Tome Architecture and management of a geological repository - Andra

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6 – C waste repository zoneFigure 6.1.3Standard vitrified waste container (CSD-V)The AVM glass waste packages group all the vitrified waste produced since 1978 in the COGEMAvitrification shop at Marcoule. As indicated above, the vitrified solutions come in the main fromreprocess UNGG in the site's UP1 plant. Note that four different glass formulations exist for one ormore vitrification campaigns.The vitrified waste is conditioned in a stainless steel container presented in Figure 6.1.4Figure 6.1.4AVM vitrified waste containersThe average weight of each package is 410 kilograms, including 360 kilograms of vitrified waste. Thepackages' heat rating comes principally from medium-lived fission products, 90 Sr et 137 Cs. A heatrating of 155 watts by 2025 has been adopted for the study. Note that this heat rating is higher than forthe previous packages; it is nevertheless significantly less than those of the C packages presentedbelow. The equivalent β-γ dose rate in pseudo-contact (5 cm) with the package, around 235 Sv/h, willbe around 150 Sv/h by 2025.Dossier 2005 granite - ARCHITECTURE AND MANAGEMENT OF A GEOLOGICAL REPOSITORY162/228
6 – C waste repository zone6.1.2 Vitrified waste packages from current or planned productions in the short termThese packages contain solutions of fission products produced by reprocessing PWR UOX/enrichedrecycled uranium fuels in the COGEMA La Hague plants, conditioned as glass in a so-called CSD-Vstainless steel container (see Figure 6.1.3). It is assumed that production and conditioning of the wastewill occur after an average fuel storage time of eight years, after unloading from the reactors. Theconditioned waste weighs 400 kilograms on average per package.A first group of waste packages relates to current industrial productions from a heat transfer viewpoint.Following the hypotheses adopted, the vitrified waste is made up of a mixture of solutions of fissionproducts from UOX1 (average combusion rate of 33 GWj/t), UOX2/enriched recycled uranium(average combustion rate 45 GWj/t) and UOX3 (average combustion rate 55 GWj/t) fuels.A second group of waste packages relates to packages with a slightly-increased heat rating. Thevitrified waste is made up of a mixture of solutions of fission products from UOX2/enriched recycleduranium and UOX3 fuels, with, as previously, average combustion rates of 45 GWj/t and 55 GWj/trespectively.The heat ratings of these packages are illustrated in Figure 7.1.1 (blue and red curves). They raise theissue of storage time before being placed in a repository and of repository module design, to limit thetemperature to acceptable levelsThe equivalent β-γ dose rate on pseudo contact (5 cm) with the packages is about 240 Sv/hr after 60years’ cooling.6.1.3 Future hypothetical packagesThese waste packages relate to potential glass productions on the COGEMA La Hague site. Thesepackages have been defined in the scenarios adopted for the study on the assumption that productionand conditioning of the waste will take place, as for the previous glasses, after an average fuel storageperiod of eight years, after unloading from the reactors. Note that other possibilities could beenvisaged.Here also, the conditioned waste weighs 400 kilograms on average per package. The containers aresimilar to the ones presented above (CSD-V).A first waste package group describes glasses produced by conditioning solutions of fission productsfrom UOX and MOX fuels. Their make-up is defined as a mixture of 15% MOX and 85% UOX2.A second waste package group describes vitrified waste from reprocessing UOX fuels, which containsa small additional plutonium load. The plutonium incorporation rate in the glass is fixed specifically atone percent, about 4 kilograms per package. The incorporated plutonium comes from the UOX2 fuels.The heat ratings produced by these packages are illustrated above in Figure 3.2.15 (green and orangecurves).The equivalent β-γ dose rates in pseudo-contact (5 cm) with the packages will be 235 Sv/h for theUOX/MOX glasses and 265 Sv/h for the glasses with plutonium respectively, after a sixty-yearcooling period.6.1.4 Production scenariosTable 6.1.2. Table 6.1.2 gives the numbers and volumes of C waste primary packages arising from thestudy scenarios presented in Table 6.1.2.Dossier 2005 granite - ARCHITECTURE AND MANAGEMENT OF A GEOLOGICAL REPOSITORY163/228
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Contents4.1 Surface installations..
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Contents7.2.1 Harnessing the favour
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ContentsFigure 5.1.7 Non-alloy stee
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ContentsFigure 7.6.3 CU package emp
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11Study approach1.1 The main stages
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1 - Study approachSome studies were
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1 - Study approachIn France spent f
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2 - High-Level and Long-Lived waste
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3 - Design study of a repository in
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44General architecture of thereposi
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5 - B waste repository zoneThis cha
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5 - B waste repository zoneproduced
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5 - B waste repository zoneFigure 5
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5 - B waste repository zonerepaired
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5 - B waste repository zoneFigure 5
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5 - B waste repository zone5.1.4 Ce
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5 - B waste repository zonein the o
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5 - B waste repository zoneThe pack
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7 - Spent fuel repository zone7.4.2
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7 - Spent fuel repository zone7.6 D
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7 - Spent fuel repository zoneevent
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88Conclusions8.1 Generic architectu
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8 - ConclusionsThe envisaged two-le
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Références bibliographiquesRefere
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Références bibliographiques[39] A
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Photo Credits:ANDRA - SKB - NAGRA -
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Tome Architecture and management of a geological repository - Andra

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