Radionuclide Release and Transport RNT-2008

POSIVA 2008-06 Radionuclide Release and Transport RNT-2008 Mikko Nykyri Henrik Nordman Nuria Marcos Jari Löfman Antti Poteri Aimo Hautojärvi December 2008 POSIVA OY Olkiluoto FIN-27160 EURAJOKI, FINLAND Phone (02) 8372 31 (nat.), (+358-2-) 8372 31 (int.) Fax (02) 8372 3709 (nat.), (+358-2-) 8372 3709 (int.)

POSIVA 2008-06 Radionuclide Release and Transport RNT-2008 Mikko Nykyri Safram Oy Henrik Nordman, Jari Löfman, Antti Poteri VTT Nuria Marcos Saanio & Riekkola Oy Aimo Hautojärvi Posiva Oy December 2008 Base maps: National Land Survey, permission 41/MML/08 POSIVA OY Olkiluoto FI-27160 EURAJOKI, FINLAND Phone (02) 8372 31 (nat.), (+358-2-) 8372 31 (int.) Fax (02) 8372 3709 (nat.), (+358-2-) 8372 3709 (int.)

ISBN 978-951-652-166-3 ISSN 1239-3096

Posiva-raportti Posiva Report Posiva Oy Olkiluoto FI-27160 EURAJOKI, FINLAND Puh. 02-8372 (31) Int. Tel. +358 2 8372 (31) Raportin tunnus Report code POSIVA 2008-06 Julkaisuaika Date December 2008 Tekijä(t) Author(s) Mikko Nykyri, Safram Oy Henrik Nordman, VTT Nuria Marcos, Saanio & Riekkola Oy Jari Löfman, VTT Antti Poteri, VTT Aimo Hautojärvi, Posiva Oy Nimeke Title Toimeksiantaja(t) Commissioned by Posiva Oy RADIONUCLIDE RELEASE AND TRANSPORT RNT-2008 Tiivistelmä Abstract The Finnish nuclear waste disposal company, Posiva Oy, is planning an underground repository for spent nuclear fuel to be constructed on the island of Olkiluoto on the south-west coast of Finland. The repository design is based on the KBS-3V concept, where spent fuel elements in copper-iron canisters are emplaced in vertical holes in the bedrock and surrounded by bentonite clay. This report presents the radionuclide release and transport analysis RNT-2008 that forms a part of Posiva's safety case. The analysis covers the subject from the release of radionuclides from spent nuclear fuel to their arrival in the biosphere. RNT-2008 is a deterministic analysis. Its assumptions are purposely conservative, meaning that they shall ensure that the results, with high degree of certainty, overestimate radioactive releases and radiation exposures. The calculation cases are composed from several scenarios, the main assumptions being that 1) a disposal canister will have initially a penetrating hole in its wall or corrode at a high rate to form one, 2) the bentonite buffer will be partly missing its function as a barrier, 3) rock shear movements will damage the near field barriers, 4) groundwater flow rates will be higher than the flow analysis suggest, 5) groundwater chemistry will be less favourable than predicted, and 6) gas generation will enhance the transport of radionuclides. The main quantitative results are expressed as indicative dose rates from the use of a stylised drinking-water well and as activity release rates from the geosphere to the biosphere. The results suggest that the assessed repository system well complies with the regulatory criteria. Based on the WELL-2008 dose rates, the most significant radionuclides in most cases are I-129, C-14, Cs-135, Pa-231 and Cl-36, in this order. The results indicate that the engineered barriers together with the sparsely fractured host rock around the near field dominate the capacity of the repository to retain the radionuclides and retard their migration. The central role of the bedrock farther from the canisters is to provide stable and favourable chemical and physical conditions for the engineered barrier system, and to hinder inadvertent human intrusion into the repository. Avainsanat - Keywords Safety case, safety assessment, long-term safety, spent nuclear fuel, nuclear waste, crystalline bedrock, KBS-3 ISBN ISSN ISBN 978-951-652-166-3 ISSN 1239-3096 Sivumäärä Number of pages Kieli Language

Posiva-raportti Posiva Report Posiva Oy Olkiluoto FI-27160 EURAJOKI, FINLAND Puh. 02-8372 (31) Int. Tel. +358 2 8372 (31) Raportin tunnus Report code POSIVA 2008-06 Julkaisuaika Date Joulukuu 2008 Tekijä(t) Author(s) Mikko Nykyri, Safram Oy Henrik Nordman, VTT Nuria Marcos, Saanio & Riekkola Oy Jari Löfman, VTT Antti Poteri, VTT Aimo Hautojärvi, Posiva Oy Nimeke Title Toimeksiantaja(t) Commissioned by Posiva Oy RADIONUKLIDIEN VAPAUTUMINEN JA KULKEUTUMINEN RNT-2008 Tiivistelmä Abstract Posiva Oy suunnittelee loppusijoituslaitoksen rakentamista käytetylle ydinpolttoaineelle Eurajoen Olkiluodon kallioperään. Laitossuunnitelma perustuu KBS-3V -konseptiin, jossa käytetty polttoaine tulee olemaan kupari-valurautakapseleissa bentoniittisaven ympäröimänä kallioon tehdyissä pystysuorissa rei'issä. Tämä raportti esittelee radionuklidien vapautumis- ja kulkeutumisanalyysin RNT-2008, joka muodostaa osan Posivan turvallisuusperusteluista. Analyysi kattaa aiheen alkaen radionuklidien vapautumisesta kapselista ja päättyen siihen, kun radioaktiiviset aineet saavuttavat biosfäärin. RNT-2008 on deterministinen analyysi. Sen oletukset ovat tarkoituksella konservatiivisia, mikä tarkoittaa että oletuksilla varmistetaan tulosten yliarvioivan radioaktiivisia päästöjä ja säteilyvaikutuksia elinympäristössä suurella varmuudella. Laskentatapaukset on muodostettu useasta skenaariosta, joiden tärkeimpiä oletuksia ovat seuraavat: 1) loppusijoituskapselissa on alunperin seinämän läpi ulottuva reikä tai korroosio puhkaisee sen nopeasti, 2) osa bentoniittisaven estevaikutuksesta puuttuu, 3) kallion lohkoliikunnat vioittavat vapautumisesteitä, 4) pohjaveden virtaamat ylittävät virtausanalyysin antamat tulokset, 5) pohjavesikemia on arvioitua epäedullisempi ja 6) kaasunkehitys nopeuttaa radionuklidien kulkeutumista. Tärkeimmät kvantitatiiviset tulokset ilmoitetaan sekä juomavesikaivon käytöstä aiheutuvina viitteellisinä säteilyannosnopeuksina että radionuklidien vapautumisnopeuksina elinympäristöön. Suunniteltu loppusijoituskonsepti täyttää saatujen tulosten mukaan viranomaisvaatimukset. Merkittävimmät radionuklidit ovat I-129, C-14, Cs-135, Pa-231 ja Cl-36, annetussa järjestyksessä. Tulokset viittaavat siihen, että tekniset vapautumisesteet yhdessä lähikallion kanssa vaikuttavat eniten laitoksen kykyyn pidättää radionuklideita ja viivästää niiden kulkeutumista. Kallioperän keskeinen rooli on pysyvien ja edullisten olosuhteiden luominen teknisille vapautumisesteille ja ihmisen tahattoman tunkeutumisen estäminen. Avainsanat Keywords Turvallisuustodisteet, turvallisuusanalyysi, pitkäaikaisturvallisuus, käytetty ydinpolttoaine, loppusijoitus, kiteinen kallioperä, KBS-3 ISBN Sivumäärä Number of pages 164 978-951-652-166-3 ISSN Kieli Language ISSN 1239-3096 Englanti

1 TABLE OF CONTENTS ABSTRACT TIIVISTELMÄ TERMS AND ABBREVIATIONS... 3 FOREWORD... 5 1 INTRODUCTION... 7 2 SAFETY CRITERIA... 11 3 DISPOSAL CONCEPT... 15 3.1 KBS-3V concept... 15 3.2 Olkiluoto site and repository layout... 17 4 GROUNDWATER FLOW AND SOLUTE TRANSPORT... 19 4.1 Equivalent porous medium (EPM) modelling of groundwater flow... 19 4.2 Discrete fracture network modelling (DFN)... 35 4.3 Consistency of the EPM and DFN results... 56 5 RADIONUCLIDE RELEASE AND TRANSPORT... 63 5.1 Assessment scenarios and modelling concepts... 63 5.2 Transport processes... 68 5.3 Mathematical formulations... 75 6 CALCULATION CASES... 79 6.1 General assumptions... 79 6.2 Data... 90 7 RESULTS... 107 7.1 General... 107 7.2 Defective canister calculation cases... 107 7.3 Additional calculation cases... 122 7.4 Supplementary cases behaviour of system... 124 8 OVERVIEW OF RESULTS AND DISCUSSION... 129 8.1 General... 129 8.2 The most important factors affecting the results... 130 8.3 Uncertainties... 133 8.4 Regulatory compliance... 134 8.5 Comparison to past safety assessments and KBS-3H... 136 9 SUMMARY... 141 REFERENCES... 145 APPENDIX 1: ILLUSTRATION OF GEOSPHERE RETENTION... 151 APPENDIX 2: DESCRIPTION OF CODES... 155 APPENDIX 3: LIST OF CALCULATION CASES... 163

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3 TERMS AND ABBREVIATIONS AP after present BWR boiling water reactor DFN discrete fracture network DP dual-porosity EDZ excavation damaged zone EdZ excavation disturbed zone EPM equivalent porous medium EPR European Pressurised (water) Reactor HZ hydrogeological zones IRF instant release fraction: the amount of radionuclides, located in any fuel microstructure and for which no long-term confinement properties can be anticipated LHZ local hydrogeological zones LO1 Loviisa nuclear power plant unit 1 (operating) LO2 Loviisa nuclear power plant unit 2 (operating) MHZ major hydrogeological zones NPP nuclear power plant OL1 Olkiluoto nuclear power plant unit 1 (operating) OL2 Olkiluoto nuclear power plant unit 2 (operating) OL3 Olkiluoto nuclear power plant unit 3 (under construction) OL4 Olkiluoto nuclear power plant unit 4 (planned) ONKALO underground site characterization facility at Olkiluoto PSAR preliminary safety analysis report PWR pressurized water reactor RNT radionuclide transport RNT-2008 radionuclide release and transport analysis, finished in the year 2008 SFR sparsely fractured rock SKB Svensk Kärnbränslehantering AB STUK Radiation and Nuclear Safety Authority TDS total dissolved solids TVO Teollisuuden Voima Oy VTT VTT Technical Research Centre of Finland VVER-440 pressurised water reactor of Soviet design; nominal capacity 440 MWe WCA well characterised area WCF water-conducting background fractures

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5 FOREWORD This study was supervised by Aimo Hautojärvi of Posiva Oy and coordinated by Mikko Nykyri of Safram Oy on behalf of Posiva. The RNT-2008 working group consisted of Henrik Nordman (VTT), Nuria Marcos (Saanio & Riekkola Oy), Jari Löfman (VTT), Antti Poteri (VTT), Aimo Hautojärvi (Posiva), Ari Ikonen (Posiva) and Mikko Nykyri (Safram). The present report was written by Henrik Nordman, Nuria Marcos, Jari Löfman, Antti Poteri, Aimo Hautojärvi and Mikko Nykyri. The radionuclide transport calculations were carried out by Henrik Nordman. The groundwater flow analysis was carried out by Jari Löfman and the solute transport analysis by Antti Poteri. The report was reviewed as an early draft by the following individual experts: Jan-Olof Selroos (Swedish Nuclear Fuel and Waste Management Co), Paul Smith (Safety Assessment Management Ltd.) and Michael Thorne (Mike Thorne and Associates Limited).

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7 1 INTRODUCTION The Finnish nuclear waste disposal company, Posiva Oy, is preparing the actualization of the plans for an underground repository for spent nuclear fuel. The repository will be constructed at the island of Olkiluoto on the south-west coast of Finland, at the site where there are two operating power reactors, one under construction, and one under consideration. The repository design is based on a multiple-barrier concept referred to as KBS-3V, where spent fuel elements in copper canisters are emplaced in vertical holes in the bedrock (one canister per hole) and surrounded by a buffer of bentonite clay. The original KBS-3 concept was introduced in the report SKBF/KBS (1983). The previous Finnish performance assessment for the KBS-3V concept, TILA-99 (Vieno & Nordman, 1999), was produced in support to the siting decision between the four candidate sites. Based on TILA-99 and other studies, Posiva promoted the selection of Olkiluoto for the repository site. The selection was favoured by the Finnish Government's Decision in Principle in the year 2000 and became ratified by the Parliament in May 2001. The on-going programme stage is targeted for the preparation of the construction licence application and for showing that the plans to construct and operate the repository are feasible and that the long-term safety targets will be met. This includes site-confirming investigations at Olkiluoto and the construction of ONKALO, an underground rock characterization facility that will be a part of the repository (Saanio et al., 2006). A safety case is a prerequisite for the construction license of the repository and it is scheduled to become completed on the year 2012. This report is one of the interim reports and represents the element "Radionuclide transport" in Posiva's safety case report portfolio 1 (Figure 1-1). It describes the radionuclide release and transport analysis named RNT-2008. The domain of the report starts from the release of radionuclides from fuel elements and extends to the activity releases from the geosphere to the biosphere. RNT-2008 gathers information from all the other portfolio reports at the three upmost levels of Figure 1-1, except the biosphere assessment, and outputs the fluxes of radioactive substances entering the biosphere. Thus, the transport analysis feeds to the biosphere assessment. While the radiation dose rates are principally calculated in the biosphere assessment, the so called indicative dose rates are calculated in RNT-2008 using the dose conversion factors of a stylised drinking water well scenario, WELL- 2008 (ICRP, 1991). Both activity fluxes and well dose rates serve the transport analysis itself as immediately available safety indicators. As for the whole safety case, the regulatory dose rate constraints are applied for the first 10 000 years after the repository closure. Thereafter, the constraints for the activity release rates to the environment apply. However, in RNT-2008 both the release rates to the environment and the corresponding indicative dose rates are presented without the 10 000 year limit, up to one million years. 1 The structure of Posiva's safety case reports was revised in 2008 and the new structure, presented in Posiva (2008), will be adopted in the reporting of coming transport analyses. The radionuclide release and transport analysis will be presented in an Analysis of Scenarios report.

