Investigation of 244 Cm dynamics production in a WWER-1200 reactor for fuel based on 235 U and MOX fuel

244Cm is a very important nuclide due to the high specific heat release and it may become reasonable to recycle the spent nuclear fuel to extract this isotope. In this work, the dynamics of 244Cm isotope production for the fuel assembly of the WWER-1200 reactor with an enrichment of 4.95% and burnup up to 70 MW · day/kg was calculated using Serpent Monte Carlo code. A similar calculations was performed for MOX-based fuel assembly with the same irradiation characteristics. It has been shown that production of 244Cm in high burnup uranium dioxide fuel, and especially in MOX fuel, reaches a high value, which can cause problems during spent fuel management. The main problems associated with the curium relates to the residual heat and neutron activity.


Introduction
During operation of a nuclear reactor nuclear fuel is exposed to a high-density neutron flux [1]. 238 U contained in large quantities in the uranium dioxide fuel captures neutrons. The resulting products also continue to capture neutrons, which leads to the production of a significant number of nuclides with mass numbers of 239 and above. These nuclides have a high impact on the neutron-physical characteristics of the nuclear fuel, especially when studying the issues of spent nuclear fuel (SNF) management. In particular, 239 Pu and 241 Pu are the main isotopes producing in the fuel and providing a significant increase of the fuel cycle duration [2].
For many decades the fuel burnup did not exceed 50 MW · day/kg. Only in the recent years, fuel of a new reactors types began to reach burnup of about 70 MW · day/kg. A relatively small increase in the fuel burnup (about 40%), as shown in calculations, can lead to a significant increase in the actinide production (several times and more). The calculations presented in this work shows that 244 Cm in the fuel with high burnup (especially in the MOX fuel) will have one of the defining values for the residual heat amount in the SNF (this leads to a serious problem associated with the SNF management). 244 Cm is a very important nuclide, due to the high specific heat release [3], and it may probably become reasonable to recycle the SNF to extract this isotope for its further usage.
Curium-244 isotope production scheme 244 Cm isotope in thermal-neutron reactors is almost completely accumulated from 238 U by capturing six neutrons sequentially. A simplified scheme (it does not show β -decays and fission reactions) of the 244 Cm during operating time is presented in Figure 1. The production rate of each successive nuclide depends on the actual amount of the preceding nuclide. Thus, a significant rate of the particular nuclide production reaches only after building up the entire chain of its predecessors. This leads to the fact that the nuclides such as 244 Cm (with chains similar to shown in Figure 1) have a strongly nonlinear production dynamics.
The amount of 244 Cm produced in a thermal nuclear reactor with uranium dioxide fuel is almost completely determined by the burnup value. The burnup dynamics have a little impact on the result due to the fact that all nuclides in the chain have a half-life significantly longer than the duration of the fuel cycle. 238 U isotope concentration in uranium dioxide fuel is about the same for different types of fuel assemblies and is only slightly dependent on burnup.

Calculation model
Serpent Monte Carlo code was used for burnup calculations [4]. The computational model consists of a single fuel assembly of the WWER-1200 reactor with reflective boundary conditions and 235 U enrichment of 4.95% [5,6]. ENDF/B-VII [7] library of evaluated nuclear data was used in this calculations. Model of fuel assembly is shown in Figure 2. Calculations were carried out for burnups from 0 to 70 MW · day/kg. This burnup values is a preliminary estimate for the WWER-1200 fuel [8]. For standard types of WWER-1000 fuel assemblies, the maximum burnup value was about 40 MW · day/kg [9] (61.8 MW · day/kg using experimental fuel assemblies and up to 72 MW · day/kg for single fuel elements [10,11]).

