor during several years of operation. The results of the fuel composition analysis were also used to design a new reactor core configuration in order to increase the core reactivity control margin using the available irradiated fuel elements.

Outline An methodology for the study of burn-up of nuclear fuel has been developed with simulations by means of Monte Carlo codes MCB, implemented in the code MCNP ver. 4C, and direct measurements in the TRIGA Mk II reactor The methodology has been validated by comparing the outputs of MCB code with the solutions of the system of differential equations (Bateman Equations) that describes the time evolution of the concentration of actinides and fission products The developed methodology was further validated by direct measurements to assess the production rates of long half-life fission products (FP) and of transuranic elements (TRU) in neutron fields with different energy spectrum and in different nuclear materials (natural uranium and thorium)

Outline The burn-up of fuel elements of TRIGA Mk II reactor of LENA has been calculated with the code MCB code until 31th December 2011. The reactivity of the simulated current core configuration has been compared with measures of core excess reactivity carried out on 2nd January 2012. A new core configuration has been studied with the objective of increasing the core excess reactivity with respect to the Technical Requirements. This allows to extend the life of the plant optimizing the use of the available fuel respecting the security parameters.

Core Excess Reactivity & Shutdown Margin Core in “cold conditions” Critical Reactor at zero power: 1.5 W Core Excess Reactivity = sum of the residual reactivity of the control rods (shown in red) Shutdown Margin= reactivity available to shut down the reactor (shown in blue) Fuel elements: green

Neutron Flux Evaluation in each Fuel Element position (1) Neutron flux evaluation by MCNP simulation in each grid position (both for aluminum elements and stainless steel elements) First Core Configuration (16th November 1965) Statistical error: σ < 0.1 %

Neutron Flux Evaluation in each Fuel Element position (2) For each fuel element, a weighted flux w has been calculated. This is equal to the average of the energy spectrum of the flux j weighted on the hours of operation in each location, where each fuel element was during his lifetime in the core up to 31th December 2011: weighted integral flux

MCB Simulation of Fuel Element I rradiation Inward cosine irradiation: neutrons generated on a sphere surface of radius 20 cm, directed inwardly the sphere, with an energy distribution in groups given by ϕj calculated for each energy interval The fuel is a sphere of radius 1.5 cm and its density was decreased by a 100 factor in order to minimize neutron self-absorption effects Irradiation time is variable depending on the element 108 neutrons have been generated for each simulation run FUEL

Determination of fuel composition Many nuclides are reported in MCB output Contribution to fission: 235U, 238U, 239Pu Saturable poisons: 149Sm (1.5 x 104 b) and 151Sm (4 x 104 b) Not-saturable poisons: virtual nuclide with a weight of about 50 barn for fission (reported in literature), has been identified in 45Sc (σ0a ≈ 27 b) with an initial weight of 2 atoms per fission Contribution of not-saturable poisons depends on many factors such as fuel enrichment, burn-up and energy distribution of the neutron flux Aim: evaluate the weight of not-saturable poisons for the TRIGA Mk II LENA reactor in order to simulate properly the poisoning effects

Simulations vs. Core Excess Reactivity Measurements MCNP simulations of 3 different core configurations For each configuration the simulated reactivity has been compared with the reactivity obtained by Core Excess measurement. From the first comparison relative to the 1965 core configuration an offset of 0.24 $ between simulation and measurement has been found. It is due to impurities of real materials and approximations in the core geometry This offset is present in all simulations and it has been subtracted from the value of the simulated reactivity 1st adjustment

2nd adjustment Final Result Simulations vs. Core Excess Reactivity Measurements 2nd adjustment From the second comparison relative to the 2009 core configuration a “fine-tuning” of the not-saturable poisons content was performed adjusting the amount of 45Sc to an optimal value of 2.4 atoms per fission of 235U (i.e. a thermal absorption cross section of about 65 barn per fission) instead than 2 atoms per fission Taking into account the previous adjustments, a third simulation of the current core configuration (31st December 2011) was performed and the result showed a good agreement with the measured core excess reactivity Final Result Configuration Sim. Reactivity Sim. Reactivity - offset Meas. Reactivity 16/11/1965 3.26±0.04 $ 3.02±0.04 $ 20/01/2009 2.70±0.04 $ 2.46±0.04 $ 2.45±0.04 $ 02/01/2012 2.47±0.04 $ 2.23±0.04 $ 2.26±0.04 $

LENA TRIGA Mk II Fuel Elemnts Burn-Up Current Core Configuration 235U fiss ≈ 372 g Energy ≈ 354.5 MWd Avg Burn-up ≈ 10.3 % Max Burn-up = 17.86 % 239Pu = 86 g 239Pu fission contributtion ≈ 1.7 % Quantity ρ($) Core Excess 2.26 Shut Down - 3.87 SHIM 3.09 TRANS 1.95 REG 1.09

Optimizing of Core Configuration Aim: knowing each fuel element burn-up, find a new core configuration just using available fuel elements, in order to obtain a greater Core Excess satisfying security parameters Rules on fuel elements movement: Instrumented elements must remain in the same position Elongation limits for fuel elements (5.08 mm for B & C rings - 6.35 mm for D, E & F rings) Technical requirements for the operation of the reactor: CE < (SHIM + TRANS + REG) / 2 TRANS + REG – CE > 0.50 $

Optimizing of Core Configuration Future Core Configuration Quantity ρ($) Core Excess 3.03 Shut Down - 3.64 SHIM 3.05 TRANS 2.39 REG 1.23 Core Excess increased of 0.77 $ Optimization of the use of the available fuel Extention of reactor operation without new fuel

Validation through actual element reshuffle to Coming up….. Validation through actual element reshuffle to measure effective CORE EXCESS For further information please contact: Dr. Michele Prata michele.prata@unipv.it LENA (Applied Nuclear Energy Laboratory) University of Pavia Via Aselli 41, 27100 Pavia - Italy tel.    +39 0382 987300 fax.   +39 0382 987302 www.unipv-lena.it Dr. Daniele Alloni daniele.alloni@unipv.it LENA (Applied Nuclear Energy Laboratory) University of Pavia Via Aselli 41, 27100 Pavia - Italy tel.    +39 0382 987300 fax.   +39 0382 987302 www.unipv-lena.it

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