8 Beside TILA-99, a more recent reference work for this assessment is the radionuclide transport analysis for KBS-3H (Smith et al., 2007). KBS-3H is a KBS-3 variant with horizontal deposition of fuel canisters. The relationship between these three analyses are methodologically close to each other. The foundation of this assessment was laid down by applying the Finnish regulatory requirements. It also is in accordance with the methodology broadly acknowledged by the international community of experts. There is strong scientific evidence suggesting that the copper walls of the canisters will resist corrosion for such a long period of time that the residual radioactivity of the content will decay below harmful levels, before penetrating holes will develop in the canister. In the main scenario of Posiva's safety case, it is assumed that all canisters will be intact at the deposition time (Pastina & Hellä, 2006). Except for initially defective canisters or those breached due to extensive rock block movements, most canisters are expected to last more than one million years. The design criterion for the corrosionlimited lifetime of a canister in the expected repository conditions shall be at least 100 000 years (Posiva, 2006). After approximately 250 000 years, the activity remaining in the fuel will be similar to that of a large uranium ore body, and will still be isolated from the surface by the bedrock. Despite the expected long lifetime, it is assumed in all the assessment scenarios of RNT-2008 that radionuclides will start to leak out of a canister either immediately after the repository closure or later. Characteristic to the safety cases of radioactive waste repositories is the pronounced need for treatment of uncertainties. In RNT-2008 the uncertainties are covered principally by excessive modelling assumptions and in the second place by parameter variations. The methods used for the analysis include assumptions that are conservative. This means that they shall ensure that the results, with high degree of certainty, overestimate the radiation exposures or radioactivity releases. A purely deterministic approach was adopted in this analysis, meaning that the data for each repository calculation case were individually specified, without sampling from probabilistic distributions. In RNT-2008 it is assumed that the disposal system will behave in many unexpected and adverse ways, depending on the calculation case: 1) a disposal canister is assumed to be initially defective or through-corrode at a high corrosion rate, 2) the bentonite buffer is assumed to initially partly miss its safety function or become later strongly eroded, 3) rock shear movements will damage the canister, 4) groundwater flow rates will be higher than the flow analyses suggest, 5) groundwater chemistry will be less favourable than predicted, or 6) gas generation will enhance the release of radionuclides. The robustness of the repository design and the relative importance of the components of the system are illustrated in the report through a number of sensitivity cases and hypothetical "what if" cases. Human intrusion into the repository will be treated within the biosphere assessment. The calculation cases only include single-canister failures, leaving multiple-canister failures to be handled in the next analysis phase. The derivation of scenarios and calculation cases was attempted to be made in a clear and comprehensible way.

9 This report is a concise and focussed presentation of the radionuclide release and transport analysis and avoids unnecessary repeating of the contents of the other reports in Posiva's report portfolio. It is aimed to offer transparency, traceability, and true reproducibility of calculations. The groundwater flow analysis and the solute transport analysis, that provide most essential input to RNT-2008, are summarized in this report from a background report by Löfman & Poteri (2008). The central parts of the report are organised as follows: Chapter 2: the general safety regulations for spent fuel disposal issued by the Radiation and Nuclear Safety Authority (STUK). Chapter 3: the description of the disposal concept, repository and site. Chapter 4: a summary of the groundwater flow analysis and the solute transport analysis. Chapter 5: the assessment scenarios, the release mechanisms from fuel elements, and the considered transport processes with their mathematical formulations. Chapter 6: the description of calculation cases and the input data for the analysis. Chapter 7: the results, presented as indicative drinking water well dose rates and activity fluxes to the environment. Chapter 8: the overview of results and discussion. Figure 1-1. The main reports in Posiva's safety case portfolio as presented in the 2005 Safety Case Plan by Vieno & Ikonen (2005).

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11 2 SAFETY CRITERIA The regulatory requirements for a spent fuel repository are set out in the Government Decision on the safety of the disposal of spent nuclear fuel (STUK, 1999) and, in more detail, in Guide YVL 8.4 issued by the Radiation and Nuclear Safety Authority, STUK (2001). A detailed discussion of regulatory requirements related to the safety case is given in Posiva's TKS-2006 report on its programme for research, development and technical design (Posiva, 2006). Some key points relevant to the present report are summarised below. Guide YVL 8.4 distinguishes between the adequately predictable future (also referred to by the regulator as several thousand years ), during which conservative estimates of dose must be made (i.e. estimates that tend to over-estimate dose where there is uncertainty), and the era of large-scale climate changes when periods of permafrost and glaciations are expected, and radiation protection criteria are based on constraints on nuclide-specific activity releases from the geosphere to the environment. Posiva s interpretation of the duration of the adequately predictable future is typically 10 000 years. The annual effective dose constraint for the most exposed members of the public applicable to the adequately predictable future is 10-4 Sv, while the average annual effective doses to other members of the public should, according to the regulations, remain insignificantly low. It is also stated in YVL 8.4 that the radiation exposure of flora and fauna shall remain clearly below the levels that would cause decline in biodiversity or other significant detriment to any living population on the basis of the best available scientific knowledge. Moreover, rare animals and plants as well as domestic animals shall not be exposed detrimentally as individuals. Compliance with these requirements is not discussed in the present report, but is considered in the biosphere analysis. YVL 8.4 also gives a qualitative requirement that: The barriers shall effectively hinder the release of disposed radioactive substances into the host rock for several thousands of years. In the long term, Guide YVL 8.4 states that the sum of the ratios of nuclide-specific activity releases to their respective activity release constraints shall be less than unity in order to satisfy regulatory requirements. Activity release constraints, as set out in Guide YVL 8.4, are shown in Table 2-1. Guide YVL 8.4 covers all the safety relevant radionuclides considered in the present report, with the exception of Mo-93. For the purposes of this report, an activity release constraint of 3 GBq per year is assigned to this radionuclide. This is based on the activity of Mo-93 needed to give an annual dose of 10-4 Sv in the indicative stylised drinking water well scenario described in Chapter 6.2 of this report. According to Guide YVL 8.4, when comparing calculated activity releases with the constraints, the calculated values can be averaged over 1 000 years at most. Guide YVL 8.4 gives some indication as to the types of evolution scenarios to be considered when evaluating doses and activity releases. It states that:

12 A scenario analysis shall cover both the expected evolutions of the disposal system and unlikely disruptive events affecting long-term safety. The scenarios shall be composed systematically from features, events and processes, which are potentially significant to long-term safety and may arise from: mechanical, thermal, hydrological and chemical processes and interactions occurring inside the disposal system; external events and processes, such as climate changes, geological processes and human actions. Table 2-1. Activity release constraints, as set out in Guide YVL 8.4. Radionuclides Activity release constraints (GBq a -1 ) Long-lived alpha-emitting Ra, Th, Pa, Pu, Am and Cm isotopes 0.03 Se-79; I-129; Np-237 0.1 C-14; Cl-36; Cs-135; long-lived uranium isotopes 0.3 Nb-94; Sn-126 1 Tc-99; (Mo-93 see main text) 3 Zr-93 10 Ni-59 30 Pd-107; Sm-151 100 The Guide goes on to state on unexpected evolutions: The base scenario shall assume the performance targets defined for each barrier, taking account of the incidental deviations from the target values. The influence of the declined overall performance of a single barrier or, in case of coupling between barriers, the combined effect of the declined performance of more than one barrier, shall be analysed by means of variant scenarios. Disturbance scenarios shall be defined for the analysis of unlikely disruptive events affecting long-term safety. The importance to long-term safety of unlikely disruptive events shall, according to the Guide, be assessed. According to STUK, these events are to include at least boring a deep water well at the disposal site; core drilling hitting a spent fuel canister; and a substantial rock movement occurring in the environs of the repository. Section 2.4 of Guide YVL 8.4 states that, whenever practicable, estimates of the probabilities of activity releases and radiation doses arising from unlikely disruptive events impairing long-term safety should be made. These probabilities should be multiplied by the calculated annual radiation dose or activity in order to evaluate the importance to safety of an event. In order to satisfy regulatory requirements, the expectation value should remain below the radiation dose or activity release constraints referred to above. If, however, the resulting individual dose implies deterministic radiation impacts (dose

13 above 0.5 Sv), the order of magnitude estimate for its annual probability of occurrence should be 10-6 at the most. The human intrusion cases will be addressed in the biosphere assessment instead of the radionuclide transport analysis. In the very long term, after at least several hundred thousand years, Guide YVL 8.4 states that no rigorous quantitative safety assessment is required, but the judgement of safety can be based on more qualitative considerations. The types of considerations relevant to safety in the very long term are discussed further in Ruokola (2002). In the present study, safety in the very long term is addressed simply by extending the release and transport calculations up to a million years. Further handling of the farthest future evolution will be found in a Complementary Considerations report (see Figure 1-1 and the Safety Case Plan report [Posiva, 2008]). The present report focuses on the comparison of calculated indicative doses (from a drinking water well) and activity releases with the regulatory guidelines.

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15 3 DISPOSAL CONCEPT The disposal concept is based on a multibarrier principle. A design premise of the principle is that the barrier components complement each other to the extent that failure of one component does not compromise repository safety. The safety rests on the longterm isolation of radionuclides within copper-iron canisters surrounded by a buffer of bentonite clay located deep underground in crystalline rock. This radionuclide release and transport assessment relates to the KBS-3 disposal concept (SKBF/KBS, 1983) in which copper canisters equipped in the present design with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at a depth of approximately 400 500 m in saturated, sparsely fractured crystalline rock. The fuel canisters will be emplaced in vertical deposition holes bored down from the floors of horizontal disposal tunnels. In order to distinguish the reference KBS-3 concept from its horizontal alternative, KBS-3H, we refer to the vertical concept by the abbreviation KBS-3V. The disposal facility will be located at Olkiluoto, an island in the community of Eurajoki on the south-west coast of Finland (Figure 3-1). The assessment utilizes site information presented in the Posiva site description of the year 2006 (see Chapter 3.2). Figure 3-1. Location of Olkiluoto. 3.1 KBS-3V concept In the KBS-3V concept the spent fuel is encapsulated in canisters with a cast-iron insert and a copper overpack with a nominal thickness of 50 mm. Once filled and sealed, the copper-iron canisters will be emplaced individually in vertical deposition holes down from the floors of the deposition tunnels. The space between the canisters and the rock wall of the borehole will be filled with compacted bentonite. The deposition tunnels are planned to be backfilled primarily with pre-compacted bentonite blocks, but the use of

16 in-situ compaction of loose bentonite is an option (Saanio et al., 2006). The tunnels and shafts will be backfilled, and sealing plugs will be emplaced to block pathways for groundwater flow and to hold buffer and backfill materials in place. The function of the disposal canister is to contain and isolate the spent fuel from the surrounding environment. It can be regarded as the most important barrier in the repository. According to the design basis, the canister overpack shall provide corrosion resistance for at least 100 000 years in the repository. The canister insert is judged to be able to provide sufficient mechanical strength to withstand the loads caused 1) by hydrostatic pressure from groundwater at the disposal depth even during glacial cycles, 2) by the pressure from the swelling of the buffer, and 3) by minor rock shear movements. The function of the multibarrier system is detailed in Figure 3-2. The fuel canisters can be retrieved from the deposition holes, if needed or decided for some reason, during the operational phase as well as after the repository closure (Saanio & Raiko, 1999). BEDROCK isolates the repository from biosphere provides protection against surface and near surface processes provides favourable and predictable rock mechanical, geochemical and geohydrological conditions limits and retards inflow and release of harmful substances to and from the repository TUNNEL BACKFILL AND SEALING STRUCTURES prevent the tunnels and excavation disturbed zones (EdZs) for becoming significant transport pathways keep the buffer and canister in place in the deposition hole contribute to keeping the tunnels rock-mechanically stable chemically and mechanically stable no harmful effects on other barriers BUFFER plastically isolates the canister from rock and protects it against minor rock displacements keeps the canister in place in the deposition hole mass transport predominantly by diffusion conducts the heat from canister to the rock has sufficient permeability to gases filters colloids and prevents growth of microbes chemically and mechanically stable no harmful effects on other barriers COPPER-IRON CANISTER under the influence of the expected evolution remains intact at least for 100 000 years withstands mechanical loads remains subcritical conducts the decay heat and attenuates the radiation from the spent fuel no harmful effects on other barriers Figure 3-2. Functions of the bedrock and engineered barrier system in the KBS 3V disposal concept (Vieno & Ikonen, 2005).

