Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor
© Guerra et al.; licensee Springer. 2013
Received: 16 May 2013
Accepted: 29 October 2013
Published: 9 November 2013
This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed.
The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. The evaluation of the thermal and epithermal neutron flux in the sample position was done considering the experimental data obtained from a vertical neutron guide, already existent in the reactor, and the simulated model for the facility.
The experimental determination of the neutron flux was obtained through the standard procedure of using Au monitors in different positions of the vertical tube. In order to validate both, this experiment and calculations of the simulated model, the flux was also determined in different positions in the core used for sample irradiation. The model of the system was developed using the Monte Carlo code MCNP5.
The preliminary results suggest the possibility of obtaining a beam with minimum thermal flux of magnitude 106 cm-2 s-1, which confirm the technical feasibility of the installation of PGAA at the TRIGA IPR-R1 reactor. This beam would open new possibilities for enhancing the applications using the reactor.
In 2001, the CDTN started a program for the optimisation of the TRIGA´s utilisation. Since then, several new projects were initiated like the implementation of the k0-standardization method; the investigation of the mechanisms involved for the improvement of the Brazilian gemstones and production of new radioisotopes for radiopharmaceutical studies, Leal et al. (2006a, 2006b; 2007, Leal et al. (2013), Gonçalves et al. (2011), Soares et al. (2012). For allowing the enhancement of the reactor´s utilisation, a study of raising the power of the reactor from 100 to 250 kW was already concluded and it is now in the final process of licensing.
This work describes the suggested set-up for the PGAA facility, the computational model of the system and the reactor core and the experimental data of the neutron flux.
Availability and requirements
Preliminary design of the PGAA facility
Guide tube. An aluminium tube with a diameter of 5 cm guides the neutron from the top of the reflector to the sample chamber. The beam will be shut flooding the tube with water. The upper part of the tube will be covered with lead shielding.
Sample chamber. The beam is guided into the sample chamber from below. The chamber is made of at least 5-cm-thick lead. It has to be lined with a shielding material containing Li (with some percentage of Li-6).
Detector. A 20–30% relative efficiency, n-type HPGe detector will be purchased. It will be completely covered with lead shielding leaving with just a small hole in the direction of the sample (Collimated detector). The sample chamber and the shielding of the detector is practically one box that has a collimator (a wall with a hole) in the middle.
Beam-stop. The beam-stop is above the sample line. It will have an inner layer containing boron and hydrogen.
The Monte Carlo programs MCNP5 and MCNPX were used for modelling the design of the system and calculations of the neutronic parameters to evaluate the feasibility of installing the PGNAA at the TRIGA reactor, Dalle (2005). The MCNP computer code was used in the calculation of the reaction rates and fluxes. MCNP is a general-purpose, continuous-energy, generalized-geometry Monte Carlo transport code. The calculations reported in this paper were performed with version 5.1.40 of the code and with the ENDF/B-VII.0 cross-section library (processed at the National Nuclear Data Centre at Brookhaven, obtained from the Radiation Safety Information Computational Centre at Oak Ridge), Briesmeiter (1997).
Neutron flux in the vertical tube
Three different set of experimental data (1), (2) and (3), considered in this work. In the experiment (1), performed in the 80’s just after the installation of the tube, a metallic 115In foil was used as a flux monitor and the cadmium ratio, used for the determination of the epithermal flux, Parry (2003). In that time, the configuration of the reactor core was different from the present one, Figure 7, due to the replacement of some fuel elements.
In the experiments (2) and (3), bare and cadmium-covered samples of gold (Au 100%), disks of 125 mg, 1.25 cm in diameter, and 0.06 mm height were used as flux monitors. The irradiation time was 4 h and 8 h, respectively, for bare and cadmium covered samples. All the experiments were performed with the reactor operating at 100 kW. After the irradiation, the gamma spectra of each sample was obtained using a system equipped with an HPGe detector GC 5019 and Genie 2000, v2.0 Spectroscopy software, provided by Canberra Industries, Inc. The counting time was adjusted to provide a net peak area of at least 10,000 counts for the 411.8 keV peak from the 198Au, Parry (2003). The thermal and epithermal flux were determined from the gamma spectra using the formalism of Høgdahl convention, De Corte (1987).
Flux in the carrousel
For the determination of the neutron flux in the carrousel, Al-Au(0.1%) foils (disks of 5 mm diameter and 0.2 mm thick) were irradiated in 5 different positions, as indicated in Figure 7. Samples were simultaneously irradiated for 2 h and the activity was measured using high-purity germanium detector (HPGe).
The activation reaction considered in this experiment is: 197Au (n,γ) 198Au, where 411.8 keV is the energy for the most prominent gamma 198Au photopeak. The cross section for the reaction 197Au (n,γ) is the highest for energies below 10 keV, making this reaction useful for the investigation of thermal and epithermal neutrons.
