A high-conductivity oxide fuel concept

The relatively low thermal conductivity of uranium dioxide (UO2) has a significant impact on the performance of commercial nuclear fuels. It is accepted as a necessary compromise because UO2 fuel pellets are readily fabricated and have performed well in nuclear energy systems over the past 40 to 50 years. However, there is an emerging need to push modern fuels to higher power ratings, higher burnup, and improved performance while maintaining safety and reliability [1-3]. One means of enabling current LWR nuclear energy systems to meet this need is to enhance the thermal conductivity of the UO2 using high-conductivity ceramic additives to the fuel microstructure.

An innovative nuclear fuel approach that uses beryllium oxide (BeO) as an engineered additive is under development to enhance the thermal conductivity of UO2. Initial development research at Purdue University evaluated silicon carbide (SiC) and BeO additives as part of a US Department of Energy Nuclear Energy Research Initiative [4-6]. In that study, BeO was selected over SiC, in part, because of the higher temperature stability of the UO2–BeO mixture. Lab-scale processing methods were established to create a novel UO2–BeO fuel microstructure (patent pending) with exceptional thermal behaviour [4] and a thermal model was created to predict the conductivity of this novel material [5]. It has been known for some time that the addition of BeO to UO2 improves the ceramic’s thermal conductivity [7]. The uniqueness of this new concept arises from the tailored ceramic-ceramic composite microstructure that enhances thermal conductivity as well or better than previous UO2–BeO concepts.

Concept and fabrication

Figure 1 shows a schematic description of the UO2–BeO structures generated during the laboratory-scale experiments [4-6]. The fabrication process comprises powder methods similar to commercial fuel fabrication methods; proprietary engineering-scale improvements to lab-scale methods are currently under development (patent pending). The structure consists of primary UO2 microspheres (50 to 500 μm) embedded in a mixed-oxide matrix containing fine grained UO2 and BeO. The BeO matrix is continuous; this feature was confirmed when a representative sample was exposed to hot nitric acid, dissolving away the entire UO2 content of the pellet; the continuous BeO network was intact and relatively strong.

Lab-scale UO2-BeO samples were fabricated using depleted UO2 according to the process flow diagram shown in Fig. 2 [4]. The UO2 powder was pressed and granulated to produce green granules. The granules were then self-milled and mixed with fine BeO powder and the final mixture was pressed and sintered to form dense pellets for microscopy and thermal diffusivity measurements. Figures 3 and 4 show images from a lab-scale UO2–BeO sample where the primary microspheres and oxide matrix are clearly visible. The dark phases are BeO and the bright phases are UO2.

Neutronic simulations

Neutronic pin cell simulations for a typical PWR were performed using the lattice physics code DRAGON [12]. DRAGON is a single code with various numerical techniques and calculation methods to solve the neutron transport equation. Primary modules include resonance self-shielding calculations, analysis of various geometries to generate a tracking file for collision probability evaluation; multigroup collision probability integration; solution of the of the multigroup neutron transport equation using the collision probability method; solution of the multigroup neutron transport equation using the method of characteristics; isotopic depletion. Development support comes from École Polytechnique de Montréal, Hydro-Québec and the Hydro-Québec chair in nuclear engineering, the Natural Science and Engineering Research Council of Canada (NSERC), Atomic Energy of Canada limited (AECL) and the CANDU Owners Group (COG). The DRAGON simulation was a two-dimensional infinite lattice comprising one-eighth of a fuel assembly.

Table 1 contains the basic data and parameters used for the computations. To effectively perform these simulations, estimates were required for the ‘effective’ fuel temperature, Teff, for each fuel composition; Teff is the representative temperature used to compute the average neutronic behaviour of the fuel pin and is an average of the fuel centreline and surface temperatures.

The fuel temperature profiles presented in Fig. 5 were determined using the conductivity properties of UO2 and UO2–10 vol% BeO fuels. For the parameters listed in Table 1, nominal values for Teff were estimated as ~627°C (900 K) for UO2 and ~527°C (800K) for UO2-10 vol% BeO.

A set of parametric studies was conducted to consider the impact of uranium-235 mass displacement on reactivity, the sensitivity of the fuel concept to the estimation of Teff, and to generate a first estimate of the temperature reactivity coefficients.

