Spent fuel

Radioactive decay and decay heat

As noted in Section 4.5.2, individual canisters of SF and HLW will have various heat outputs at the time of emplacement in the repository. A thermal constraint of a maximum of 1500 W per canister at the time of waste emplacement has been selected, based on calculations of repository temperature evolution (see Section 4.5.2). This can be met for PWR and BWR UO2 fuel canisters by allowing 40 years storage prior to emplacement for fuel with an average burnup of 48 GWd/t_IHM. In the case of canisters with 3 UO₂ assemblies and 1 MOX assembly of the same burnup, 55 years cooling are required for the canister decay heat to decrease to ~ 1500 W. The time-dependent decay heat of canisters of SF and HLW is illustrated in Fig. 5.3-1.

Fig. 5.3-2 shows the total α and total β/γ activities for a canister containing 1 MOX assembly and 3 UO₂ assemblies and for a HLW canister with BNFL glass. For canisters with 4 PWR or 9 BWR UO₂ assemblies, the β/γ activity is similar, but the α activity and decay heat are

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approximately 30 % lower than for the 1 MOX / 3 UO₂ canister in the time period from 40 to 10⁵ years (McGinnes 2002).

Fig. 5.3-1: Calculated time-dependent heat production of disposal canisters as a function of time from reactor discharge for: a SF canister with 3 UO₂ assemblies and 1 MOX assembly (1.50 t_IHM), a SF canister with 9 BWR UO₂ assemblies (1.60 t_IHM), all with an average burnup of 48 GWd/t_IHM and a canister with an average flask of BNFL HLW glass (based on McGinnes 2002)

Fig. 5.3-2: Calculated time-dependent total α and total β/γ activity of i) a SF canister with 3 UO2 assemblies and 1 MOX assembly (1.50 tIHM), all with an average burnup of 48 GWd/t_IHM, and ii) an average flask of BNFL HLW glass (based on McGinnes 2002)

Decayheat[W/canister] 1000

2000

Time from reactor discharge [a]

40 100 1 000 10 000

500

0

SF (3 UO + 1 MOX)2

SF (9 UO )2

1500 HLW

Activity[Bq/canister]

10¹²

Time from reactor discharge [a]

10³ 10¹¹

10¹³ 10¹⁴ 10¹⁵ 10¹⁶

10² 10⁴ 10⁵ 10⁶

HLW total activity

SF (3 UO + 1 MOX) total2 activity SF (3 UO + 1 MOX) total activity2

HLW totalactivity

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Nuclear criticality

Nuclear criticality is a sustained nuclear chain reaction, in which fissioning atoms release neutrons which induce further fission reactions. The reactions produce radiation, heat and new fission products. In order to occur, criticality requires sufficient quantities of fissile isotopes in a suitable geometry and a neutron moderator, such as water, to slow the neutrons. The principal fissile isotopes present in spent fuel are ²³⁵U, ²³⁹Pu and ²⁴¹Pu, which are present at levels that are sufficient to obtain criticality under certain conditions. Criticality, were it to occur, could pro-duce elevated temperatures (several hundred degrees) in the EBS, which could potentially damage canisters and backfill material and induce groundwater movement. It is necessary, therefore, to select a layout and adopt other measures that ensure sub-criticality and to perform calculations to demonstrate that, for the spent fuel compositions and geometries in question, including any evolution of the EBS, criticality will not occur.

Calculations for MOX and UO₂ fuel of average burnup indicate that the reactivity⁷⁶ decreases for the first few hundred years after discharge of the fuel from the reactor due to ingrowth of

241Am, a strong neutron absorber, from decay of ²⁴¹Pu. It gradually increases back to its initial value over hundreds of thousands of years as ²³⁵U grows back in. This means that the potential for criticality requires evaluation for the entire period for which assessment calculations are performed. An analysis summarised in Kühl et al. (2003) shows that, when loaded with spent UO₂ fuel, canisters would be sub-critical, both when intact, and after failure when the void space in the canister would eventually become filled with water, provided a burnup of at least 15 GWd/t_IHM is reached (22 GWd/t_IHM for canisters with both MOX and UO₂ fuel)⁷⁷. In the rare cases in which the minimum burnup is not attained by a fuel assembly, the co-placement of high burnup assemblies in the canister or the use of inert filler (e.g. sand) in the void space in the canister would reduce reactivity to a sub-critical level. For the long term, calculations show that changes in geometry, such as diffusion of large quantities of uranium into the bentonite, followed by precipitation within pore spaces, would not result in criticality. Similar conclusions have been reached by Oversby (1996), whose results are based on observations from the Oklo natural reactor.

The identification of a minimum burnup requirement to preclude criticality represents a waste acceptance criterion related to the operation of a spent fuel encapsulation facility, an issue which is discussed in more detail in Section 4.5.2.4.

Radiation damage

Uraninite is regarded to be highly resistant to radiation damage, as evidenced by studies of natural uraninites. Although spent fuel will sustain an even higher radiation dose over hundreds of thousands of years, this will not lead to loss of crystallinity, as discussed by Weber (1981) and Johnson & Shoesmith (1988).

5.3.1.2 HLW

The time-dependent heat production of a canister of BNFL HLW glass after 40 years of cooling is shown in Fig. 5.3-1. The α- and β/γ- activity of BNFL glass is shown in Fig. 5.3-2. As noted

76 Reactivity is a measure of the number of excess neutrons produced by fission

77 The use of burnup credit in criticality calculations is reviewed by Dyck (2001). It is recognised that considerable developments are required before it is applied in the context of spent fuel disposal. The principles as well as the technologies to implement it are nonetheless well established.

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in Section 4.5.2.2, the COGEMA glass has a slightly lower activity and decay heat. More detailed data are available in McGinnes (2002).

Radiation damage

Accumulated radiation damage effects appear to cause only small increases (a factor of 2 to 3) in the dissolution rate of HLW glass (Lutze 1988). These small effects are considered in the selection of the reference glass dissolution rates discussed in Section 5.3.4.6.