Multiple Pu recycling1 February 2003
The Nuclear Energy Agency has carried out a benchmark exercise to address the number of times plutonium can effectively be recycled in a PWR. The findings are reported in "Physics of Plutonium Recycling".
The recycling of plutonium in the form of MOX fuel is well established on a commercial scale. The discharge burnup of MOX fuel is essentially the same as that of UO2 fuel. Thus the MOX fuel currently used in PWRs is intended to be discharged at burnups of 40-45MWd/kg. The initial plutonium content needed to achieve such burnups varies depending on the precise source of the plutonium. The experience of MOX use in PWRs has been positive, and there are no outstanding<
However, the situation is not static. The fundamental changes are that discharge burnups are increasing, and that there will be a greater need to recycle plutonium from discharged MOX assemblies. Both of these changes will result in a decrease in isotopic quality of the plutonium available for recycling. For thermal reactors, the even isotopes of
plutonium (238, 240 and 242) do not contribute significantly to fissions. The ratio (Pu-239 + Pu-241)/(total plutonium) is the fissile fraction of the plutonium, and is a measure of plutonium quality for thermal reactor MOX. The problem is that plutonium quality decreases as the discharge burnup increases, and it decreases yet further after the recycle of the plutonium recovered from MOX. It will therefore be necessary to significantly increase the total plutonium content of the MOX fuel.
Compared with conventional UO2 fuel, MOX fuel is already significantly different from a neutronic point of view, with a much smaller thermal flux for a given rating. This is due to the combined effects of the higher fission and absorption cross-sections of the Pu-239 and Pu-241 compared with U-235, exacerbated by the significant absorption of the Pu-240 and Pu-242. The difference in spectrum affects the core performance because the control, reactivity coefficient and transient behaviours are all altered. Increasing the total
plutonium content beyond present levels exaggerates all these effects. The deterioration in parameters such as control rod reactivity, boron reactivity and moderator void and temperature coefficients may become a barrier to further use of MOX in PWRs, at least in conventional lattices.
The question of how many times plutonium recovered from MOX assemblies can be reused is important strategically and logistically. Strategically, it is important because it affects the energy potential available from plutonium. Logistically, it determines whether there will be a need to store or dispose of MOX assemblies or plutonium or both if indefinite recycling does not prove practical.
Each recycle generation involves irradiation of MOX fuel (typically lasting 4-5 years), followed by pond cooling (typically 5 years), followed by reprocessing and refabrication as MOX (taking 2 years). Thus each generation of multiple recycle will last at least 11 years. Multiple recycle scenarios extend over very long periods measured in decades. Over such timescales, there may well be major changes in world energy requirements and strategies. As a result, scenarios that are considered and analysed may be overtaken by events before the early generations of recycling are completed. Nevertheless, it is important to analyse such scenarios, in order to be certain that they are practical technically, strategically, logistically and to establish their impact on environmental and safety considerations.
The present benchmark is for a standard lattice PWR and a highly moderated PWR operating with a moderately high burnup fuel cycle (51MWd/kg), which should be representative of PWR operation in the next decade. The specification calls for the reprocessing of MOX fuel along with a certain fraction of UO2 assemblies. The emphasis was on specifying a benchmark problem that was as realistic as possible, but keeping within the bounds of what is known from current technology and not relying on extrapolation to an uncertain future. This constraint may mean that the benchmark may turn out to have been pessimistic by not accounting for technological developments, but this is unavoidable.
The primary objectives of the benchmark are:
•To compare reactivities, reactivity coefficients and isotopic evolution calculations obtained with different lattice codes and their associated nuclear data libraries.
•To determine at what point, if any, the calculations diverge to such an extent that the physics predictions must be considered unreliable.
•To determine at what point, if any, further generations of recycle are excluded on technical grounds such as unacceptable reactivity coefficient characteristics.
•To evaluate the environmental impact of multiple recycle.
The rationale behind including a highly moderated PWR lattice is that, theoretically, such a lattice may show technical advantages in multiple recycles. The idea would be to dedicate a small number of new PWRs to MOX usage only. With no need to accommodate UO2 fuel, the reactor designer could then choose to optimise the moderator/fuel volume ratio for plutonium. The optimum occurs at a moderator/fuel ratio of about 3.5, compared with 2.0 for uranium fuel in a standard lattice. This could be achieved by preserving the fuel rod design and dimensions, and simply increasing the rod-to-rod pitch. The reactor core would be marginally larger in its radial dimensions, but otherwise the reactor design and the associated equipment would be similar to that for a conventional PWR.
