Fission gasses

Almost all of the radioactive gasses (H, C, Kr, Xe, Cl, and I) are released during voloxidation1 (if performed) and dissolution. They are carried in a relatively dry stream from the voloxidizer if present or in a wet stream containing significant nitrogen oxide species from the dissolver. These gaseous components can be selectively captured for immobilization. In the US, Cl and I must be captured for virtually all fuels and Kr must be captured for fuels cooled less than 30 years and 3H from fuels cooled less than -50 years. Figure 5.5 shows an example of the gaseous fission products evolved from various fuel recycling steps. Of primary interest are 3H, 1291,14C, and 85Kr.

Tritium removal from voloxidizer off-gas is performed using a desiccant (such as CaSOi) or a molecular sieve (such as Linde type 3A). The water content of air fed to the voloxidizer is controlled to obtain the desired decontamination factor in the tritiated water removal bed without signifi­cant increases in the tritium waste stream volume. Capture is performed near room temperature followed by release of the tritiated water at higher temperature. The captured (H, D,T)2O is then immobilized for decay storage and disposal. With a 12.3 year half-life, tritium immobilization does not require a robust waste form. Current process development activities assume that the tritium waste form is sufficiently low in long-lived radionuclides to qualify for near-surface disposal and the target waste form is generally considered to be a low-water cement. The other leading candidate is cemen­tation of the loaded sorbent.

Iodine-129 and 36Cl are significant dose contributors for nearly all reposi­tory environments because they are highly mobile, have long half-lives (15.7 x 106 years and 0.3 x 106 years, respectively), and are efficiently concentrated in the human body. Therefore, every reprocessing nation has strict toler­ances on the capture of 129I at a minimum. Various past studies have shown

Voloxidation is a potential process step employed primarily to remove 3H from the fuel meat prior to dissolution so that waste streams from all downstream solvent extraction processes are not 3 H contaminated. Tritium capture may not be necessary in all countries to the same level as required in the US (40CFR61 and 10CFR20), so processes with and without voloxida — tion will be considered.


(Mass basis: 1 MT initial heavy metal UNF;

55 GWD/MTIHM; 5 year cooling)

5.5 Schematic of off-gas treatment components from a typical UNF (per kg initial U after 55 GWd/MTHM and 5 years of cooling) (Jubin et al., 2009).

that 94-99% of iodine reports to the dissolver off-gas. A large fraction of the iodine in the off-gas was found to be associated with organic compounds (e. g., methyl iodide). A range of technologies have been employed to capture iodine from the plant off-gas streams including (IAEA, 1980):

• silver saddles (AgNO3 on ceramic substrate) ^ Hanford and Savannah River

• silver faujasite (AgX) ^ Sellafield

• silver mordenite (AgZ) ^ Hanford

• AgNO3 on silica (e. g., AC-6120) ^ WAK and Mayak

• silver on alumina ^ LaHague, Rokkasho

• carbon ^ Hanford

• wet caustic scrub (2 m NaOH) ^ La Hague, Tokai, Krasnoyarsk, Mol, and Sellafield

• IODOX (20+ m HNO3)

• mercurex (mercuric and nitric acids) ^ Dounreay and West Valley

• cadmium faujasite (CdX)

Advancements in materials science have allowed for the development of improved solid getter materials for iodine. Chief among them are silver — loaded aerogels (Strachan et al., 2010a; Matyas 2012); metal organic frame­works (Nenoff et al., 2011; Sava et al., 2011) and chalcogenide-based glass aerogels (chalcogels) (Strachan et al., 2010a). However, these materials are currently in the development phase and are not ready for full implementation.

Iodine waste form development and waste management are closely coupled to the separations technique employed. For example, at La Hague

in France and Sellafield in the UK, iodine is managed by ocean disposal (isotope dilution) which leads to the most appropriate capture method of caustic scrubbing. Other than ocean disposal, the immobilization/manage — ment of iodine is still a significant technical challenge faced by the industry in general. Several waste forms have been proposed and are being devel­oped for the disposal of radioiodine.

Silver-loaded adsorbers (AgZ, AgX, AC-6120, alumina, etc.), for example, can be encapsulated in cements (Toyohara et al., 2002; Scheele et al., 2002) or low melting metals (Vance et al., 2005) or glasses (Garino et al., 2011; Perera et al., 2004), or hot pressed into a durable waste form (JAEA, 2007). Scheele et al. found that adding CaI2 to the grout would significantly reduce the leaching rate of 1 29 I by isotopic dilution in the pour water solution (Scheele et al., 2002). However, for some repository design concepts, the presence of cement is a disadvantage because of the impact of alkaline cement leach solution on the corrosion of HLW glass and SNF. For example, the Yucca Mountain repository design specifically excluded cement wher­ever possible. The loaded AgI containing ceramics or glass can be hot — pressed into a final waste form (Sheppard et al., 2006 ).

Alternatively, the iodine can be eluted from the capture media and immo­bilized. Pure halide waste can be immobilized in:

• bismuth oxide-based ceramics (Krumhansl and Nenoff, 2011)

• sodalite-like minerals (Strachan and Babad, 1979; Winters, 1980; Naka — zawa et al., 2000 )

• apatite-like minerals (Uno et al., 2001, 2004)

• glass by low temperature vitrification (Sakuragi et al., 2008 ; Mukunoki et al., 2009 ).

Table 5.2 summarizes several potential iodine waste forms along with their loading and anticipated performance. To date, the authors are not aware of any of these processes being utilized on an industrial scale.

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