Importance for Long-term Safety Analysis

The reduction of the radiotoxicity of the nuclear waste is often used to argue in favor of a P&T fuel cycle. Figure 1.1 showed that the radiotoxicity of untreated nuclear waste declines very slowly. It takes geological time scales to reach a level compared to that of a uranium ore. If the minor actinides are transmuted into stable isotopes or isotopes with a much shorter half-life, the time scale is significantly reduced. The value of transmutation of minor actinides appears obvious. But this figure mirrors a simplified case: the used definition of the radiotoxicity assumes that the total waste inventory is consumed by the population. This can be seen as a very conservative

estimation since ingestion of radionuclides yields the highest dose rates. Yet, it neglects the probability of such an ingestion. Consequently, it is usually not used as a reference value when assessing a deep geological repository in regard to the risk it would pose to future generations. If the potential exposure to the public by such a deep geological repository is to be calculated, there are many more parameters that need to be taken into account (Brasser et al. 2008). In contrast to the intrusion or repository-breach scenario, where the whole radionuclide inventory is ingested, the underlying scenario for the long term safety assessment is a leach-and-migrate scenario. In this case, the repository functions more or less as planned and is not compromised from the outside. Regardless, after a certain time span, the man-made barriers will fail. Some of the nuclides will eventually dissolve from the host matrix and will be transported to the surface where they might affect the population.

This relevant safety scenario goes beyond the sole composition of the radioactive waste in the geological repository which is used for the radiotoxicity calculation. Instead, the repository system as a whole has to be analyzed, because different nuclides behave differently in the host matrix. Possible questions are: how long will the host matrix stay intact and prevent radionuclides from diffusion and subsequent release into the biosphere? Which radionuclides are more likely to dissolve? Which release paths are dominating? The answers to these questions also depend on the geological formations, which are site-specific and might vary over time. Beside ongoing changes such as erosion and denudation processes, sudden events, such as seismic activity or volcanic eruptions, can affect the repository. The long term safety analysis of a deep geological repository tries to cover all these aspects as best as possible. Among others, it estimates the radioactive dose humans will receive in the future from the repository. Unlike the radiotoxicity index, where the value is compared to the value for an unspecified natural uranium ore, the estimated dose rate originating from the future repository is compared to what is considered to be an acceptable exposure today. This also explains why they result in very low values compared to the calculated ingestion dose rates.

The approach explained above is by far more complex than simply using the total radioactive inventory and multiplying it by the appropriate conversion factors to calculate the ingestion dose rates. It starts with the fact, that for example the host matrix in which the deep geological repository is placed must be considered. Granite, clay, salt, or tuff affect the mobility of certain nuclides differently.

But in all cases, the dose rate emitted from the deep geological repository is dominated by some long-lived fission products when looking at time scales beyond 10,000 years (IAEA 2004, p. 98). In Mol, Belgium, the clay dome repository SAFIR-2 is planned. Exemplary calculations show that, in the long term, among the most influential nuclides are selenium-79, technetium-99, tin-126, and iodine-129 (Preter et al. 2001; G. Schmidt et al. 2013). For the planned swiss deep geological repository, also to be placed in clay, selenium-79 and iodine-129 are calculated to be the most relevant isotopes after 100,000 years (Nagra 2002, p. 261). For the long-term safety of repositories in other host matrices, zirconium-93 and cesium-135 play an important role as well (IAEA 2004, p. 98). zirconium-93 is in so far interesting as it is produced by neutron irradiation of zirconium alloys which are often used as cladding material for the fuel pins.

Assuming that an efficient transmutation fuel cycle can be implemented, and that uranium, plu- tonium and the minor actinides are almost completely removed from the spent fuel, the main contributors to the radiotoxicity will be the fission products. Figure 7.12 shows that in this case, after 1000 years of decay, the radiotoxicity is dominated by only a few fission products for typical high level waste made from PWR spent fuel. In a time frame of up to 100,000 years, the dose rate is dominated by the technetium-99 inventory. Afterwards, iodine-129 dominates. This mirrors the fact that even the long-lived fission products greatly differ in regard to their lifetime: while technetium-99 has a half-life of 211,100 years, the half-lives of iodine-129 and cesium-135 are more than ten times longer. Table 7.15 summarizes these values.

Contr

ibution to R

adioto

xicity in %

Time After Discharge in Years 0 20 40 60 80 103 104 105 106 Tin-126 Technetium-99 _Iodine-129 Cesium-135 Zirconium-93

Figure 7.12.: Relative contribution of selected long-lived fission products to the total radiotoxicity if all transuranium elements are removed from the spent fuel (Kloosterman et al. 1995). The five fission products are responsible for more than 90 % of the total radiotoxicity from all fission products.