Two pin cells were examined (see Figure 1). The first was representative of a modern PWR, specifically representing a 17x17 MOX assembly with 25 guide and instrument tubes and 264 fuel rods. The second pin cell was intended to represent a proposed assembly design for a highly moderated MOX fuelled PWR.
For the three main material components of fuel pellet, clad and moderator, the following observations apply:
The fuel pellet consists of UO2/PuO2 MOX at a density of 10.02g/cm3. This corresponds to a nominal density of 95% of the theoretical density, with allowance for the loss of volume due to end dimples and pellet chamfers. The fuel temperature is representative of nominal full power operation.
The clad is assumed to be composed of natural zirconium. In reality, Zircalloy is used, but ignoring the minor alloying elements has no implications for the benchmarks. The temperature is again representative of full power operation.
The moderator is light water at a density of 0.71395g/cm3, representative of the average density in nominal full power operation. At the nominal working pressure of 155.5bar, this density corresponds with a moderator temperature of 306ºC. The specification calls for the pin cell depletions to be carried out with 500ppm dissolved boron. This is roughly the lifetime average value in a modern PWR. It is important to carry out the depletion in this manner in order to correctly capture the average spectral history effect.
The fuel depletion specifications are the same for each recycle generation. The underlying assumption is that fuel management remains the same throughout all five recycle generations considered. This is unlikely to be the case in practice.
Fuel management is a three-batch 18-month fuel cycle, with a discharge burn up of 51MWd/kg. Thus at each refuelling outage, one third of the fuel assemblies resident longest in the core are discharged and replaced with fresh fuel. Such a fuel management scheme is a little outside
current experience with both UO2 and MOX, but it is realistic for a few years hence. Since each refuelling cycle lasts for 18 months, the complete fuel depletion lasts for 4.5 years.
Multiple recycle logistics
The benchmark calls for the plutonium to be followed through five recycle generations.
The first generation simply involves taking the plutonium obtained from reprocessing a number of UO2 assemblies and recycling it as MOX. The UO2 assemblies contain 12.4kg/tU total plutonium at discharge. With the isotopic composition of this plutonium, the initial plutonium content of the first generation MOX needed to achieve the 51MWd/kg fuel cycle was determined to be 101.5kg/tHM. This value was taken to be the specification for the first generation MOX assembly.
In the second, third, fourth and fifth generations, the plutonium from the MOX assemblies was assumed to be recovered by co-reprocessing with conventional UO2 assemblies in a ratio of 3 UO2: 1 MOX. There are two reasons why this is appropriate.
Firstly, it is not expected that reprocessing plants specifically dedicated to MOX will be built, so that the MOX assemblies would have to be reprocessed in plants primarily intended for reprocessing UO2. It would be uneconomic to recycle MOX assemblies on their own, as well as being technically impractical, due to heat generation in the reprocessing liquors (which is higher in MOX) and activity level from minor actinides (also higher).
Secondly, it is not intended to irradiate MOX assemblies alone in the core; there will also be UOX assemblies. Therefore, a utilitiy's spent fuel will consist of a mix of the two, and it would make sense to reprocess them together.
Therefore, in second and later MOX generations, the plutonium from each MOX assembly derives from a number of MOX assemblies and three times that number of UO2 assemblies.
The MOX assemblies, at discharge, contain upwards of 80kg/tHM of plutonium, compared with just 12.4kg/tU for the UO2 assemblies. Thus in terms of plutonium, the dilution factor is much smaller than 3:1, and this means that the degradation of plutonium isotopic quality proceeds quickly in spite of the dilution.
The logistics between generations depends on the initial plutonium content needed in the next generation of MOX and the plutonium mass at discharge of the previous generation.
End of cycle reactivity
The concept of equivalent end of cycle reactivity is used to determine the initial plutonium content of a MOX assembly. A utility would usually wish a MOX assembly to substitute directly for a UOX assembly. Although there are occasionally exceptions, this normally means that the MOX assembly will contribute the same reactivity as a UOX assembly when averaged over its lifetime in the core.