17 The influence of the excavation disturbed zone (EdZ) or excavation damaged zone (EDZ) in the repository is considered in this work only through the groundwater flow modelling (for the explanation of EdZ and EDZ, see e.g. Chapter 3.1.3 in Pastina and Hellä [2006]). In the equivalent porous medium (EPM) model, used for the flow analysis, the effects of the EdZ were taken implicitly into account in the backfilled tunnels by modelling those tunnels with larger diameter than that of the real tunnels. More details are given in Chapter 4.1.3. 3.2 Olkiluoto site and repository layout The analyses of this report are based on the 2006 site description (Andersson et al., 2007) and the 2006 repository layout (Saanio T. et al., 2006). As typical sparsely fractured hard rock, the bedrock at Olkiluoto is characterized by a network of fractures and fracture zones. The frequency, spatial distribution, size distribution, shape and orientation of these structures affect both the hydraulic and mechanical properties of the rock. The fracture zones constitute dominant paths for groundwater flow and their size also affects the scale of any rock shear movements that may take place in the event of earthquakes (Andersson et al., 2007). The repository layout has been adapted to avoid major fracture zones (Figure 3-3, Kirkkomäki 2006). A potential extension of the repository to the eastern part of the island has been evaluated and is shown in Figure 3-4. Figure 3-3. The repository layout at a depth of 420 m (Kirkkomäki, 2006).

18 In principle it also is possible to enlarge the repository to greater depths. In such a case the safety impacts of e.g. higher rock stress and higher groundwater salinity on the repository need be examined. Figure 3-4. The repository site at Olkiluoto. The green-marked disposal tunnels are included in the current layout and a possible extension area stretches in the eastern region of the island up to the black line. (Saanio T. et al., 2006)

19 4 GROUNDWATER FLOW AND SOLUTE TRANSPORT Two distinct approaches were applied for the groundwater flow and solute transport analyses. Equivalent porous medium (EPM) modelling was applied for the groundwater flow in the geosphere and the repository tunnels. The other approach, discrete fracture network (DFN) modelling was applied for the groundwater flow and solute transport solely in the geosphere, excluding the repository tunnels. Both of the analyses provided input for the radionuclide release and transport analysis. The EPM simulations provided averaged groundwater flow rates, while the DFN analysis was used to assess the hydrodynamic control of geosphere retention, expressed as the parameter WL/Q along the potential migration paths (Equation 4-1 in page 43). In the EPM simulations it also is possible to handle thermally induced flow, the salinity of groundwater, and time-dependence of the system. The flow is assumed to take place through blocks of porous material. The applied DFN model in turn is able to handle steady-state flow and solute transport in the fracture system of sparsely fractured bedrock. Both kinds of modelling were carried out independently, with all needed boundary conditions and other input. The results of the EPM modelling were applied for the selection of groundwater flow rates in the near-field release and transport analysis, whereas the results of the DFN modelling were used to provide the transport resistance for the far-field transport analysis. The consistency of the results of these two analyses is discussed in Chapter 4.3 (page 56). 4.1 Equivalent porous medium (EPM) modelling of groundwater flow The objective of the site-scale equivalent porous medium (EPM) modelling of groundwater flow is to provide support for the radionuclide release and transport analysis. The EPM simulations were carried out in a similar manner to those in the supporting groundwater flow modelling for a study on the expected evolutions of the spent nuclear fuel repository at Olkiluoto (Pastina and Hellä, 2006). The current simulations, however, were based on the latest flow model 2006 (Andersson et al., 2007), in which e.g. the properties of the hydrogeological zones and the sparsely fractured rock between the zones have been updated. The evolution of the groundwater flow conditions was simulated from the start of the ONKALO excavations in September 2004 until 10 000 years after the emplacement of the first canister in 2020 (it was assumed that the current climate conditions will prevail at least 10 000 years). The simulations covered all the main phases of the repository operation, i.e. the excavation of ONKALO in 2004 2012, the preparatory phase in 2012 2019 (the excavation of the first repository tunnels),

20 the operational phase in 2020 2112 (the excavation of the repository tunnels, the disposal of the canisters, the backfilling and resaturation 2 of the disposal tunnels), the backfill phase in 2112 2130 (the backfilling and resaturation of the central tunnels, the shafts and the access tunnel), and the post-closure phase after 2130 (the reversion of groundwater flow conditions back to natural conditions). In each of these phases there are characteristics that dominate the groundwater flow conditions. Until the post-closure phase the flow is mainly affected by the disturbances introduced by open tunnels. After that the hydrological disturbances will cease and both post-glacial land uplift (Påsse, 1996; Pastina and Hellä, 2006) and the thermal effects of the decay heat of the spent fuel start to play a major role. The decay heat will raise the temperature of the repository and the surrounding bedrock several tens of degrees for many centuries, which may cause changes to the flow conditions over distances of several hundred metres from the repository. After a few thousand years, the thermal effects will diminish, whereas the land-uplift will continue to affect the flow conditions until the end of the simulated period of time of 10 000 years. The climatic conditions were not assumed to change over the simulated period of time. Consideration of the prevailing processes affecting the flow during the construction, operational and post-closure phases required coupled and transient modelling of groundwater flow, solute (salt) transport and heat transfer. A summary of the modelling approach and assumptions as well as the applied input data and the results are provided below. The details are presented in a background report on the modelling of the groundwater flow and solute transport (Löfman and Poteri, 2008). 4.1.1 Modelling approach In this study, the modelled bedrock volume is conceptually divided into two kinds of hydraulic units: planar hydrogeological zones (HZ), the sections of the bedrock featuring a higher fracture density and a greater ability to conduct water, and the sparsely fractured rock (SFR) between the zones. In both units, the averaged hydraulic characteristics of fractured rock were used (Andersson et al., 2007). From a standpoint of groundwater flow and heat transfer the hydraulic characteristics of the HZs and the SFR were modelled with the equivalent porous medium (EPM) approach. Thus, the fractured system is treated as a single continuum with representative averaged characteristics and water is assumed to flow everywhere in the system. From the point of view of flow, the matrix blocks with essentially stagnant water are not important and they can be approximated to be in pressure equilibrium with the fractures, which justifies the use of the EPM approach. Thermal conduction was assumed to be 2 The repository was assumed to be initially saturated, i.e. resaturation was not simulated.

21 the dominant heat transfer mechanism, due to the low hydraulic conductivity of the bedrock. From a standpoint of solute transport the fractured rock is modelled with the dualporosity (DP) approach. In the DP model, the system is assumed to consist of two overlapping continua: the fractures with flowing water and the matrix blocks with essentially stagnant water representing the rest of the system. Advection and dispersion are the dominant processes within the water-bearing fractures, whereas in the matrix solutes are transported only by diffusion. In the DP approximation the matrix diffusion of solutes in the rock blocks with stagnant water moderates the transport of solutes in the water-bearing fractures. The mathematical model describing groundwater flow, solute transport and heat transfer consists of three coupled partial differential equations written for the pressure, the total concentration of the dissolved solids (TDS) and temperature. In the equations the density and viscosity of water are dependent on salinity and temperature. The hydraulic conductivity increases with increasing temperature and decreasing salinity (Figure 4-1). The water table drawdown resulting from the inflow of water into the open tunnels during the construction and operational phase was ignored in the simulations, because the objective of this study is to analyse the flow conditions in the future and the water table will recover back to the natural level within few years after the closure of the tunnels. Furthermore, even during the construction and operation phases the impact of drawdown can be considered small due to the efficient grouting of the tunnels. The model calculates the disturbances caused by the open tunnels on the salinity field. The flow, solute transport and heat transfer equations were solved numerically using the finite-element program package FEFTRA (Löfman et al., 2007) developed at VTT for groundwater flow modelling in the site investigations for the spent fuel repository. The complex geometry of the HZs is discretised into a finite-element mesh with an adaptive and recursive octree algorithm that adapts the element size to the so-called well characterized area (WCA, the rock volume where the borehole investigations have been focused) and the HZs. The WCA is bordered from east to west between [4750, 7000] and from south to north between [1500, 3500]. The HZs were gridded with sets of triangles of appropriate hydrogeological properties fitted onto the faces of tetrahedra along their planes. The element size ranged from 8 m (the tunnels) and 16 m (at a depth of 0 200 m inside the WCA) to 130 m (under the sea). The resulting mesh consists of approximately 800 000 nodes and 5 million tetrahedral and triangular elements, which occupied 4.5 Gbytes of computer memory.

22 Figure 4-1. The impact of salinity and temperature rise on the relative hydraulic conductivity (K r, the hydraulic conductivity relative to the reference temperature of 20 o C). The curves are based on the dependence of the viscosity and the density on the salinity and temperature (Löfman and Poteri, 2008). c denotes salinity (TDS). 4.1.2 Site-specific flow model The modelling is based on the latest (2006) site description of Olkiluoto (Andersson et al., 2007). As stated before, the modelled bedrock volume is conceptually divided into two hydraulic units: planar hydrogeological zones (HZ), and the sparsely fractured rock (SFR) between the zones. The geometry of the HZs is based on the geological brittle deformation zone model, which had been reconciled with available measurements of the transmissivity and the hydrogeological responses to various field activities such as pumping tests, water sampling, etc. (Ahokas et al., 2007). The geometry consists of 19 deterministic zones (5 large scale lineaments outside the island and 14 local zones located in the centre of the island; Figure 4-2). The depth of the modelled volume is 2 km. The SFR was divided into five hydrologic layers with different properties in each layer (Löfman and Poteri, 2008). The hydraulic properties of the HZs and the SFR between the zones as well as other properties of bedrock and water required in the simulations are based on the site information discussed in detail in Andersson et al. (2007). Values for the flow porosity were obtained with an approach that couples the porosity to other properties of rock such as transmissivity and/or hydraulic conductivity (see details of the approach in the background report of the groundwater flow modelling; Löfman and Poteri, 2008). The thermal properties of rock (thermal conductivity, diffusivity, specific heat) were based on the averages of the measured properties of mica gneiss samples taken from the boreholes at Olkiluoto (Kukkonen, 2000). These properties depend on the temperature.

23 The conductivity and diffusivity decrease and the specific heat increases with increasing temperature. Especially, the thermal diffusivity, which is the most important parameter regarding the maximum canister temperatures, may decrease by up to 17 % from 20 to 100 o C. Since the laboratory measurements were carried out at room temperature (20 o C), Kukkonen (2000) applied simple corrections to them in order to obtain estimates for the average thermal properties at 60 o C, which according to the disposal scenarios was considered to be the highest temperature of the rock surface around a deposition hole. Thus, the average values appropriate to a temperature of 60 o C were used in this work. The same thermal properties were assumed for both the bedrock (HZ and SFR) and the backfilled tunnels. A short discussion of the sensitivity of the calculated temperature and salinity field to the thermal properties is given by Löfman (2005). Figure 4-2. ONKALO, the repository tunnels and the selected nearby hydrogeological zones of the site model 2006 (Andersson et al., 2007). 4.1.3 ONKALO and the repository tunnels The ONKALO layout version of June 2005 was used in the analysis. The repository layout is based on the preliminary layout adaptation for a one-storey repository (Kirkkomäki, 2006). All the ONKALO and repository tunnels were modelled explicitly according to the layout taking into account a stepwise programme of excavation (Figure 4-3) and associated disposal schedule (Kirkkomäki, 2006). Accordingly, each disposal tunnel is open and ventilated for only a short period, after which the decay heat begin to

24 raise the temperature of bedrock and groundwater. The disposal holes are not included in the model. Finite element model of the tunnels The conceptual geometry of the 3-D tunnel system was simplified to a wireframe model, in which each tunnel segment was represented by a line located in the centre of the actual segment, i.e. no physical extension, such as could be defined by a radius, was considered (Figure 4-4). Each line (tunnel) segment was modelled as a set of nodes of the finite element mesh by using an appropriate internal boundary condition for each node. The boundary condition was selected considering the processes related to the different tunnel segments. For example, the open tunnel nodes were treated as hydraulic sinks applying an atmospheric pressure corresponding to the elevation of the node, whereas the backfilled repository tunnel nodes were treated as point heat sources by assigning a time-dependent decay heat power source to each node. The finite element mesh was refined around the tunnels so that the average length of the sides of the tetrahedral and triangular elements was about 8 m. Grouting Grouting, which will be applied during the excavation and operational phase to limit the inflow to ONKALO and the repository, reduces the conductivity of the grouted zones and rock around the tunnels. In the modelling this was taken into account by reducing the hydraulic conductivity of the SFR around the tunnels in the topmost 50 m bedrock from 6.0 10-8 to 5.0 10-9 m/s in the horizontal direction. In the vertical direction and deeper than 50 m the hydraulic conductivity of the SFR is small enough to render reduction unnecessary. The hydrogeological zones (HZ001, HZ002, HZ004, BZF099, HZ19 and HZ20) intersecting the tunnels (ONKALO and/or the repository) are assumed to be grouted, so as to have a transmissivity of 1.0 10-8 m 2 /s. In the finite element model the grouting was assigned to the elements (the average length of sides 8 m) adjacent to the tunnels. Thus, the grouting in the model extends deeper than the estimated penetration depth of the grouting cement in reality. Backfilled tunnels and EDZ The excavation disturbed/damaged zone (EdZ/EDZ) and the tunnel backfill were assigned uniform, averaged hydraulic properties, because the resolution of the mesh adjacent to ONKALO was an order of magnitude larger than the extent of the EDZ. In a base case of the hydraulic calculations the conductivities of the closed tunnels were assumed to be equal to that of the surrounding rock before the excavation (9.0 10-11 m/s), which is nearly the same as the guideline value in the design criteria (Gunnarsson et al., 2003; Gunnarsson et al., 2007). The three additional simulation cases consider higher hydraulic conductivity for the backfilled tunnels (and their associated EDZs): 10-9, 10-8 and 10-7 m/s (at the reference temperature of 20 o C). In these cases, the hydraulic properties of the tetrahedral elements next to the tunnels were modified (Figure 4-4). Thus, the backfilled tunnels were modelled with somewhat larger main dimensions (a diameter of approximately 16 m) than those of the real tunnel design (4 m). The hydraulic properties of the rest of the tunnels were assumed to be similar to those of the disposal tunnels.