Considering the discussion above, it seems only consequent that in the original partitioning and transmutation concepts, the destruction of the long-lived fission products was included as well (NRC, Committee on Separations Technology and Transmutation Systems 1996; Jameson et al. 1992; United States Department of Energy 1999). They investigated also the transmutation of at least technetium-99 and iodine-129. When assessing the practical feasibility of the transmutation of fission products, a first step is the calculation of the transmutation half-life (NEA 2005). The transmutation half-life is a rough estimate for the time period it takes to incinerate half of the initial mass. It is defined as

T₁t r ans_/2 = ln2

σn,γ· Φ · 3.15 · 107

years. (7.12)

The transmutation half-life depends on the cross-section for (n,γ) reactions σ_n,γand the neutron

fluxΦ. Because it neglects the ongoing build-up of fission products in the core, this definition

is mostly useful when looking at heterogeneous transmutation. In this case, special elements rich on the targeted nuclides are positioned in the core. Table 7.15 gives average values for the transmutation half-life in a thermal and a fast neutrons spectrum (NEA 2002, p. 265). The energy of the incoming neutrons is assumed to be 1 eV and 0.2 MeV for the two spectra. A neutron flux of 1014n/(s·cm2_{) and10}15_n/(s·cm2_{), respectively, is selected.}

Explicit transmutation in a nuclear reactor is only useful when the transmutation half-life is significantly shorter than the natural half-life of the targeted radionuclide. For the long-lived fission products, at least in principle this holds true for the thermal and the fast neutron spectrum. Yet, the

Table 7.15.: Characteristic values for selected long-lived fission products. The half-lives are taken from the Janis database NEA (2012b). Average isotope production from a pressurized water reactor with a burn-up of 50 MWd/kgHM is tabled in NEA (1999, p. 170). The ingestion dose and the transmutation half-lives as listed in NEA (2002, p. 265) are shown for a simplified thermal and fast neutrons spectrum. Note that a 1 GWe nuclear reactor can produce up to 8 TWh per year.

Isotope Half-life Isotope Quantity Ingestion Dose T₁t r ans_/2 ,ther mal T₁t r ans_/2 , f ast

1000 years kg/TWh nSv/Bq years years

Se-79 335 0.018 2.3 220 730 Zr-93 1,610 2.8 0.42 790 730 Tc-99 211 3.2 0.34 51 110 Sn-126 230 0.079 5.1 4400 4400 I-129 15,700 0.66 74 51 160 Cs-135 2,300 1.40 1.9 170 310

thermal spectrum leads to a higher transmutation efficiency, in particular for technetium-99 and iodine-129. Therefore this already contradicts the objective of efficient minor actinide transmutation in a fast neutron spectrum.

The difference between the fast and thermal transmutation half-lives of the isotopes seems small. But when considering human time scales, there is a significant difference between 51 years and 160 years for the example of iodine-129. It takes long irradiation periods for efficient reduction in the overall inventory, because the neutron absorption cross sections and thus the probability of the absorption reaction is small. However, once again, material endurance is an open question. It is impossible that cladding and structure last for such a long time, considering the heavy neutron bombardment.

One limiting factor especially in a thermal neutron spectrum is the neutron balance. Each trans- mutation of a fission product is a net neutron loss due to absorption since no new neutrons are produced. Even in a fast spectrum it might be necessary to add external neutrons when the transmutation of long-lived fission products is an actual objective during operation.

Additional, element specific challenges when transmuting long-lived fission products became evident over the years. One example is the long-lived cesium-135 isotope. It comprises only about 10 % of the total cesium vector in the spent fuel. Other isotopes, such as cesium-133 and cesium-134, are also present. These isotopes also absorb neutrons, producing on the one hand more cesium-135 isotopes while on the other hand parasitically using neutrons that are supposed to be used for transmutation of cesium-135. Possible transmutation schemes for cesium-135 were developed, but the general opinion is that cesium-135 transmutation is too complicated to be deployed on an industrial scale (NEA 2005, p. 215). The transmutation of technetium-99 and iodine-129 seems more feasible, but there were only limited irradiation experiments with different target materials and geometries. Iodine targets for example bear the problem that a considerable amount of xenon gas is produced during irradiation. The gas needs to be removed by venting of the target (NEA 1999, p. 171).

The emphasis has mostly shifted to the sole treatment of the minor actinides in the spent fuel because of these challenges. Therefore, current publications only discuss the incineration of minor actinides when assessing the transmutation efficiency for a certain system (Mansani et al. 2012; Mueller 2013; Renn 2013; Sarotto et al. 2013). All arguments in favor of transmutation only consider the hazard generated by a deep geological repository. However, it is acknowledged even by advocates of a P&T fuel cycle, that a deep geological repository is required.

For the design and capacity of a deep geological repository, the emitted dose rate is not the limiting factor. The initial heat generation of the spent fuel and, subsequently, of the high-level waste, is far more important in this context (Bonin 2010, p. 3307). The residual heat of typical LWR spent fuel is dominated by the more short-lived fission products, such as cesium-137 and strontium-90 (Mueller and Abderrahim 2010). Both isotopes have a half-life of about 30 years. For the P&T-fuel, the situation is different since especially curium-242 adds a significant amount of heat during this time period. While much could be gained with an extended interim storage, this would contradict the objective not to shift the burden of nuclear waste treatment to next generations. The long-term consequences are also the topic of the following section, in which findings on long-term effects caused by the generation of long-lived fission products in a P&T fuel cycle are presented.