NEUTRONIC ANALYSIS OF A FUEL ELEMENT WITH VARIATIONS IN FUEL ENRICHMENT AND BURNABLE POISON

In this work, the goal was to evaluate the neutronic behavior during the fuel burnup changing the amount of burnable poison and fuel enrichment. For these analyses, it was used a 17 x 17 PWR fuel element, simulated using the 238 groups library cross-section collapsed from ENDF/BVII.0 and TRITON module of SCALE 6.0 code system. The results confirmed the effective action of the burnable poison in the criticality control, especially at Beginning Of Cycle (BOC) and in the burnup kinetics, because at the end of the fuel cycle there was a minimal residual amount of neutron absorbers ( 155 Gd and 157 Gd), as expected. At the end of the cycle, the fuel element was still critical in all simulated situations, indicating the possibility of extending the fuel burn. .


INTRODUCTION
The burnable poison (or simply poison) is a substance, which has a high neutron absorption cross-section. Then it reduces the reactivity of a reactor core [1]. These neutron absorbers are chosen because they transmute by neutron capture into isotopes with low capture cross sections, somewhat faster than fuel burnup, thus leaving a residual minimum amount of burnable poison at the end of the fuel cycle. [2]. Such characteristics of burnable poison improve fuel utilization, contribute to a more even distribution of power in the reactor core, and are able to control nuclear reactivity.
Burnable poisons are rare on earth and they are particularly useful for the control of nuclear reactivity, among which are the elements Samarium (Sm), Europium (Eu), Dysprosium (Dy), Erbium (Er) and Gadolinium (Gd).
The presence of burnable poison in adequate amounts reduces the use of control rods [ 3]. This practice reduces the amount of actinides and fission products, and does not change the thermal conductivity [ 4]. The burnable poisons remove neutrons by absorption and thus effectively reduce the excess in the nuclear core reactivity. This effect is generally desirable at BOC, due to the excess of reactivity at loading a new nuclear fuel; therefore, a high concentration of burnable poisons is necessary. The best way to improve the fuel utilization is based on the burnup extension and poison kinetics at BOC, since the lowest remaining residual concentration of burnable poison is expected at the end of the cycle in order to stop neutrons absorption and to reduce further plus core reactivity. Cochran and Tsoufanidis (1999) [ 5] reported that gadolinium (Gd) seems to be more attractive because it can be mixed directly with UO2. It has several isotopes and the natural abundance of 155 Gd and 157 Gd are 14.7% and 15.7%, respectively. Their absorption cross sections for thermal neutrons are 5.8x10 4 and 2.4x10 5 barns, respectively.
A relevant way to improve nuclear fuel efficiency is using Gd by enriching the percentage of natural isotopes 155 Gd and 157 Gd [ 6]. This enrichment can eliminate the presence of parasite absorber at EOC [7]. Although Schlick (2001) [ 8] reported that the contribution of Gd2O3 in values close to 2% does not affect the thermal conductivity of the fuel, is known that the amount used in PWR is generally higher than these values.
The aim of this work was to evaluate the neutronic behavior during burnup changing the amount of burnable poison and fuel enrichment. Neutronic parameters such as infinite multiplication factor and composition of burnable poison, both during burnup and at EOC, have been analyzed. The gadolinium pins were simulated in a homogeneous mixture with uranium oxide (UO2) containing 0.2% of 235 U and with different proportions of gadolinium in the mixture of UO2 + Gd2O3.

Description of the modeled system
The simulations were carried out considering a Pressurized Water Reactor (PWR) fuel element arranged on an assembly of 17 x 17, with a central guide tube, which makes a total of 24 guides tubes [ 9] , but without boron, initially, without Burnable Poison Rods (BPR) and after with 16 BPR. The rods of UO2 have enrichment described in Table 1, according to the literature [9 and 10]. For fuel element analysis of the benchmark, was used the nuclear code SCALE 6.0 [11], altough 238 groups collapsed from ENDF/BVII.0 library of cross section [ 12]. 2.00% 4.00% 6.00% 8.00% Figure 1 shows fuel rod assembly with guide tubes, with 16 BPR and without BPR, and was generated using the SCALE 6.0 code. To calculate the infinite multiplication factor (kinf), geometry of the bundle model was reflected from all sides. In this way, neutrons are not allowed to escape from the system. Figure 2 shows pellets fuel, fuel rod and fuel element.    Using the same methodology and maintaining the same geometry, 16 rods of UO2 were replaced by burnable poison. Firstly, the arrangement was studied without BPR e after with 16 BRP, in proportion gadolinium 2.00%, 4.00%, 6.00% and 8.00%, as mentioned in table 1. This study was developed with fuel depletion calculations using the nuclear code SCALE 6.0.

Simulations: fuel depletion with SCALE 6.0
The SCALE code estimates the infinite multiplication factor (kinf) with the respective standard deviation of the model (σST). This work evaluated: (a) criticality the system with different enrichment and different proportion burnable poison during fuel depletion and (b) impact of burnable poison in the kinf.
For fuel depletion calculation, TRITON module [ 14] was used through T6-depl command with 10000 particles and 2200 generations. The library ENDF/BVII.0 was used with 238 collapsed groups (V7-238), because it presents less deviation from the average, besides being able to be used in the calculations during fuel evolution [ 15]. The specific power density was 38 W/gU during 789.48 days, producing an overall burnup of 30 GWd/tHM [ 9]. To analyze the burnup and the impact of burnable poison in the reactivity, the time was divided at smaller intervals, including periods of decay, as shown in Table 3.
During fuel evolution , infinite multiplication factor (kinf) was compared as a function of gadolinium, especially 155 Gd and 157 Gd, as isotopes with the greatest abundance, 14.7% and 15.7% and the highest absorption cross section for thermal neutrons equal to 5.8.10 4 and 2.4.10 5 barns, respectively [ 5].

Figures 3 and 4 show the variations of
kinf values changing fuel enrichment and proportion burnable poison. The standard deviation estimated by the used code has magnitude of 10 -4 . All cases, the main difference between graphics is predominant at the Beginning Of Cycle (BOC), period of effective action of the burnable poison.
It is important to emphasize that the value of kinf is always lower in all simulations, during the period when there is a higher concentration of burnable poison, as shown in Figures 3 and 4.. This allows reduction of use of control rods at the Beginning Of Cycle (BOC). .    Table 4 shows the initial and final composition of 155 Gd and 157 Gd, during burnup of the fuel As observed the final amount is small compared to the beginning of the cycle.

CONCLUSIONS
This neutronic study confirmed the importance of burnable poison in the control of reactivity without use of control rods, especially at the beginning of the cycle, when there is the greatest excess of reactivity during all the burnup.
This study showed that the consumption of burnable poison does not occur homogeneously, because, at the end of the cycle, there is a larger amount of 155 Gd, not maintaining the observed proportion at the beginning of the cycle.
The next study is to make comparisons of the results with other nuclear codes such as MCNP-X and Monteburns and to use enriched burnable poison in order to achieve better efficiency in their use as neutron absorbers, besides to try to establish a relation between the enrichment of the burnable poison with the amount of actinides generated.