25 Figure 4-3. The excavation schedule of the tunnels, the different colours designating the different stages (Kirkkomäki, 2006).

26 Figure 4-4. Finite element model of the tunnels. The conceptual geometry of the 3D tunnel system was simplified to a wireframe model, in which each tunnel segment was represented by a line located in the centre of the actual segment i.e. no physical extent (such as would be represented by a radius) was considered. Each line (tunnel) segment was modelled as a set of nodes of the finite element mesh by using an appropriate (internal) boundary condition (e.g. hydraulic sink or heat source) for each node. 4.1.4 Results The primary results needed for the canister-scale radionuclide release and transport (RNT) simulations were the groundwater flow rates, the salinity and the temperature in the tunnel backfill at different times. In addition, the results characterizing the evolution

27 of the groundwater flow conditions during the construction, operational and postclosure phases were provided (the inflow of the groundwater into the tunnels, the groundwater flow rates in bedrock, the groundwater salinity in the vicinity of the tunnels, the temperature rise in bedrock, the hydraulic gradient). The evolution of the groundwater flow conditions was analysed assuming that the hydraulic properties of the closed tunnels were identical to those of the surrounding rock before the excavation. In addition, groundwater flow conditions in the tunnel backfill were analysed with higher hydraulic conductivities in the tunnel backfill. Evolution of groundwater flow conditions During the ONKALO excavations and the operational phase the open tunnels draw water from all directions (fresh water above and saline water below). Most of water is drawn by the ONKALO tunnels and the four repository shafts, which are open until the end of the operational phase and intersect the HZ19 and HZ20 zones. In the operational phase the inflow rate evolves along with the disposal of canisters and the backfill of the tunnels (each disposal tunnel is open for only a short period). The calculations show that the total inflow rate is at its highest about 135 L/min in the year 2080. The open tunnels result in upconing of deep saline groundwater around and below ONKALO or the repository tunnel system. The calculations indicated that locally, the salinity may rise to rather high values, even if the inflow is limited by the grouting. At the repository level the maximum salinity will rise from 13 g/l to 23 g/l and at the lowest point of the tunnel system at a depth of 550 metres from 20 g/l to 50 g/l. Immediately after the emplacement of the first canister into the deposition hole the decay heat of spent fuel starts to increase the temperatures in the surrounding rock. The temperature distribution is linked to the disposal schedule (Figure 4-3), and the maximum values (approx. 40 o C) in the individual tunnels are reached accordingly (Figure 4-5). After the closure of the uppermost parts of ONKALO and the shafts at 110 years after the disposal of the first canister in 2020, the inflow rates will fade out and the flow and salinity fields will start to gradually recover towards the natural conditions from the hydrological disturbances caused by the inflow. Under natural conditions, the water table variations corresponding to variations in the topography are the main driving force of the groundwater flow resulting in a downward flow below the present area of the island. However, during the post-closure phase the groundwater flow will also be affected by the thermal disturbances resulting from the heat generation of spent fuel. The simulations indicate that the higher temperatures will affect not only the magnitude of the flow rates but also the flow directions (Figure 4-6 and Figure 4-7). The temperature rise will decrease the viscosity and density of water resulting in a buoyancy effect, which, at its peak, tends to change the flow direction upwards and/or strengthen the already upward flow, especially at times when the temperatures are high. Thus, during the post-closure phase there will be two opposite driving forces affecting the flow conditions: 1) the topographical gradient, and 2) the thermal buoyancy. If temperature gradients are high enough, the thermal effects will exceed the topographical driving forces and the flow directions will change to upwards (if not already upward). Although the temperature will start to decrease along with the declining heat output, it will still

28 remain high enough for hundreds of years, to delay the recovery of the flow conditions towards the natural state. After thousands of years all the disturbances will cease and the water table variations, which will be affected by the post-glacial land uplift, will remain the dominant process affecting the flow conditions. The land uplift, which is expected to be around 40 metres during the next 10 000 years, will drive the more saline groundwater downwards, and the fresh and brackish groundwater presently above the repository depth will start to gradually replace the saline groundwater at the repository depth. Groundwater flow in the tunnel backfill The main results needed for the radionuclide transport calculations are the groundwater flow rates, the hydraulic gradient, the salinity and the temperature in the tunnel backfill at different times. The results for groundwater flow in the tunnel backfill presented below are the averages of the nodal values in the first disposal tunnel (the preparatory construction stage 0; see Figure 4-3), which was considered to be representative of the repository from the point of view of canister-scale modelling. In addition to using the same hydraulic conductivity for both the closed tunnels and the surrounding rock, the three simulation cases were introduced, in which higher hydraulic conductivities (1.0 10-9, 1.0 10-8 and 1.0 10-7 m/s, the initial values at a temperature of 20 o C) were assumed for the tunnel backfill. The calculated results in the first disposal tunnel are presented in Table 4-1. The flow rate in the first tunnel was obtained by multiplying the magnitude of the average Darcy velocity vector by the cross-sectional area of the real tunnel (14 m 2 ). The disposal tunnels are intersected by the sub-vertical zones HZ001 and BFZ099 (see Figure 4-2), which extend from the surface to deeper than the repository depth. The low transmissive bottom part of the zone HZ001, which intersects the first tunnels (stage 0), and the moderately transmissive upper part of the zone BFZ099, which intersects the tunnels of the construction stages 3 6, connect the disposal tunnel system to the lowhead area of the surface. In addition, the zones HZ001 and BFZ099 together with the tunnels, in which the backfill conductivity is one to three orders of magnitude higher than that of the surrounding rock, constitute the flow system, in which the water flows from the surface downwards along the shafts and the access tunnel (located in higher elevation of topography) further on along the disposal tunnels towards BFZ099 (a better connection to the surface than HZ001), where the flow direction is upwards heading to the surface (due to the lower elevation of topography). The results show that assuming a similar conductivity for the closed tunnels and the surrounding rock the groundwater flow in the disposal tunnels changes with time according to the temperature distribution (Figure 4-6 and Figure 4-8). Under natural conditions the flow direction in the disposal tunnels is downwards. With rising temperature, however, the magnitude of the downward flow starts to decrease, and after the temperature has increased enough, the thermal driving forces exceed the topographical driving forces and the flow directions change to upwards. On the other hand, after the bedrock and the water have cooled down enough the topographical driving forces start to dominate the flow again. The aforementioned effects are naturally strongest in the

29 tunnels where the temperature rise is at the highest, i.e. in the centre of the repository panels. In the first disposal tunnel, which is located at the periphery of the first panel, the temperature rise affects somewhat the flow directions, but not the magnitude of the flow, from which the average flow rates were calculated for the radionuclide transport calculations (Table 4-1). The higher backfill conductivity strengthens the connection of the tunnels to the surface, turns the flow directions more and more along the disposal tunnels towards the zone BFZ099, and increases the magnitude of the Darcy velocity and the flow rates in the tunnels. This tendency is strengthened further by thermally induced driving forces, which in this case increase the magnitude of the flow along the tunnels, instead of changing the flow directions to upwards. Slight variations in the flow rates reflect the combined effects of changes in temperature, salinity and elevation caused by the land uplift. The hydraulic gradient in the repository tunnels at 1 000 years after the disposal of the first canister is presented in Figure 4-9, which shows that if the hydraulic conductivity of the backfilled tunnels is assumed to be equal to that of the surrounding rock the gradient is between 0.1 1.0 % in most of the tunnels. The maximum values are 1.5 2 % and they occur in the tunnels that are intersected by the hydrogeological zones BFZ099 and HZ001. If the conductivity of the backfilled tunnels is assumed to be higher than the surrounding rock, the gradient field becomes more even and the magnitude of the gradient decreases. With conductivity values of 10-9, 10-8 and 10-7 m/s, the maximum values of gradient are between 1.0 1.5 %, 0.5 1 % and 0.1 0.5 %, respectively, and these are not necessary associated with the intersecting zones.

30 Figure 4-5. The calculated temperature rise ( o C) in the repository. The maximum temperatures are presented at the top classified by the construction stages (see Figure 4-3.) The temperature rise in all the tunnels during the post-closure phase at 110 years after disposal of the first canister in 2020 is shown at the bottom. In the upper figure, the maximum means a maximum value anywhere in the repository at a given stage and time.

31 Without thermal effects With thermal effects Downwards Upwards Figure 4-6. The calculated average groundwater flux Q (m/a) 3 through a horizontal crosssection (1.9 x 2.4 km 2, at a depth of 400 m) during the post-closure phase at 110 years after the disposal of the first canister in 2020 (the time when the last parts of the tunnel system will be closed). The flux was obtained by dividing the cross-section into 50 x 50 m 2 sub-areas and calculating the perpendicular component of the Darcy velocity at each sub-area. The left side of the figure presents the results of the coupled simulation of pressure and concentration ignoring the thermal effects of the decay heat of spent fuel on the density and viscosity, and on the right side there are the corresponding results with the thermal effects included. The downward flow direction is presented on the top row and the upward direction on the bottom row. The repository is located at a depth of 420 metres. The hydraulic conductivity of the backfilled tunnels was assigned equal to that of the surrounding bedrock. See Figure 4-5 for the corresponding temperature rise. 3 Reduced from [m 3 m -2 a -1 ].

32 Without thermal effects With thermal effects Figure 4-7. Magnitude of the Darcy velocity q (m/s) in the horizontal cross-section (1.9 x 2.4 km 2, at a depth of 400 m) at the post-closure phase at 110 years after disposal of the first canister in 2020. On the left, the coupled simulation of pressure and concentration ignoring the thermal effects of the decay heat of the spent fuel on the density and viscosity, and on the right the corresponding simulation with the thermal effects. The hydraulic conductivity of the backfilled tunnels was assigned equal to that of the surrounding bedrock. See Figure 4-5 for the corresponding temperature rise.

K backfill = 9.0 10-11 m/s K backfill = 1.0 10-9 m/s K backfill = 1.0 10-8 m/s K backfill = 1.0 10-7 m/s 33 The 1 st disposal tunnel Figure 4-8. The calculated Darcy velocity q (m/s) vectors in the first repository panel (the preparatory construction stage 0, Figure 4-3) during the post-closure phase at 300 years after disposal of the first canister in 2020. The results calculated for the radionuclide transport modelling (Table 4-1) were the averages of the nodal values in the first disposal tunnel, which was considered to be representative for the repository.