Cm-244 production in thermal nuclear reactors with uranium dioxide fuel
Obtained dependence of the specific content of 244 Cm depending on the burnup is shown in Figure 3.
As can be seen, at the initial stage of the fuel cycle and up to a burnup of 20 MW · day/kg, the 244 Cm isotope is practically not produced. This is due to the formation of the intermediate nuclides in the chain, after which curium production begins to occur relatively intensively.
In the fuel with burnup of about 40 MW · day/kg the amount of 244 Cm produced during operation is several tens of grams per ton and does not make a significant contribution to the radiation or thermal characteristics of the SNF. In this sense, the WWER-1200 reactor with burnup up to 70 MW · day/kg becomes the first practically used reactor type in which the problem of curium is relevant.
It is important to note that the only channel for reducing the amount of 244 Cm in a nuclear reactor is the reaction of radiation capture (except for α -decay, with a half-life of 18 years) with a very low cross section of about 10 barn for thermal neutrons [3]. For this reason, the amount of 244 Cm in reactor core decreases very slowly.

Cm-244 production in thermal reactors with MOX fuel
The rapid accumulation of 244 Cm in uranium dioxide fuel after building up the chain of precursors (Figure 1), starting from burning out 20 MW · day/kg and above suggests a high rate of production of this isotope in MOX fuel, where this chain is already built. To estimate the 244 Cm production time in MOX fuel, a similar model was created for a fuel based on uranium oxide, but using MOX fuel. The composition of MOX fuel was determined as follows: instead of enrichment natural uranium is diluted with plutonium and the 239−242 Pu isotope proportion that which is contained in the spent uranium fuel of the WWER-1200 reactor at the end of the fuel cycle (the assumption is based on the potential spent fuel of the WWER-1200 reactor management with the emission of plutonium from it). The amount of plutonium in MOX fuel was calculated on the basis that the resulting concentration of fissile isotopes ( 235 U , 239 Pu , 241 Pu ) is the same as in uranium fuel with an enrichment of 4.95%. MOX fuel consist of uranium and plutonium dioxide: UO 2 and PuO 2 . The composition of the fuel thus obtained is presented in Table 1. Production of 244 Cm isotope depending in the fuel burnup (composition of fuel in Table 1) is shown in Figure 4. As can be seen, 244 Cm production in MOX fuel can reach a significant value up to kilograms per ton. Even with relatively small burnup's characteristic of the most WWER-type reactors, the mass of 244 Cm reaches 1 kg per ton.

Residual heat and neutron activity
High decay energy in combination with a relatively short half-life leads to the fact that 244 Cm creates a high heat release of ∼ 2.83 kW per 1 kg of 244 Cm . Figure  5 presents the residual heat of 1 ton of spent nuclear fuel from WWER-1000 and WWER-1200 reactors in comparison with the heat generated by 1 kg of 244 Cm [12]. On the other hand, 244 Cm is an important isotope due to the high heat release over a long period of time. 228 Pu used in the manufacture of electrical energy sources in space technologies [13][14][15] have a similar property.
The extraction of curium isotopes for subsequent use is a feasible task with the prospect of further development [16]. For example, in work [13] was discussed the issue of separating curium from spent fuel with a burnup of 33 MW · day/kg in which the production of 244 Cm was only 31.9 g/ton. At the same time, in the WWER-1200 reactor with uranium dioxide fuel this value is higher by an order. In the future, the curium production time can reach kilograms per ton by using MOX fuel.
Another important fact is that the 244 Cm isotope creates high neutron activity in the SNF due to spontaneous fission. The neutron activity of one gram of 244 Cm is ∼ 1.6 · 10 7 neutr./s [17]. 244 Cm also makes a significant contribution to the neutron activity due to reactions ( α , n) on oxygen 17 O and 18 O contained in the nuclear fuel [18].

Conclusion
The latest generations of light-water reactors with high burnup fuel (from 60 MW · day/kg and above) pose new challenges for researchers related to the spent nuclear fuel management. One of these tasks can be the solution of the problem with significantly increased 244 Cm isotope production. The mass of this isotope in the fuel with a burnup of 70 MW · day/kg exceeds the corresponding mass at a burnup of 40 MW · day/kg by about 6 times.
At the same time the small amount of 235 U isotope in natural uranium (is approximately 0.7%) require use of new types of fuel in the future. Usage of MOX fuel can be one of the solution for this problem. But as was shown by calculations it can cause serious problems with spent fuel management.