1. 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

2. 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)

3. 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

4. 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.

5. 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)

6. 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 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.

7. 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

8. 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 %

9. 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

10. 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

11. 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

12. 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 $ 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

13. 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 $

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

15. 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 mm for D, E & F rings)
Technical requirements for the operation of the reactor:
CE < (SHIM + TRANS + REG) / 2
TRANS + REG – CE > 0.50 $

16. Optimizing of Core Configuration Future Core Configuration
Quantity ρ($)
Core Excess
Shut Down
SHIM
TRANS
REG
Core Excess increased of $
Optimization of the use of the available fuel
Extention of reactor operation without new fuel

17. 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
LENA (Applied Nuclear Energy Laboratory) University of Pavia
Via Aselli 41, Pavia - Italy
tel.   
fax.  
Dr. Daniele Alloni
LENA (Applied Nuclear Energy Laboratory) University of Pavia
Via Aselli 41, Pavia - Italy
tel.   
fax.  

18. Thank you
for your attention!