34 Table 4-1. The calculated average temperature rise (dt), the average salinity (TDS), the average Darcy velocity (q) and the average flow rate (Q) in the backfill of the first disposal tunnel. The flow rate was obtained by multiplying the magnitude of the average Darcy velocity by the cross-sectional area of the real tunnel (14 m 2 ). The time column denotes the years after disposal of the first canister in 2020. The last tunnels at ONKALO will be closed after 110 years, after which the higher backfill conductivity was applied instantly to the whole tunnel system. The Darcy velocity vectors in the first repository panel at 300 years are presented in Figure 4-8. Time (a) dt ( o C) TDS (g/l) Hydraulic conductivity of the tunnel backfill (m/s) 9e-11 1e-9 1e-8 1e-7 q (m/s) Q (L/a) Q (L/a) 0 0.0 10.1 9.4e-10 415.0 100 22.6 7.0 2.2e-11 9.7 200 20.7 7.0 9.2e-13 0.41 1.63 9.7 41.9 300 23.3 6.9 9.4e-13 0.42 1.68 10.2 43.7 1000 19.9 6.4 9.2e-13 0.41 1.72 11.0 48.6 2000 13.1 5.7 9.0e-13 0.40 1.77 11.0 48.6 5000 8.0 3.7 9.2e-13 0.41 1.81 11.5 53.0 10000 5.5 1.7 9.7e-13 0.43 1.94 12.4 48.6

35 K backfill = 9.0 10-11 m/s K backfill = 1.0 10-9 m/s K backfill = 1.0 10-8 m/s K backfill = 1.0 10-7 m/s Figure 4-9. The calculated hydraulic gradient (%) in the disposal tunnels at 1 000 years after the disposal of the first canister (2020). Four different values were applied for the hydraulic conductivity of the backfilled tunnels (K backfill ). The gradients were obtained from the coupled simulation of pressure and concentration considering the thermal effects of the decay heat of spent fuel on the density and viscosity of groundwater. 4.2 Discrete fracture network modelling (DFN) The main objective of the Discrete Fracture Network (DFN) modelling is the assessment of the hydrodynamic control of retention along the potential release paths. The flow modelling is based on a site-scale DFN model and detailed description of the fracturing in the near field of the deposition holes. Small background fractures are important in the near field, because the deposition holes are located in sparsely fractured background rock (SFR). On the other hand, flow paths tend to be associated with larger

36 hydraulic features that diminish the significance of small fractures further away from the deposition holes. The DFN modelling is an independent flow modelling approach that is based on statistical description of the fracture observations and hydraulic measurements in boreholes. DFN simulations give an alternative representation of the geosphere flow conditions compared to the EPM simulations. The EPM simulations aim to realistic description of the flow-related processes by incorporating density dependent flow, repository tunnels, heat produced in the disposal canisters and transient modelling of the evolving boundary conditions. The DFN simulations aim to handle realistic distribution of the flow in fractured rock under steady state flow conditions determined by the water table and assuming that repository tunnels do not interfere with release paths. 4.2.1 DFN model parameters The hydraulic features were divided into three categories according to their size: The largest features are major hydrogeological zones (MHZ). They are complicated structures on a scale of hundreds of metres to kilometres and contain a multitude of individual fractures. Local hydrogeological zones (LHZ) may be composed of several fractures and LHZ may extend over several hundreds of metres. Water-conducting background fractures (WCF) are individual fractures that vary in size from a few metres to some hundreds of metres. A hydro-dfn model applies a simplified description of the hydraulic features in such a way that both fracture zones and individual water-conducting fractures are planar and two-dimensional, i.e. fracture zones have zero thickness in the model. There is neither increased fracturing nor any other zone of influence around the fracture zones and the rock matrix between fractures is considered to be impermeable. The representation of the hydraulic features in the DFN model is based on a division into deterministic and stochastic hydraulic features: Deterministic features include MHZs that extend over the whole of the Olkiluoto island and smaller LHZs that are deterministically observed only in the well characterised area (WCA) of the Olkiluoto site. The WCA is defined to be the area at Olkiluoto that is bordered from east to west between [4750, 7000] and from south to north between [1500, 3500]. Stochastic hydraulic features include LHZ's outside WCA and individual WCFs. It is assumed that the LHZ structures observed in the WCA are representative for the whole site 4. The statistical pattern of LHZ structures observed in the WCA is extended over the whole model domain. 4 In the EPM modelling it was assumed that the average hydraulic conductivity for the background rock is five times as high outside the WCA than inside the WCA.

37 Deterministic hydrogeological zones Definition of the deterministic MHZs and LHZs follows the 2006 Olkiluoto hydrostructural model (Andersson et al., 2007). The hydrostructural model of the year 2006 contains 19 different deterministic homogeneous fracture zones, as listed in Table 4-2. The hydrostructural model is based on a description of the structures by homogeneous effective transmissivity. In practice, deterministic zones are modelled as "transmissive surfaces" in ConnectFlow. In the transport model these zones give only minor contribution to the transport resistance of entire routes. Deterministic zones are illustrated in Figure 4-10. The high intensity of the deterministic LHZs in the WCA is clearly visible in the middle of the modelling domain. Table 4-2. Hydraulic transmissivities of the deterministic major (in bold) and local hydrogeological zones in the hydro-dfn model. Name Transmissivity (m 2 /s) Name Transmissivity (m 2 /s) HZ001 (z>-250 m) 1.0e-5 HZ20B_alt 3.2e-6 HZ001 (z<-250 m) 2.5e-8 HZ21 1.6e-8 HZ002 1.0e-6 HZ21B 7.9e-7 HZ003 6.3e-7 BFZ099 (z>-500m) 5.0e-7 HZ004L 1.6e-7 BFZ099 (z<-500m) 7.9e-9 HZ008 1.0e-5 LIN1 1.0e-5 HZ19A 1.6e-6 LIN2 1.0e-5 HZ19B 2.0e-6 LIN3 1.0e-5 HZ19C 3.2e-6 LIN4 1.0e-5 HZ20A 7.9e-6 LIN5 1.0e-5 HZ20AE 1.0e-6

38 Figure 4-10. Deterministic major and local hydrogeological zones (MHZ and LHZ) included to the Olkiluoto DFN model (a view from south). Zones are coloured by transmissivity. The well characterised area (WCA) is the region filled by LHZs in the middle of the model. Stochastic local hydrogeological zones A deterministic description of the LHZs in the WCA is possible due to the dense pattern of boreholes in the WCA. The lack of LHZ-scale deterministic features outside the WCA is caused by a limited number and sparser spatial pattern of the boreholes outside the WCA. In the present hydro-dfn model, the description of LHZs outside the WCA is based on the assumption that the statistical pattern of LHZs is similar inside and outside the WCA. The deterministic zones that are observed only in the WCA are interpreted to be deterministic LHZ structures. These structures are HZ001, HZ002, HZ003, HZ19A, HZ19B, HZ19C, HZ20A, HZ20AE, HZ20B_alt, HZ21, HZ21B and BFZ99. The mean size of them is about 1400 m x 1400 m. They are all dipping sub-horizontally towards south-east and transmissivities range from 7.9 10-9 m 2 /s to 1.0 10-5 m 2 /s. Based on these data, the pattern of LHZs is extended outside the WCA by applying following simplifications:

39 The WCA in the Hydro-DFN model is located between east [4750, 7000], north [1500, 3500] and in depth from [-1000, 0] (i.e. vertically through the whole model). The statistical LHZ model is used only for the model domain outside the WCA. All stochastic LHZs have a fixed size that is the average size of the deterministic LHZs observed in the WCA. Thus the size of the stochastic LHZs outside the WCA is 1400 m x 1400 m. The number of deterministic LHZs in the WCA is quite small so that fitting of the LHZ size distribution is not meaningful. Orientations of the observed LHZ in the WCA follow the major sub-horizontal fracture set of the background fracturing. Orientations of the stochastic LHZ structures are sampled from the subhorizontal fracture set given in Table 4-3. Orientations of the stochastic LHZ are more widely spread than the observed orientations, because the number of deterministic LHZs is too limited (only 12) for reliable estimation of the dispersion of the LHZ orientations. Transmissivities of the stochastic LHZs are represented by a truncated lognormal distribution. Upper and lower truncation limits are defined by the lowest and highest transmissivities measured for the deterministic LHZs in the WCA (from 7.9 10-9 m 2 /s to 1.0 10-5 m 2 /s). The variance of the transmissivity distribution is determined by the variance of the transmissivities of the deterministic LHZs. The mean of the transmissivity distribution is calibrated to give the same effective hydraulic conductivity inside and outside the WCA. Stochastic LHZs outside the WCA are uniformly distributed. The intensity of the non-isolated stochastic LHZs is calibrated to give the same mean number of borehole intersections that has been observed in the Posiva Flow Log measurements in the boreholes OLKR1 OLKR28. Parameters for the stochastic LHZs are collected in Table 4-3 and one realisation of stochastic LHZs with deterministic LHZs and MHZs is presented in Figure 4-11. The statistical extension of the LHZ structures in Figure 4-11 can be compared with Figure 4-10 that contains only deterministic MHZ and LHZ structures. Table 4-3. Calibrated parameters of the stochastic LHZs. Parameter Definition Fracture intensity P32 0.0125 m 2 /m 3 Orientation distribution (fracture pole vector) Fisher distribution mean orientation (trend/plunge) 330 / 70 Fisher κ value 6.9 Transmissivity distribution Truncated lognormal distribution Mean Log 10(T [m 2 /s]) -6.25 Std Log 10(T [m 2 /s]) 0.96 Range (T [m 2 /s]) [7.9 10-9, 10-5 ] Fracture size Fixed size 1400 m x 1400 m Model region Outside WCA

40 Figure 4-11. Deterministic MHZs and LHZs represented together with one realisation of the stochastic LHZs outside the WCA. This figure can be compared with Figure 4-10 that does not include stochastic LHZs outside the WCA. Water-conducting fractures The statistical model of the water-conducting background fractures (WCF) is based on up-to-date Posiva Flow Log (PFL) measurements and on a DFN geometry that integrates fracture observations from different scales (LaPointe and Hermanson, 2002). The intensities and transmissivity distributions of the fracture sets were calibrated using PFL measurements from boreholes KR1 KR28. The DFN model is composed of square fractures that are divided into three different sets based on the fracture orientations. Fracture sets are parameterised using Fisher distributions for fracture orientation and power-law distribution for fracture sizes. The minimum and maximum sizes of WCFs are 10 m and 500 m, respectively. It is assumed that fractures smaller than 10 m are not hydraulically important and, on the other hand, features larger than 500 m are likely observed deterministically (in the WCA) and thus represented by LHZs in the DFN model. The same fracture sets and fracture orientation and size distributions were also applied in the earlier DFN modelling of the Olkiluoto KBS-3H repository (Lanyon and Marschall, 2006).

41 Transmissivity data for the individual water-conducting fractures were measured using the PFL. There are a total of 2 244 measured individual fracture transmissivities in 54 boreholes. The present work applies data from deep boreholes OLKR1 OLKR28, where the inclination data of the boreholes were readily available. The number of measured individual fracture transmissivities in these boreholes is 1 409. Some of the PFL-measured fractures are associated to the known hydrogeological zones and they are not used to describe background WCF. The dataset of PFL fractures outside known hydrogeological zones contains 1234 individual fracture transmissivities. The analysis of borehole data indicates a clear depth-dependent trend both in the mean fracture intensity along the boreholes and in the fracture transmissivities. Earlier analysis of the hydraulic and fracture measurements at Olkiluoto suggested that it would be appropriate to divide fracturing into five layers according to depth (Ahokas et al., 2007): 0 50 m, 50 100 m, 100 200 m, 200 400 m and 400 900 m. Fracture intensity and transmissivity distributions are represented using these five depth layers. Parameters describing WCFs are presented in Table 4-4. Table 4-4. Calibrated DFN parameters of the water-conducting background fractures (WCF). Parameter Definition Set 1 Set 2 Set 3 Weight Relative intensity 56 % 13 % 31 % Orientation distribution Size distribution Fisher distribution (for the fracture pole vector) Power law Mean (trend/plunge) 330 / 70 77 / 2 150 / 17 Fisher κ value 6.9 16 2.0 Minimum size (x min ) 10 m 10 m 10 m Exponent (b) 3.5 3.72 3.66 Fracture intensity Transmissivity P32 (m 2 /m 3 ) Depth 0 50 m 0.5236 0.1215 0.2898 Truncated lognormal distribution Mean(Log 10 (T)) / Std(Log10(T)) T [1.0 10 9, 2.5 10 5 ] m 2 /s Depth 50 100 m 0.2689 0.0635 0.1495 Depth 100 200 m 0.0935 0.0215 0.0517 Depth 200 400 m 0.0567 0.0127 0.0314 Depth > 400 m 0.0573 0.0133 0.0311 Depth 0 50 m 7.95 / 1.09 Depth 50 100 m 8.08 / 1.00 Depth 100 200 m 8.38 / 1.09 Depth 200 400 m 8.38 / 1.09 Depth > 400 m 8.38 / 1.09 7.95 / 1.09 8.08 / 1.00 8.38 / 1.09 8.38 / 1.09 8.38 / 1.09 7.95 / 1.09 8.08 / 1.00 8.38 / 1.09 8.38 / 1.09 8.38 / 1.09

42 4.2.2 Boundary conditions The scope of the solute transport analysis was to assess hydrodynamic control of retention under natural flow conditions and after saturation of the repository. Hydrological conditions at Olkiluoto will slowly evolve due to the post-glacial land up-lift 5. The gradual change of hydrological conditions was taken into account by repeating the calculations as steady state simulations for a time soon after the saturation of the repository (1 000 years AP) and for a time when the deflection caused by the last glaciation will be fully recovered (10 000 years AP). The top of the model was set to a fixed hydraulic head that followed the estimated level of the water table at 1 000 years and 10 000 years after present (Figure 4-12). No-flow conditions were applied on all the other boundaries of the model. Figure 4-12. Fixed-head boundary condition at the top of the DFN model (the sea level) followed the estimated levels of the water table at 1000 years and 10 000 years after present. The boundaries of the hydro-dfn model are indicated by a black rectangular and the present-day shoreline of Olkiluoto by a gray line. 4.2.3 Hydrodynamic control of retention The main scope of the DFN simulations is to estimate flow related retention properties along potential radionuclide release paths. Hydrodynamic control of retention can be represented by a single parameter in case of sorption and matrix diffusion as the principal retention processes. This parameter has been notated in many different ways: β (e.g. Cvetkovic et al., 1999), FWS/q (e.g. Neretnieks, 2002), F quotient (Andersson et al., 1998) and WL/Q (e.g. Vieno & Nordman, 2000). The first three notations are fully interchangeable, but WL/Q = F/2. In the present analysis hydrodynamic control of retention is represented by WL/Q, which is derived from F quotient values calculated by ConnectFlow. WL/Q describes integrated flow properties along the transport channel as given in Equation 4-1. 5 Conditions will also change at later times due to permafrost and glaciation effects.

