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Waste sources and classification

29

2.7.2.1. Radiological data

Individual radionuclide and total activities A summary of some of the methods used to determine radionuclide inventories is given in Table 2.5.

This information is required not only as input for a repository safety assessment, but is also used for dose rate, heat output and radiotoxicity calculations.

The number of radionuclides presented in a radionuclide inventory is programmespecific, e.g., in the Swiss programme, for model inventories all radionuclides of halflife greater than 60 days plus important short-lived daughters are initially considered (i.e., a total of 147 radionuclides). Following a simplified safety analysis (screening exercise), this list is reduced to those radionuclides that are safety-relevant and only for this inventory is the full safety assessment study performed.

As an example, the nuclides selected after the simplified safety assessment, those selected after the analysis of the near-field and, finally, those that dominated releases from the far-field in the case of the studies performed for a potential Swiss L/ILW repository (Nagra, 1993) are given in Table 2.6.

Heat output Heat outputs are determined using WBq 1 conversion factors based on the determined radionuclide inventory. For L/ILW, heat output is not normally of great interest (apart from ensuring that it is not too high for ILW-LL). However, for HLW wastes, this parameter determines the size of the repository, in that disposal canister heat output limits are required to ensure that the engineered barrier system (EBS) functions as required. These limits are a compromise between the spacing between waste packages and the dimensions and spacing of the tunnels (see also chapters 3 and 5).

Dose rates If no production data exist, these can be calculated with varying degrees of sophistication. A simple, user-friendly code is Microshield (Grove, 1998) but, like all tools, the limitations must be understood if it is to be used for more than simple modelling. If more accurate values are required, and especially if complicated

Table 2.5

Overview of methods used to determine radionuclide inventories

Waste category

Radionuclide inventory determination

 

 

Reactor operational

Correlation factors (measurement of samples of raw waste to obtain a radionuclide

wastes

spectrum. As the relationship between the nuclides is relatively stable, this

 

spectrum can then be correlated to a few key measured nuclides or dose rates).

 

Reactor modelling. In Belgium, e.g., reactor parameters (cooling water chemistry,

 

etc.) have been exhaustively correlated to the wastes produced and hence

 

inventories can be determine in this manner.

Reactor core

Modified fuel depletion code validated by a few measurements.

components

 

Reactor bioshield

Neutron transport code supplemented by a few measurements.

Reprocessing waste

Fuel depletion codes, information from the reprocesser (BNFL, Cogema).

Spent fuel

Fuel depletion codes.

Accelerator wastes

Spallation, medium energy reaction codes supplemented by a few measurements.

Medicine, industry and

Production declarations, government statistics concerning import and/or production

research

of sources.

 

 

30

D.F. McGinnes

Table 2.6

Summary of nuclides in the initial (reference) inventories and at subsequent stages of the safety assessment for a L/ILW repository in Switzerland.

Nuclides

Nuclides

Nuclides

Nuclides

Nuclides

Nuclides

remaining

selected after

dominating

remaining

selected after

dominating

after initial

the analysis

releases from

after initial

the analysis

releases from

pre-screening

of near-field

the far-field

pre-screening

of near-field

the far-field

exercise

releases

 

exercise

releases

 

 

 

 

 

 

 

H-3

 

 

Be-10

 

 

C-14

C-14

C-14

Na-22

 

 

Cl-36

Cl-36

Cl-36

Ar-42

 

 

K-40

K-40

K-40

Ca-41

 

 

Fe-55

 

 

Fe-60

 

 

Co-60

 

 

Ni-59

Ni-59

 

Ni-63

Ni-63

 

Se-79

Se-79

Se-79

Sr-90

Sr-90

 

Zr-93

 

 

Nb-94

 

 

Mo-93

Mo-93

Mo-93

Tc-99

 

 

Ru-106

 

 

Pd-107

 

 

Ag-108M

Ag-108M

 

Sn-126

 

 

I-129

I-129

 

Cs-134

 

 

Cs-135

 

 

Cs-137

Cs-137

 

Sm-151

 

 

Eu-152

 

 

Ho-166M

 

 

Pb-210

 

 

Po-210

 

 

Ra-226

Ra-226

Ra-226

Ra-228

Ra-228

 

Ac-227

Ac-227

 

Th-228

Th-228

 

Th-229

Th-229

Th-229

Th-230

Th-230

 

Th-232

Th-232

 

Pa-231

Pa-231

Pa-231

U-232

U-232

 

U-233

U-233

U-233

U-234

U-234

 

U-235

U-235

 

U-236

U-236

 

U-238

U-238

 

Np-237

Np-237

Np-237

Pu-238

Pu-238

 

Pu-239

Pu-239

Pu-239

Pu-240

Pu-240

Pu-240

Pu-241

Pu-241

 

Pu-242

Pu-242

 

Am-241

Am-241

Am-241

Am-242M

Am-242M

 

Am-243

Am-243

 

Cm-243

Cm-243

 

Cm-244

Cm-244

 

Cm-245

Cm-245

 

Cm-246

Cm-246

 

geometries are involved, more sophisticated codes such as RANKERN (AEA, 2003) or MCNP (LANL, 2005) should be considered.

Surface contamination If no production data exist, these are normally set at the handling limits that are approved by the competent regulator.

Radiolytic gas production This can be measured but, more often, is based on conservative calculations of the theoretical gas production rate using so-called G-values. These G-values are experimentally determined for different types of materials, e.g., concrete, PVC, paper, rubber, etc., and describe the amount of gas produced from the deposition of a discrete amount of radiation in this material. Generally, these values are determined for - and -radiation, with having proportionally higher values due to the