Potential methods to reduce the thermal impact of vitrified waste from spent mixed oxide fuel on geological repositories

Compared to the vitrified, high-level radioactive waste (HLW) from spent UO2 fuel (UO2-HLW glass), that of mixed oxide (MOX) fuel (MOX-HLW glass) exhibits high heat generation over a long period of time. Thus, to reduce the associated waste volume and repository footprint, it is essential to prevent bentonite buffer alteration by mitigating heat generation. In this study, we evaluated five potential methods of decreasing the bentonite buffer temperature in a geological repository for MOX-HLW glass. Of these five methods, the separation of minor actinides (MA) from high-level liquid waste (HLLW) was the most effective. This is because the main nuclide involved in heat generation in MOX-HLW glass is Am-241. An MA separation rate of 81% was required to maintain the bentonite buffer temperature below the Japanese upper disposal limit of 100 °C. Thus, it is important to measure the amount of Am-241 when disposing of MOX-HLW glass. Assuming that only Am-241 was disposed of, the maximum Am-241 waste loading under a bentonite buffer temperature of 100 °C was 0.33 wt%, equating to the disposal of only 1.32 kg of Am-241 per 400 kg of MOX-HLW glass. Thus, there is a need for new methods of treating and disposing of Am-241. Relationship between bentonite buffer temperature and (a) minor actinide (MA) separation ratio from high-level liquid waste (HLLW), (b) mixed oxide (MOX) high-level radioactive waste (HLW) glass and UO2 HLLW mixing ratio, (c) waste loading in MOX-HLW glass, (d) MOX-HLW glass storage time, and (e) canister footprint, representing five methods of mitigating heat generation from disposed MOX-HLW glass.


Introduction
Recently, Japan has begun storing spent mixed oxide (MOX) fuel discharged from commercial nuclear power plants [1]. Compared to spent UO 2 fuel 15 years after discharge, spent MOX fuel has 5.3 times the content of minor actinides (g/ tHM), generates 3.0 times more decay heat (kW/tHM) [2], and takes approximately 300 years to cool down before generating the same amount of heat. Therefore, it is necessary to develop an appropriate disposal process while considering the reprocessing and vitrification of spent MOX fuel to promote the use of Pu. It is particularly crucial to evaluate the amount of heat generated by vitrified waste during disposal. Minor actinides (MA) exhibit a significant influence on the disposal of vitrified high-level radioactive waste (HLW) from MOX fuel (MOX-HLW glass), as determined in our previous study [3]. In this study, we considered and evaluated five methods of decreasing the bentonite buffer temperature in a geological repository for MOX-HLW glass. These five methods were as follows: (1) separating MA from high-level liquid waste (HLLW), (2) mixing MOX-HLW glass with UO 2 and MOX-reprocessing HLLW, (3) low waste loading of MOX-HLW glass, (4) long-term storage, and (5) expanding the repository footprint. We evaluated the effectiveness of these five processes relative to existing methods of disposing vitrified waste from UO 2 fuel (UO 2 -HLW glass).

Initial Pu isotopic ratios in MOX fuel
The initial Pu isotopic ratios of the MOX fuel used in this study were calculated based on the composition of spent UO 2 fuel. The calculation conditions are listed under Supplementary information (Table S1). The number of nuclides generated in spent UO 2 fuel and extracted during reprocessing and the amount of decay were calculated using the ORIGEN 2.2-UPJ [4] and ORLIBJ40 [5] cross-section libraries. Considering that we evaluated Japanese nuclear reactors, we assumed that Pu derived from pressurized and boiling water reactors, respectively, was mixed at a ratio of 1:1. Some Japanese reference case conditions were used for other parameters, such as specific power [6,7]. The calculated Pu isotopic ratios in the MOX fuel are listed under Supplementary information (Table S2).   [4] and ORLIBJ40 [5], respectively. The five methods of reducing the bentonite buffer temperature each comprised different parameters. Only one parameter per method was varied with respect to standard conditions. These parameters were as follows: (1) MA separation ratio from HLLW, (2) MOX-HLW glass and UO 2 HLLW mixing ratio, (3) waste loading in HLW glass, (4) HLW glass storage time, and (5) canister footprint. All other parameters were fixed based on those of UO 2 -HLW glass in Japan [6][7][8]. We then proceeded to measure the effect of the five methods on heat generation in MOX-HLW glass at a 10.8 wt% waste loading [6,9], which is the standard waste loading of UO 2 -HLW glass in Japan. Additionally, we determined the parameter levels of the five methods at which the bentonite buffer temperature after disposal remained below the upper limit of 100 °C in Japan.