43, (4-1) where L is the length of the flow path, Q is the flow rate along the channel and W(s) is the width of the channel at location s along the channel. In ConnectFlow, the F quotient is calculated along particle-tracked flow paths. The F quotient for a flow path is approximated by summing over fractures along the flow path as shown in Equation 4-2., (4-2) where t f is the travel time over a fracture and e f is the transport aperture of a fracture. The transport aperture is calculated from the hydraulic aperture by applying an aperture factor. The transmissivity of a fracture depends on the friction along the flow path, but solute transport depends on the volume of the transport channels. In the variable aperture fracture this means that the smallest apertures will determine hydraulic properties, but large apertures will contribute to the volume of the fracture. This means that the effective hydraulic aperture is smaller than the effective transport aperture, typically the aperture factor is between 5 and 20 (Hautojärvi and Taivassalo, 1994). The present analysis applies aperture factor of four, i.e. transport aperture is four times larger than hydraulic aperture. However, it is noted that the hydrodynamic control of retention (WL/Q) does not depend on the transport aperture and the selected value of the aperture factor is not crucial for the main results of the present analysis. 4.2.4 Model set-up The size of the DFN model extends well beyond the main lineaments that border the Olkiluoto island. The model covers a domain that goes from the top surface (z = 0) to a depth of 1 000 m and is located between east [3000, 9900] and north [250, 4700] (cf. Figure 4-12). The starting locations of migration paths are divided into six separate panels. Figure 4-13 shows a map of Olkiluoto and six groups of flow path starting locations. The three western panels are located in the WCA, below the major hydro-structure HZ20. The three eastern panels are located outside of the WCA, above the structure HZ20. All the starting locations are placed 420 m below ground surface. The size range of the WCF fractures included in the model has been adjusted to keep the hydro-dfn model computationally reasonable. The full range of WCF fractures, i.e. sizes from 10 m to 500 m, are included in the model near the deposition holes as shown in Table 4-5. There is a domain of small background fractures within a radius of at least about 100 m around each starting location where the smallest or least transmissive WCF fractures are truncated from the model. An example of the different DFN hydrostructures and model sub-domains for the repository panel 2 is presented in Figure 4-14. Numerical flow simulations were carried out using the ConnectFlow code (Serco Assurance, 2005). The calculations were performed as steady-state simulations, without

44 considering density-dependent flow 6. The flow paths were simulated by particle tracking, "releasing" ten particles from each starting locations. Well mixed conditions are assumed to prevail in the fracture intersections. Ten particles are thought to give a good description of the flow paths of the highest flow rates. The simulations were repeated for 21 realisations of stochastic LHZs and WCFs, in order to develop a good impression of the stochastic variability between the realisations. Figure 4-13. Starting locations (black dots) of the flow paths are grouped to six different panels. The panels 1, 2 and 5 are located below the major hydro-structure HZ20 and the panels 3, 4 and 6 are above HZ20. The preliminary repository tunnel design is presented with green lines (Kirkkomäki, 2006). 6 The recent version of ConnectFlow can handle density variations.

45 Figure 4-14. An illustration of the hydraulic features of the DFN model and the repository near-field model for the flow paths starting from the panel 2. The WCF model is constructed using five different depth layers, although they are not clearly visible in the figure. The fractures are coloured by transmissivity using the same scale in all the figures.

46 Table 4-5. Summary of the model sub-division to domains that have different fracturesize and transmissivity cut-offs. Near-field of the repository panel 1 Model sub-volume East North Depth WCF Fracture size (m) Fracture size and Transmissivity WCF Fracture transmissivity (m 2 /s) [5029, 5859] [1546, 2301] [-600, -200] r [10, 500] T [1E-9, 2.5E-5] Near-field of the repository panel 2 Near-field of the repository panel 3 Near-field of the repository panel 4 Near-field of the repository panel 5 Near-field of the repository panel 6 Depth layer [-1000, -400] Depth layer [-400, -200] Depth layer [-200, -100] [5604, 6433] [2333, 3087] [-600, -200] r [10, 500] T [1E-9, 2.5E-5] [6573, 7403] [1861, 2616] [-600, -200] r [10, 500] T [1E-9, 2.5E-5] [6136, 6965] [1215, 1969] [-600, -200] r [10, 500] T [1E-9, 2.5E-5] [5189, 5836] [2141, 2643] [-600, -200] r [10, 500] T [1E-9, 2.5E-5] [7345, 8037] [1233, 1963] [-600, -200] r [10, 500] T [1E-9, 2.5E-5] [3000, 9900] outside repository near-field [3000, 9900] outside repository near-field [250, 4700] outside repository near-field [250, 4700] outside repository near-field [-1000, -400] r [60, 500] T [1E-9, 2.5E-5] [-400, -200] r [60, 500] T [1E-9, 2.5E-5] [3000, 9900] [250, 4700] [-200, -100] r [100, 500] T [1E-8, 2.5E-5] Depth layer [-100, -50] [3000, 9900] [250, 4700] [-100, -50] r [250, 500] T [1E-8, 2.5E-5] Depth layer [-50, 0] [3000, 9900] [250, 4700] [-50, 0] r [250, 500] T [1E-8, 2.5E-5] 4.2.5 Flow paths at 1 000 years and 10 000 years after present The geometry of the flow paths starting from each panel were studied by particle tracking for the model-top boundary conditions representing the water table both at 1 000 years and at 10 000 years after present. Simulation results for four realisations are shown in Figure 4-15 and Figure 4-16. The geometry of the flow paths does not vary much between the realisations. The discharge locations are governed by the lineaments

47 that surround Olkiluoto. There is only a little mixing between the flow paths starting from different panels. The discharge locations at the ground surface disperse over a wide area, although they are originated in a very limited area of the starting points at the repository level. Changes in the water table caused by land up-lift do not influence on the geometry of flow paths. The flow path simulations for 1 000 years and 10 000 years after present give identical results (Figure 4-15 and Figure 4-16). In general, the geometry of the flow paths is also quite similar in different realisations. Flow paths of least retention (WL/Q less than 10 5 a/m) are characterised by short hydraulic connections to the major lineaments (Figure 4-17). Figure 4-17 shows also that there are fewer low-retention flow paths from the WCA (panels 1, 2 and 5) than from outside the WCA.

48 Figure 4-15. The flow paths for four realisations at 1 000 years AP. Flow paths in different realisations are indicated by different colours.

49 Figure 4-16. The flow paths for four realisations at 10 000 years AP. Flow paths in different realisations are indicated by different colours.

50 Figure 4-17. The flow paths of WL/Q less than 10 5 a/m from all the 21 realisations. Simulations are calculated for 1 000 years AP. 4.2.6 Retention takes place in the near field of the repository The statistics of the flow-related retention property WL/Q show smaller values for the paths starting from the eastern panels than from the western ones (Figure 4-18). This is natural because the locations of the western panels follow the preliminary repository layout that is adapted to the hydro-structural model of the known deterministic zones. Starting locations in the eastern panels are randomly located in the middle of the stochastic network of LHZs. This provides well conducting transport routes to the lineaments.

51 The background fractures (WCF) provide most of the overall retention along the path. Figure 4-19 shows distributions of the WL/Q in the different categories of hydraulic structures (MHZ, LHZ and WCF) along the flow paths. In all cases the WL/Q values along MHZs and LHZs are only a small fraction of the WL/Q along WCFs. Retention in the stochastic hydro-features does not depend on the starting location. The deterministic features (MHZ) show a little bit more retention for the flow paths starting from the WCA than for those starting outside the WCA. The reason is that the high transmissivity lineaments are the only MHZs outside the WCA. Again, there is no difference between the simulations for 1 000 years and 10 000 years after present. Figure 4-20 shows the accumulation of WL/Q along the flow paths in all the realisations. All the paths show strong retention at the beginning of the flow paths. The great majority of the flow paths reach the level of WL/Q = 10 5 a/m in first few tens of metres. Figure 4-20 shows that there are more low-retention flow paths starting from the eastern panels than from the western ones, as has already been observed from Figure 4-18. The distributions of the WL/Q for the flow paths in Figure 4-18 do not directly describe the WL/Q distributions of the release paths from the deposition holes. In reality, radionuclide release points may locate between the described fractures. In those cases the starting point of the migration path in the model is moved to the nearest fracture. The given WL/Q distributions shown do not include in the starting points such deposition holes that are not intersected by water-conducting flow paths, i.e. the canister locations that have an infinite WL/Q. The number of "dry" deposition holes can be estimated based on the average fracture frequency along (more or less) vertical boreholes. The proportion of "dry" deposition holes can then be used to scale given WL/Q distributions to better represent all deposition holes. The frequency of PFL fractures below a depth of 400 m is about 0.06 fractures/m. Uniform fracture frequency along a vertical line leads to the Poisson distributed number of fractures for a fixed length interval. Deposition holes are approximately 8 m long, i.e. the expected number of fractures in a deposition hole is 8 m 0.06 fractures/m = 0.48 fractures and according to the Poisson distribution the probability for zero fractures in a deposition hole is exp(-0.48) = 62 %.

52 Figure 4-18. The distribution of the log 10 (WL/Q [a/m]) in all the 21 realisations from the various panels at 1 000 years AP. The upper row includes the western panels 1, 2 and 5 that are located below HZ20, and the lower row includes the eastern panels 3, 4 and 6 above HZ20. The vertical black lines indicate WL/Q of 5 000 a/m and the vertical yellow lines WL/Q of 50 000 a/m. Figure 4-19. The distributions of the log 10 (WL/Q [a/m]) in all the 21 realisations for MHZ (light blue), stochastic LHZ (red) and WCF structures (green and blue) at 1 000 years AP. The upper row includes the western panels 1, 2 and 5 that are located below HZ20, and the lower row includes the eastern panels 3, 4 and 6 above HZ20. The distribution for MHZ include all the deterministic structures, i.e. also the LHZs in the WCA. The vertical black lines indicate WL/Q of 5 000 a/m and the vertical yellow lines WL/Q of 50 000 a/m.

53 Figure 4-20. The accumulation of WL/Q in all the 21 realisations at 1 000 years AP. The upper row includes the western panels 1, 2 and 5 that are located below HZ20, and the lower row includes eastern panels 3, 4 and 6 above HZ20. The horizontal black lines indicate WL/Q of 5 000 a/m and the horizontal yellow lines WL/Q of 50 000 a/m. 4.2.7 Large variability in flow-related retention properties The hydrodynamic control of retention varies quite significantly between the realisations. There is an order of magnitude variability in the 90 % confidence interval of the median WL/Q between the realisations. The variability is a couple of order of magnitudes for the flow paths of least retention. Figure 4-21 shows statistical measures between the realisations for all panels, after the omission of the dry deposition holes. It demonstrates that there is no difference in hydrodynamic control of retention between the chosen times (1000 years or 10 000 years AP), but there is a significant difference in retention between the realisations and some difference between the panels. Simulations indicate that there is a 50 % probability that the minimum WL/Q is at least 50 000 a/m in the WCA and over about 20 000 a/m outside WCA. However, especially for the southern panels there is an about 5 % probability that the minimum WL/Q is only about 2 000 3 000 a/m. The statistics on the flow rates passing deposition holes (Figure 4-22) show that the largest flow rates in the realisations are a few tens of litres per year in the WCA panels and some hundred of litres per year outside the WCA. Such deposition holes have been omitted from the statistics, that are not intersected by conductive fractures.

54 1000 years AP 10 000 years AP Figure 4-21. The statistics of the simulated WL/Q for the flow paths starting from various panels (only for the holes intersected by conducting fractures). The colours indicate the statistical measure used: minimum (red), 5 % percentile (gray), median (blue) and 95 % percentile (green). The markers indicate median values and the lines indicate the 90 % confidence intervals of the simulated 21 realisations. The vertical black lines indicate WL/Q of 5 000 a/m and the vertical yellow lines WL/Q of 50 000 a/m. 1000 years AP 10 000 years AP Figure 4-22. The statistics of the simulated flow rates passing deposition holes (only for the holes intersected by conducting fractures). The colours indicate the statistical measure used: minimum (red), 5 % percentile (gray), median (blue) and 95 % percentile (green). The markers indicate median values and the lines indicate the 90 % confidence intervals of the simulated 21 realisations. 4.2.8 Low retention is coupled with high flow rate at the deposition hole The correlation between the hydrodynamic control of retention (WL/Q) along a flow path and flow conditions at the deposition hole were studied by analysing particle tracking results from all the 21 DFN realisations. Each starting location of the flow path was "sampled" with 10 particles. The flow-related retention property connected to a given starting location is the minimum WL/Q among the 10 sampled paths. All starting

55 locations are positioned in water-conducting fractures, i.e. dry deposition holes are not included to analysis. There is a clear correlation between the flow rate at the beginning of the flow path and the corresponding WL/Q of the entire path. The correlation between the transmissivity of the first fracture of a path and the WL/Q of the entire path is not strong in the present simulations partly because the used DFN model version does not include correlation between fracture size and transmissivity. However, including a fracture size-transmissivity correlation would not affect the observed correlation between WL/Q of the flow path and flow rate at the beginning of the path, because that is strongly coupled to the essential character of WL/Q. Figure 4-23 contains scatter plots of the WL/Q values of flow paths as a function of the flow rate at the beginning of the flow path for simulations at 1 000 years AP. The results for 10 000 years AP are practically identical with the results for 1 000 years AP. Each marker is coloured based on transmissivity of the first fracture along the path. The feature type of the first fracture is indicated by the marker: the WCFs are plotted with plain dots and the LHZs with circled dots. The tailing of the low-retention flow paths extends to lower WL/Q outside the WCA than inside it. However, the smallest values of WL/Q are connected with the highest flow rates in a range of hundreds of litres per year. In addition, all their cases involve an intersection between a LHZ and a deposition hole. In the WCA, a flow rate of less than 1 L/a will provide hydrodynamic control of retention that is equivalent to WL/Q = 50 000 a/m or more. Figure 4-23. WL/Q vs. flow rate at 1 000 years AP in 21 realisations. The WL/Q values represent entire flow paths and the flow rates only the first fracture of each path. The upper row includes the western panels 1, 2 and 5 that are located below HZ20, and the lower row includes eastern Panels 3, 4 and 6 above HZ20. The flow paths start either from a background fracture (dots) or from a LHZ (dot and a circle). The markers are coloured by the transmissivity (m 2 /s) of the first fracture.