For a three-batch 18-month fuel cycle used in this benchmark, in the equilibrium fuel cycle, there will be one third of assemblies having been irradiated for one cycle, one third of two-cycle assemblies, and one third three-cycle assemblies. Making the simplification that the reactivity versus burnup is linear, and that the burnup accumulated in each cycle is the same, the reactivity/burnup relationship of a UO2 and a MOX assembly are shown in Figure 2.
The reactor will need to be refuelled when the average k-infinity of the three regions (the average of points A, B and C for the UOX assembly and a, b and c for the MOX assembly) is just sufficient to cover neutron leakage from the core.
Isotopics and concentrations
The benchmark specification called for the first MOX generation to start off with a total plutonium content of 10.15 w/o (101.5kg/tHM) and isotopic composition of the plutonium as shown in the Table below. The specification called for the multiple recycle strategy to be carried out in two separate stages. In the first stage, the isotopic number densities are imposed, and in the second stage they are individually calculated.
The benchmark specification called for reactivity coefficients to be calculated in each generation at irradiations of 0.5, 17, 34 and 51MWd/kg. Being able to calculate reactivity coefficients is essential if multiple recycling is to be achieved. It is likely that if any limitations to multiple recycling are identified, it will be the reactivity coefficients which will be the determining factors.
The reactivity coefficients examined were the following:
•Boron efficiency, calculated by perturbing the dissolved boron concentration from 500 to 600ppm. The boron coefficient is important for reactivity balance during normal operation and for emergency shutdown in a PWR.
•Moderator temperature coefficient calculated by perturbing the moderator temperature and densities between hot and cold states.
•Fuel temperature coefficient calculated by perturbing fuel temperature between 890 and 910K.
•Global void effect calculated by changing the moderator density from 100% to 1% of nominal density. The calculation is carried out both with a no leakage spectrum and with a critical leakage spectrum in which the buckling parameter has been adjusted.
Lifetime average reactivity
For a standard PWR, the initial plutonium in the first generation was defined as a common starting point in the specification of the benchmark. For the highly moderated PWR, participants in the exercise were asked to calculate the initial plutonium in all generations, including the first.
It was found that the rate of increase in initial plutonium between generations is much higher for the highly moderated PWR than for the standard PWR. The reason for this is that the highly moderated PWR is better at fissioning the Pu-239 and Pu-241 in the soft spectrum, so that the isotopic quality degrades more quickly between generations. Thus the highly moderated PWR starts off with a low initial plutonium concentration in the first generation, but quickly catches up with the standard PWR case. This is significant, as the initial advantage of the highly moderated PWR is eroded in later generations.
The mass balances of Pu, Am and Cm are important, as they are a determining factor for the environmental impact.
The plutonium mass balance is negative for both the standard and the highly moderated cases, due to the destruction of the odd fissile isotopes. In both cases, the plutonium destruction rate increases with recycle generation, with the highly moderated PWR being more effective at destroying plutonium.
The mass balances for Am and Cm are positive, meaning that they accumulate with burn up, and the build-up rate is higher in the later recycle generations. This applies to both cases, and is a result of the higher initial plutonium concentrations needed in later generations; the higher initial plutonium mass leads to accelerated production of the trans-plutonium nuclides.
Detailed comparisons of fission rates of the isotopes in the fuel identified the thermal range of Pu-239 as the main source for observed differences. The deviations originating from Pu-241 are more moderate.
Significant contributions to the spread in eigenvalues are caused by the absorptions of the fissile isotopes Pu-239 and Pu-241, and in addition by Pu-240. The resonance absorption of U-238 is still a source of discrepancies.
A surprising conclusion resulted from the comparison of the absorption rates of non-actinides. Calculations have shown 110pcm higher absorption rates in clad material, and 100pcm lower absorption rates for oxygen in fuel, and 80pcm lower in the moderator. The reason for the lower oxygen absorption is a different threshold of the (n,alpha) cross-section in the data used.
Differences emerging from cross-section data used, their processing and differences in the computational schemes still lead to unsatisfactory discrepancies in calculating PWR MOX pin cells by different institutions, although "state-of-the-art" data and methods were used. This means that there is still a need for further analyses of the differences of best estimate solutions with respect to measured data and data processing, and that there is also a need for improvement in deterministic methods and basic nuclear data, as well as for better experimental validation.