Bentonite buffer temperature
The temperature of the bentonite buffer overpack (the hottest region) was calculated via heat transfer analysis, using the heat generated as the input. Calculations were performed using COMSOL Multiphysics v. 5.6 (COSMOL AB, Stockholm, Sweden), assuming a disposal site. The engineered barrier design and near-field system for the vertical emplacement of the HLW and repository environment were based on the Japanese safe reference case [6] and exhibited the following parameters: a hard rock (crystalline) structure, an underground depth of 1000 m, and an initial surrounding temperature of 45 °C. The boundary conditions for heat transfer were as follows: 15 °C at the ground surface, 51 °C at the bottom of the analysis model, a temperature gradient of 3 °C per 100 m, and adiabatic on all sides.

Results and discussion
Methods of mitigating heat generation from MOX-HLW glass Figure 1 shows the changes in bentonite buffer temperature associated with representative parameters of the five methods of mitigating heat generation from MOX-HLW glass: (a) MA separation ratio from HLLW, (b) MOX-HLW glass and UO 2 HLLW mixing ratio, (c) waste loading in HLW glass, (d) HLW glass storage time, and (e) canister footprint. Table 2 shows the value of each parameter required to maintain the bentonite buffer temperature below 100 °C. These values were an 81% MA separation rate, 96% UO 2 HLLW mixing ratio, 2.9 wt% waste loading, and 901 year storage period. The bentonite buffer temperature exceeded 100 °C immediately after disposal, when the canister footprint expanded to 500 m 2 , which is more than 10 times the minimum canister footprint. Thus, the heat generated by MOX-HLW glass cannot be mitigated only by altering the canister footprint. Based on a recent study [10][11][12], an MA separation rate of 80% is reasonably achievable.
Mixing MOX-HLW glass with UO 2 HLLW is effective when the spent MOX fuel is less than 1/25 of the spent UO 2 fuel, thus requiring a large amount of UO 2 HLLW to dispose of the MOX-HLW glass. The waste loading limit of 2.9 wt% is lower than the conventional Japanese UO 2 reference case, such that the amount of MOX-HLW glass increases by a factor of 4.3. This is not recommended from the perspective of waste volume reduction. Lastly, storing MOX-HLW glass 901 years above the ground is unrealistic. Therefore, of the five processes evaluated in this study, MA

Waste loading of Am-241
To confirm the effect of Am-241 on the bentonite buffer temperature, heat transfer analysis was performed using MOX-HLW glass containing only Am-241. The associated heat generation was calculated using ORIGEN2.2-UPJ and ORLIBJ40. Figure 2 shows the resulting bentonite buffer temperatures at different waste loadings of Am-241. The maximum waste loading of Am-241 below the bentonite buffer temperature limit of 100 °C was 0.33 wt% (0.36 wt% as Am 2 O 3 ). Therefore, when considering geological disposal, the waste loading of Am-241 in MOX-HLW glass should be less than 0.33 wt%. However, this equates to the disposal of only 1.32 kg of Am-241 per MOX-HLW glass. Thus, the disposal of Am-241 isolated via MA separation is not feasible using conventional Japanese HLW glass disposal methods. This necessitates new ways of treating and disposing of Am-241.  Acknowledgments This study was conducted as part of a research program on vitrification technology for waste volume reduction and supported by the Japanese Ministry of Economy, Trade, and Industry (Grant Number JPJ010599).
Data availability All data generated or analyzed in this study are included in this published article and its supplementary information.

Conflict of interest
The authors declare that they have no known competing financial interests or personal relationships that could have influenced this study.
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