56 4.2.9 Conclusions on the DFN analysis Simulations show that there is a rather large variability in the hydrodynamic control of retention of the flow paths between the DFN realisations. This variability is thought to follow from the nature of the retention along the flow path, where a single fracture can make a great difference to the WL/Q of the whole flow path. This conclusion is also supported by properties of the flow paths from the different panels. Flow paths starting from the WCA tend to have more retention. The main difference between flow paths starting from the WCA and outside the WCA is that large structures are avoided in the WCA. Evolution of the water table due to the post-glacial land up-lift and corresponding changes in the top boundary condition does not have a great influence on the retention or flow properties. These simulations indicate that in most of the realisations the minimum WL/Q is at least 50 000 a/m in the WCA and at least 20 000 a/m outside the WCA. There is a probability of a few percent that the minimum WL/Q for a flow path is only about 2 000 3 000 a/m. However, flow paths of very low WL/Q can be avoided, because simulations indicate that the flow rate in a deposition hole has a good correlation with the hydrodynamic control of retention along the flow path and that the smallest values of WL/Q are always connected with an intersection of a LHZ and a deposition hole. Inside the WCA, a flow rate of less than 1 L/a suggests a hydrodynamic control of retention of at least WL/Q = 50 000 a/m. 4.3 Consistency of the EPM and DFN results 4.3.1 Local and site-scale flow properties of the hydro-dfn model Local and site-scale flow properties of the hydro-dfn model are calculated in order to evaluate consistency of the DFN and EPM flow model representations of the Olkiluoto site. The EPM simulations include site-specific processes and features that may affect the groundwater flow but are not included to the DFN flow analysis. Features that are included in the EPM simulations but not in the DFN modelling include e.g. the repository and disposal tunnels, spatially varying groundwater salinity- and density-dependent flow, and effects of the heat produced in the radioactive waste. Block-size effective hydraulic conductivities Block-size effective permeabilities are estimated in the hydro-dfn model by applying cellular model calculations in ConnectFlow. The sub-volume of the hydro-dfn model that is bounded by east [4000, 6000], north [1000, 3000] and depth [-1000, 0] is divided into cubic blocks of 100 metre edge length. Block-size conductivities are calculated from 20 hydro-dfn realisations for the following two cases: 1) taking into account only water-conducting background (WCF) fractures and 2) considering all hydraulic fractures, i.e. WCF, LHZ and MHZ.

57 Table 4-6 summarises the calculated block-size conductivities for the depth layers of the DFN model. Conductivities decrease below 300 m following the applied depth-dependence of fracture transmissivities. The difference in the 100 m block-size conductivities between the top and bottom of the model is almost an order of magnitude when all hydraulic features are considered. The deterministic zones and stochastic LHZ increase the average hydraulic conductivity by a factor of four to six compared with the model that contains only WCFs (Table 4-6). They also have a strong influence on the number of conductive blocks (i.e. blocks containing at least one hydrogeological zone or waterconducting fracture). With MHZ and LHZ, the percentage of the conductive blocks is 88 % as compared with 62 % when only WCFs are included in the model. Figure 4-24 shows distributions of the block-size conductivities. The hydraulic conductivity in the EPM model is clearly higher at the very top of the model. However, at the repository depth both models apply similar block-size conductivities. Table 4-6. Block-size conductivities in a 100 m scale, simulated using the hydro-dfn model. k11, k22 and k33 are the diagonal values of hydraulic conductivity tensors. Standard deviations are bracketed. WCF only, percentage of conductive blocks 62 % Depth log 10(k11 [m/s]) log 10(k11 [m/s]) log 10(k11 [m/s]) 0-100 m -9.6 (1.1) -9.6 (1.1) -9.6 (1.1) 100-200 m -9.7 (1.1) -9.7 (1.1) -9.9 (1.1) 200-300 m -9.8 (1.1) -9.7 (1.1) -10.0 (1.2) 300-400 m -9.8 (1.1) -9.7 (1.1) -9.9 (1.1) 400-500 m -9.7 (1.3) -9.6 (1.2) -9.9 (1.3) 500-600 m -9.7 (1.2) -9.6 (1.2) -9.9 (1.2) 600-700 m -9.8 (1.1) -9.8 (1.1) -10.0 (1.1) 700-800 m -9.8 (1.1) -9.8 (1.1) -10.0 (1.2) 800-900 m -9.8 (1.1) -9.7 (1.0) -10.0 (1.1) 900-1000 m -9.9 (1.2) -9.8 (1.2) -10.0 (1.2) All layers -9.8 (1.2) -9.7 (1.1) -10.0 (1.2) All hydro-features, percentage of conductive blocks 88 % Depth log 10(k11 [m/s]) log 10(k11 [m/s]) log 10(k11 [m/s]) 0-100 m -8.5 (1.1) -8.5 (1.0) -9.1 (1.0) 100-200 m -8.7 (1.2) -8.7 (1.2) -9.1 (1.1) 200-300 m -8.8 (1.2) -8.9 (1.2) -9.2 (1.2) 300-400 m -8.9 (1.2) -9.0 (1.1) -9.2 (1.1) 400-500 m -9.0 (1.1) -9.0 (1.1) -9.3 (1.1) 500-600 m -9.1 (1.2) -9.0 (1.2) -9.4 (1.2) 600-700 m -9.2 (1.2) -9.2 (1.1) -9.5 (1.2) 700-800 m -9.3 (1.3) -9.3 (1.2) -9.6 (1.2) 800-900 m -9.3 (1.2) -9.3 (1.1) -9.6 (1.2) 900-1000 m -9.4 (1.2) -9.4 (1.1) -9.7 (1.2) All layers -9.0 (1.2) -9.0 (1.2) -9.4 (1.2)

58 Figure 4-24. The distribution of the 100 m blocks-size conductivities based on 20 realisations of the hydro-dfn model. The blocks-size conductivities were calculated 1) separately for the WCF only (dashed lines), and 2) including all the hydraulic features (solid lines). Flow rates at the top of the model Flow rates were calculated through a control plane that is located at the top of the model and covers the main part of Olkiluoto (Figure 4-25). The control plane is bounded by the coordinates east [3200, 7000] and north [1150, 4000]. It gives an estimate of the groundwater infiltration to the model. Control plane is smaller than the modelling volume and discharge areas are mainly outside the control plane. Therefore, inflow through the control plane can be larger than outflow. The simulations were performed for the water table at 1 000 years AP. The results are consistent with the differences in hydraulic conductivities of the model top layers in the DFN and EPM representations. The base case calculations show by a factor of 2 4 higher inflow to the EPM model than to the DFN model. Applying the block-size conductivity of the DFN model in the top layers of the EPM model gives consistent results with DFN simulations (Figure 4-26). The maximum inflow is about twice the minimum inflow among the 20 DFN realisations. This indicates that the equivalent hydraulic conductivity varies significantly at the scale of the approximately 10 km 2 control plane.

59 Figure 4-25. The control plane on the top surface of the model was used to estimate inflows to the model. Figure 4-26. The distribution of the inflow and outflow rates through the control plane on the ground surface at 1 000 years AP. The DFN results are based on 20 realisations of the hydro-dfn model. The red markers indicate median values and the blue lines indicate 90 % confidence intervals. The EPM results were calculated using base-case hydraulic conductivities (green) and the DFN model was used to simulate block-scale conductivities (black).

60 Groundwater flow through a repository panel The central area of the repository is enclosed by a control box that measures horizontally 778 m x 778 m and is vertically 100 m thick (between -470 m and -370 m). The location of the control box is illustrated in Figure 4-27. The control box was placed in a volume that is located centrally in relation to the disposal tunnels and has been characterized better than other bedrock volumes. The DFN simulations are based on 20 realisations of the hydro-dfn model. Turnover of the groundwater flow through the repository box shows half an order of magnitude variability between those 20 realisations. The minimum flow rates are less than 20 m 3 /a and the highest flow rates around 100 m 3 /a. The corresponding EPM simulations give flow rates that are in the same order of magnitude but in the lower range of the DFN simulation results (Figure 4-28). The variability in the DFN results is larger because in some realisations stochastic LHZs intersect the control box. Figure 4-27. The flow rates are calculated through a repository panel that is indicated by the blue line. The repository layout is based on a preliminary plan by Kirkkomäki (2006).

61 Figure 4-28. The groundwater flow rate through a repository panel. The DFN results are based on 20 realisations simulated for 1 000 years AP. The red marker indicates median values and the blue lines indicate 90 % confidence interval. The EPM results were calculated using DFN block-size conductivities (magenta), for 10 000 years AP (green) and for 1 000 years AP (black). The EPM simulations were performed assigning a conductivity of 9 10 11 m/s for the backfilled tunnels and excluding heat production in waste canisters. 4.3.2 Conclusions on the consistency of EPM and DFN analyses The EPM and DFN modelling gave consistent results regarding the flow paths and flow rates at the repository level. There is some discrepancy in the flow rates at the top of the model but that is due to differences in the effective hydraulic properties in the top layers. The effective 100 m block-scale conductivities simulated in the DFN model show similar conductivities at the repository level to those that were applied in the EPM modelling. The EPM model is more conductive close to the ground surface than the DFN model. Hence, the EPM model gives larger infiltration than the DFN model. However, at the repository level the DFN model gives slightly higher flow rates. This is consistent with the differences in the block-scale conductivities between these two models. The flow paths from the six repository panel areas are very similar in both models (cf. Figures 4-7 and 4-8 for EPM in Löfman and Poteri [2008] to Figure 4-13 for DFN in the present report). Dominant flow directions in both simulations are towards the major lineaments that surround Olkiluoto. The flow paths simulated in the DFN model are more dispersed than in the EPM model due to the explicit representation of the stochastic background fracturing.

62 The EPM modelling shows that the influence of the decay heat produced in the waste canisters on the flow field and flow paths is negligible already at 1 000 years after the repository closure. The influence of the repository and disposal tunnels on the flow field in the repository panels was analysed using the EPM modelling. The results indicate that the hydraulic conductivity of the tunnel backfill and surrounding EDZ has an influence on the flow paths inside the repository and in the near field of the repository. Flow paths in the far field have not been analysed for different repository tunnel variants. The lineaments are quite dominant discharge pathways to the ground surface in all the analysed cases for both EPM and DFN simulations. However, it cannot be completely ruled out that any other discharge location is activated for some variants of the repository tunnel backfill or EDZ. Substantial changes in the flow pattern between the tunnels, EDZ and background fractures may then also affect hydrodynamic control of retention of the flow paths. The relative hydraulic properties between the porous tunnel backfill and the surrounding fracture network affect the distribution of groundwater flow in the bedrock and the tunnels. If the properties of the structures are equivalently close to each other, the tunnels can cause just marginal effects in the flow pattern. On the other hand, if the tunnel backfill is more conductive than the fracture network, the flow rates along tunnels are higher and the flow rates in adjoining fractures get reduced. However, this is thought not to have remarkable effects on the WL/Q integrated over the release paths along the fracture network. ConnectFlow is able to handle combinations of porous media and fracture network and enables further analyses of combined systems of tunnels and fractured rock in the future. Currently, there is lack of data on the EDZ transmissivities, but the matter will be handled in coming analyses. Uncertainties in the data can be covered by scoping calculations.