Results and discussion
Experimental and simulated flux in the carrousel
Experimental and simulated flux in the vertical tube
Variation of the experimental and calculated thermal neutron flux, in cm -2 s - 1 ( ϕ th ), and the factor f ( ϕ th / ϕ epi ) in the vertical channel of TRIGA reactor with the distance Z from the core
4.7 x 108
5.8 x 107
2.2 x 108
5.6 x 107
2.8 x 107
2.4 x 107
1.8 x 107
5.2 x 107
5.9 x 106
7.7 x 106
5.6 x 106
5.1 x 108
1.6 x 108
4.3 x 107
2.0 x 107
1.5 x 107
7.2 x 106
6.3 x 106
4.9 x 106
The relative error of the simulation compared to the experimental data, (Δexp), varies from 9 to 27% and the MCNP intrinsic code error, (Δcod), is lower than 10%. The code error can be improved, but it requires an upgrade of the computational resources or a huge time of processing, one week or more, which is not feasible. The (Δexp) is due to the conjugated effects of the intrinsic code error and the experimental error together. The variation of the thermal flux, higher than it could be expected for the considered positions, is due to the uncertainties in several factors: thickness of the Au monitor and cadmium capsule, different efficiency of the used detectors, variation of the reactor flux during the experience and others. From previous data of thermal neutron flux performed periodically in the carrousel, using similar procedures, it can be attributed a maximum experimental error of 10%, Zangirolami et. al. (2010).
The relative high error resulting for the factor f (ϕth/ϕepi) is due to difficulties for obtaining a more accurate experimental estimation of the epithermal flux. The evaluation of f parameter is important to give an idea of the thermalization of the beam. Because the neutron cross section are larger and the number of possible interfering reactions smaller, at lower energy, ideally, the neutrons for PGAA must have energy lower than the thermal level, Molnar (2004). One source of error is the irregularities in the kind of cadmium capsules used for the measurements and the uncertainty in the distance between the core and vertical tube, fulfilled by water. Small changes in the value of this distance modify dramatically the value of epithermal flux in the tube. A visual inspection of the pool and a new set of experiments are planned to improve the model of the system. The experiment (1) was not considered in the comparison in the Table 1 because the configuration of the core was different when the experiment was performed. From Figure 9, it can be observed a same behaviour for the calculated and measured thermal neutron flux from the experiments (2) and (3).
Simulated flux in the inclined tube
Variation of the calculated thermal neutron flux, in cm -2 s -1 (ϕ th ), and the factor f (ϕ th /ϕ epi ) in the inclined tube of TRIGA’s reactor with the vertical distance Z from the core
The relative error of the simulation Δcod(%), is lower than 11%, which can be considered satisfactory for this preliminary model. Ideally, Δcod(%), should be lower than 5%, but it requires a very high computational time due to the necessary increasing of number of neutron histories and cycles, Dalle (2005)
From these results, it can be expected, in the worst case, a thermal flux, ϕth, of 106 cm-2 s-1 in the highest possible position of the sample/detector, located approximately 3.5 m height from the reactor room level. If the set sample/detector can be positioned closer to the pool and the floor a double flux, ~2.0 × 106, will be possible.
A still higher flux in the sample/detector can be obtained increasing the diameter of the tube, but in this case, the tube must be shifted for a more external position of the core, diminishing the effect of the higher diameter. This shifting also depends on mechanical adjustments of the tube with the core and the top of the pool and the configuration of the shielding. If the reactor operates at 250 kW, in the future, a thermal flux of magnitude 107 can be obtained in the position of the sample/detector.
The measured and calculated thermal flux in the vertical neutron beam and the simulation of a proposed inclined tube from the core in the reactor pool, confirm that the installation of a PGAA facility at the TRIGA reactor can be feasible. In the worst case, considering an inclined tube of 5 cm of diameter, a minimum neutron flux of 106 would be possible in a sample/detector position located 3.5 m above the reactor room floor.
According the calculations, the flux can be increased until 107, if the rack containing the set of shielding/sample/detector can be positioned closer to the reactor pool; if a tube with higher diameter can be mechanically adapted in the core and in the pool´s surface and if the reactor operates at 250 kW. A best statistic performance of simulation with a lower Δcod(%), will require the improvement of the computational resources, including additional and faster processors.
Experimental measurements under way, of the neutron flux in the inclined tube and neutron and gamma dose in reactor pool will define the optimum design for the shielding and will establish an accurate minimum-maximum range of the possible neutron flux in a sample/detector set. The preliminary results present here suggest the possibility of using the reactor, at least, for some PGAA applications. Additional experimental data are necessary to make a more complete evaluation of cost/benefit.
The authors thank the Brazilian Agencies FAPEMIG and CNPq for their financial support.
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