Core cycle length studies were performed for two scenarios: a 3-batch core reload strategy and a 4-batch core reload strategy for both fuel types: the UO2 fuel with the BeO additive, and the pure UO2 fuel without the BeO additive. For each scenario using standard UO2 fuel, the enrichment in uranium-235 was computed to achieve criticality at the core average burnup. A value 1.035 is chosen to take into account the approximate 3.5% neutron leakage from the reactor core that it is not accounted for in the infinite lattice physics simulations performed with DRAGON.

The mass equivalence studies compared two scenarios: a fuel with the BeO additive, and a fuel without the BeO additive, but in both cases there were an equal amount of uranium-235 atoms. In this way, the effect of the BeO is more easily apparent. The first step was to determine the enrichment of a pure UO2 fuel that reaches a multiplication factor of 1.035 at the specified average burnup (3.86 wt% for case 1 and 4.75 wt % for case 2). The value 1.035 is chosen so as to take into account the approximate 3.5% neutron leakage from the reactor core that it is not accounted for in the lattice physics simulations performed with DRAGON.

The first step was to determine the enrichment of a pure UO2 fuel that reaches a multiplication factor of 1.035 at the core average burnup (3.86 wt% for scenario 1 and 4.75 wt % for scenario 2). For the UO2 fuel with the BeO additive, the same mass of U-235 was used. Since 10 vol% was taken up by BeO, a higher enrichment was required to keep the mass of U-235 equivalent for side-by side fuel comparisons (4.29 wt% for scenario 1 and 5.29 wt % for scenario 2). For the UO2-BeO fuel, in both scenarios, the lower fuel temperature and improved moderation due to Be increased the beginning-of-cycle (BOC) fuel reactivity (2800 to 2900 percent milliRho, pcm). The n,2n reaction from Be also provided a minor boost in neutron production. This resulted in a slightly extended equilibrium fuel cycle length of about 20 days with a notable increase in end-of-cycle (EOC) burnup (4000 to 5000 MWd/tHM). This increase in cycle duration was achieved despite a steeper reactivity swing from a higher power density and it was also achieved without design manipulations such as burnable poison. Alternatively, this potential increase in achievable burnup could be manipulated to enable the reduction of BOC enrichment (or more precisely, the overall U-235 content in the fuel) to maintain a given fuel cycle duration.

The reactivity coefficients, fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC) were computed using DRAGON. The FTC value estimated for this system with UO2 was estimated to be -3.44 pcm/K. The FTC value of this system with UO2–10 vol% BeO decreased to -3.71 pcm/K. This was a small change, but it is more negative, which is a benefit for safety. For the MTC at 0 ppm of boron, the estimated values were -56.6 pcm/K for UO2 and -52.7 pcm/K for UO2–10 vol% BeO. This is an increase, but it is still in a range that is comparable to UO2. The important point is that the FTC and MTC shifts arising from BeO additions to UO2 are relatively small (7 to 8%); The FTC is slightly improved and the MTC is slightly degraded.

From a safety point of view, it is also important to note the differences in the temperature profiles shown in Fig. 5. This particular case compares 5% enriched UO2 with 5% enriched UO2–10 vol% BeO. The coolant, cladding and gap temperatures are identical in both cases, but the centreline temperature in the higher thermal conductivity is ~150°C lower than the nominal UO2 case. The ΔT would be ~250°C in a peak fuel pin with a peaking factor of 1.7. The implication of this result is that potential safety benefits (that is, less fission gas swelling and release and reduced stored energy) arising from lower temperatures may be achieved.

Future work

The experimental work carried out thus far provides a solid characterization of unirradiated material. Questions regarding fuel restructuring, fission gas behaviour, and thermal conductivity changes with irradiation require irradiation performance testing. In addition, parametric studies to optimize the BeO concentration (10 vol% may be too high for practical use) must be completed.

The UO2­–BeO concept described here has matured to the point that its development must move beyond its present academic context and proceed toward implementation with industrial collaborators. This will involve continuing and expanding the above activities to evaluate the commercial viability and industrial performance of this fuel (including irradiation tests). If the fuel lives up to its promise, it has the potential to create a valuable commodity for the nuclear industry worthy of eventual licensing by the US Nuclear Regulatory Commission.

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FilesFig. 2: Fabrication process flow diagram (large)
Tables

Table 1: PWR specifications used for simulations

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