63 5 RADIONUCLIDE RELEASE AND TRANSPORT Posiva's safety case is consisted of 1) a main scenario, where no releases of radionuclides will occur within safety-relevant period of time and thus no calculation cases are needed, and 2) assessment scenarios, where radionuclides are assumed to be released (Miller & Marcos, 2007). Release from canisters is assumed to occur in all the assessment scenarios of RNT-2008 meaning that the main scenario of Posiva's safety case (Pastina & Hellä, 2006) is omitted in this analysis, although there is good evidence on the ability of fuel canisters to prevent release of radionuclides in the long-term. It assumes that the canisters will retain their containment capacity through the whole safety-relevant period. Alone the design requirement for the minimum corrosion-limited lifetime of canisters in the expected repository conditions is 100 000 years (Posiva, 2006), while the corrosion studies suggest significantly longer lifetimes (Chapter 8.1.4 in Pastina & Hellä, 2006). Nevertheless, it cannot be completely ruled out that there could be initially defective canisters or that canisters will fail later, no matter how unlikely such situations may be. The processes leading to radionuclide release and transport have been thoroughly described in Miller & Marcos (2007) and only short descriptions are given in this report. 5.1 Assessment scenarios and modelling concepts In this assessment several scenarios have been formulated to cover the possible futures of the repository system. In each scenario a general setting is defined with the system's evolution and becomes further developed into calculation cases. The features of the calculation cases are then used in deriving the input for the modelling. A KBS-3V spent fuel repository is based on a multi-barrier system to ensure that if any of the barriers fail, the others will still mitigate the release of radionuclides to the biosphere. The long-term performance of the barriers has been evaluated in predicting the evolution of the repository (Pastina & Hellä, 2006) and it is regarded very likely that the multi-barrier system will effectively hinder the release of radionuclides into the host rock, ensuring the safety of the repository for at least a million years. This is in accordance with Section 8 of the Government Decision STATE 398/91 (Council of State, 1991). However, in view of the management of uncertainties and in accordance with the regulatory requirements (STUK, 2001) to analyse the behaviour of the system with respect to possible releases of radionuclides, several assessment scenarios (see Chapter 2 in Miller & Marcos, 2007) are defined in which either one or more barriers are assumed to fail or no credit is given to their safety function. The assessment scenarios are listed in Table 5-1. These scenarios cover most of the Features, Events and Processes (FEPs) described in the Process Report (Miller & Marcos, 2007). In the defective canister scenario (DCS), it is considered that the con-

64 tainment capacity of the canister fails as a consequence of a non-detected initial penetrating defect (DCS-II) or as a consequence of corrosion (DCS-I) in cases where the canister presents other defects (e.g. thin wall, non-penetrating cracks, etc.). In the additional scenarios (AD), the canister containment capacity is assumed to fail due to either internally occurring processes or external events, that is, in AD-I, the canister fails as a consequence of a rock shear displacement. The assumed magnitude of the event depends on the time at which it occurs, since in this respect significant rock shears are much less likely before the next ice age than shortly after it. In AD-II, accelerated corrosion damages the canister as a consequence of events affecting the buffer. Such events can be initial poor emplacement of buffer and erosion of buffer material due to intrusion of glacial meltwater. In AD-III, it is considered that in a canister with a penetrating defect gas is generated due to corrosion of the iron insert or other metal parts. The gas then escapes being able to carry radionuclides that exist in gaseous form. Gas may also displace water that is contaminated by radionuclides originating mostly in the instant release fraction (IRF) of spent fuel. In AD-III, no credit is claimed for the retardation capacity of the buffer and backfill. The calculation cases derived from these scenarios are described in Chapter 6. Sensitivity cases and what if cases addressing uncertainties in the knowledge of processes, their magnitude, and their predictability are built starting from these scenarios. Supplementary calculation cases, that are used to scrutinise the robustness of the disposal system without specific reference to detrimental events and processes supported by scientific reasoning, are also presented in Chapter 6. Table 5-1. Assessment Scenarios in Posiva s Safety Case for a KBS-3V repository (after Miller & Marcos, 2007). Assessment Scenarios Defective canister scenario (DCS) Additional scenarios (AD) based on deviations in initial conditions and timing of internal or external processes Human intrusion scenario (HI) * Descriptions and divisions into variants DCS-I: delayed penetrating defect radionuclide release starting at 10 000 years after the repository closure DCS-II: early penetrating defect groundwater in contact with spent fuel at the repository closure AD-I: earthquake / rock shear: a canister fails as a consequence of the sudden displacement of a fracture intersecting a deposition hole AD-II: a canister fails as a consequence of disruptive events affecting the buffer: e.g. due to misemplacement of buffer or intrusion of diluted meltwater AD-III: gas expels water contaminated with IRF or radionuclides in volatile form (C-14) from a canister and deposition hole. The buffer and backfill are assumed to be incapable to retain radionuclides. HI-I: a deep water well is bored at the disposal site HI-II: core-drilling penetrates the wall of a canister * The human intrusion scenario is to be dealt with in the Biosphere Assessment of the Safety Case.

65 5.1.1 Radionuclide inventory The radionuclide inventory in spent nuclear fuel depends among other on the burnup history of the fuel, and the time elapsed since the end of its irradiation. The highest burnup to date in Finland is around 45 MWd/kgU. The highest burnup averaged over a fuelling batch will be 38 MWd/kgU in Loviisa fuels and 40 MWd/kgU in Olkiluoto fuels, and the highest local burnups within a fuel assembly may be as high as 48 and 52 MWd/kgU, respectively. At the planned sealing time of the repository, in the year 2 100, the average cooling time of the Olkiluoto fuel will be well over 30 years, but for the release and transport analysis in this report a cooling time of 30 years and the corresponding activity inventory was selected conservatively. The radionuclides are embedded within the fuel matrix, including its microstructures, within the gap between the fuel pellets and the cladding, and within the cladding and other metal parts. In the fuel matrix the radionuclide inventory is partitioned between a very immobile fraction in the ceramic granules of mixed fissionable and fertile materials (principally UO 2 ), and a much more mobile fraction in grain boundaries, fractures and the rim zone of restructured fuel matrix. Radioactive gases are found in the before mentioned gaps in small amounts but in readily released form. In the metallic parts of fuel elements there are neutron activation products, caused by neutron radiation. The radionuclide inventory in the gap zone and grain boundaries forms a rapid source term of segregated nuclides, and is termed instant release fraction or IRF. See also Chapter 5.1.2. The material compositions of typical Olkiluoto and Loviisa fuel assemblies are presented in Table 5-2 (Anttila, 2005). Figures and dimensions for the Olkiluoto and Loviisa fuel assemblies are presented in TVO-92 (Vieno et al., 1992), TILA-96 (Vieno & Nordman, 1996) and Pastina & Hellä (2006). The nuclear fuel types taken into account in this report are BWR 7 and EPR 8 in the case of Olkiluoto NPP and VVER-440 9 in the case of Loviisa NPP. The activity inventories, heat generation and other radioactive properties of Olkiluoto and Loviisa fuel assemblies with different burnups, void history 10, and enrichment are presented in Anttila (2005) and concisely in Pastina & Hellä (2006). 7 Boiling Water Reactor, the reactor category of the power units OL1 and OL2. 8 European Pressurised (water) Reactor 9 VVER-440 is a pressurised water reactor of Soviet design with nominal capacity of 440 MW e. 10 In an operating boiling water reactor, voids (steam bubbles) in the reactor coolant affect the local fuel temperature and power distribution through differences in neutron moderation.

66 The spent fuels considered in the release and transport analysis were specified to have a uniform burnup of 40 MWd/kgU and the currently highest enrichment level. The activity inventories of the reference spent fuels are shown in Chapter 6.2 (Table 6-8 Table 6-10). Table 5-2. Material composition of fuel bundles (per 1 000 kg U) for each spent fuel type of Olkiluoto and Loviisa (after Anttila, 2005). Fuel bundle component BWR Mass (kg) VVER-440 Mass (kg) EPR Mass (kg) Fuel (U) 1000 1000 1000 Cladding (Zry-2 in BWR, ZrNb1 in VVER, M5 in EPR) Flow channel (Zry-4 in BWR, ZrNb2.5 in VVER) 265 342.62 282.55 190 112.31 Inconel parts 2.9 4.8 Stainless steel parts 53 194.30 50.90 5.1.2 Release of radionuclides from fuel Radionuclides can be released from fuel by a number of processes. Once in contact with groundwater, there will be rapid release of a certain fraction of the inventory at grain boundaries (IRF), followed by a slow congruent release from the fuel matrix by dissolution and alteration reactions. As the burnup of fuel increases, the structure of fuel pellets evolves to more altered states and towards increased reactive surface areas (Poinssot C. et al., 2006). Radionuclide release from fuel is mainly affected by the radiation intensity (depends on the inventory) and the volume of water in a canister. The radiation intensity controls the radiolysis of water and radiolytically generated oxidants can cause oxidative conversion and dissolution of the UO 2 matrix. The volume of water in the canister and the rate at which this volume turns over, are primary controls over whether or not radionuclides released to the groundwater reach their solubility limits, and do thus control the overall radionuclide release rate from a canister. The slow long-term release of radionuclides from fuel matrix is handled in this analysis by constant fuel degradation rates, leading to constant release rates of radionuclides from the matrix. Werme et al. (2004) reviewed the available data under the relevant reducing conditions inside a canister and proposed fractional degradation rates of 10-6 to 10-8 per year. For the purpose of this report the central value of 10-7 a -1 was used in most of the calculation cases, with a few exceptions, where 10-6 a -1 was applied. Once released from fuel, the transport behaviour of radionuclides is largely controlled by their solubility and speciation characteristics in the ambient hydrogeochemical environment.

67 The maximum possible volume of water in a canister is relatively small compared with the inventory of radionuclides. This combined with the very slow rate of water turnover in a canister (due to the diffusive barrier provided by the bentonite buffer) means that the concentration of many radionuclides in the water contained by a canister will be solubility limited (see Chapter 5.1.3). The composition of water in the canister interior (particularly the redox conditions) and temperature are major controls over radionuclide solubility and speciation, and thus also strongly affect the maximum release rates from fuel. 5.1.3 Solubility Solubility refers to the total aqueous concentration of an element in all dissolved chemical forms, which are in equilibrium with each other and with a pure crystalline or amorphous phase. If equilibrium is reached, a maximum concentration of all soluble species can be estimated with the help of chemical thermodynamics. In the solubility calculations for safety assessments, a solubility-limiting phase is assumed for each radioelement. The most probable solid phases expected to form under the existing chemical conditions are identified. The speciation calculations are important for the evaluation of solubility and transport properties of radionuclides. The results of chemical thermodynamic modelling are validated against available concentrations determined in natural environments and laboratories. Several assessments have been made on the radionuclide solubility limits relevant for the repository near field (Duro et al., 2006, Grivé et al., 2007). The solubility limits used in this report have been reported in Grivé et al. (2007; see Table 6-12 in this report). Radionuclide solubilities are dependent on the chemical environment in the canister cavity. Generally, the most important aquatic chemistry parameters that affect solubility are the concentrations of strong complexing ligands, e.g. carbonate content and ph. High-salinity groundwaters affect the solubility equilibria through their ionic strength. Additionally, the redox conditions in the water are relevant for elements with several oxidation states. In the case of e.g. uranium, reducing conditions in a canister due to the presence of hydrogen gas or corroding iron are the most important factors in limiting uranium solubility. The temperature affects the kinetics of dissolution/precipitation processes as well as the solubilities and complexes formed. The temperature has a strong effect on ph and hence on the solubilities of hydrolysable elements. Grivé et al. (2007) use, for the estimation of solubility limits, an Olkiluoto specific groundwater composition, taking into account the possible changes in groundwater composition with time and depth. Changes in temperature are not taken into account, which adds to the uncertainty in the solubility results. Radionuclides, which do not go into solution because of solubility constraints, can be directly incorporated into newly formed solid phases, which form by the alteration of spent fuel or engineered barrier materials, or from direct precipitation from solution. Alternatively, radionuclides may be sorbed onto surfaces, such as those of the iron oxyhydroxides formed by corrosion of the iron insert of a canister. Colloids may affect the transport of poorly soluble radionuclides due to sorption of radionuclides onto their

68 surfaces. However, these two last processes are omitted from the RNT-2008 calculations (see Chapter 5.2.3 in this report). Several elements (Ni, Se, Sr, Zr, Nb, Pd, Sn, and Mo) can be found as stable isotopes forming part of e.g. the fuel rods and bundles and are also generated by decay of some radionuclides. The amounts of these stable isotopes often determine whether or not the solubility limits are reached, as they tend to be larger than the amounts of radioactive isotopes of the same elements (Table 6-13). Because of the limitations of the REPCOM code, the solubility limits are applied in the calculation cases of this report in the near-field model only: inside the canister and at the buffer-rock interface, but not throughout the buffer. This restriction is of minor significance for release and transport calculation results. 5.2 Transport processes 5.2.1 Advection and Diffusion Advection and diffusion are major processes related to the migration of radionuclides in the context of this report. Diffusion is spontaneous net movement of particles from an area of high concentration towards an area of low concentration in a given volume of fluid, and advection is transport of dissolved substances with flowing water. They may occur in the engineered barrier system (EBS), in the geosphere, and at the interface between system components. Fission products generated in fuel will slowly migrate in the spent fuel matrix by two types of diffusion: thermally activated diffusion and athermal diffusion. Diffusion of radionuclides could, in principle, lead to elevated fission product concentrations at grain boundaries and an increase in the rapid release inventories when water eventually comes into contact with the fuel after a canister failure. Within the canister void space, aqueous diffusion and advection can transport radionuclides in aqueous solution. Movement into and out of the canister will be controlled by the hydraulic conductivity of the bentonite buffer that provides a diffusion barrier. Thus the rate of radionuclide movement into and out of the canister will be very slow, and the dominant process will be diffusion. In the buffer and backfill, advection is of importance soon after their emplacement, when the groundwater saturates the buffer and backfill from outside. However, at that time no radionuclides are expected to have been released from canisters to buffer or backfill. Advection may be important also in the case where dilute glacial meltwaters enter repository depths and damage bentonite (chemical erosion, see Chapter 5.2 in Miller & Marcos, 2007). After bentonite has generated adequate swelling pressure due to the uptake of water, transport will occur predominantly by diffusion. That is, advection is relevant soon after the emplacement of buffer and backfill (although radionuclides will not be released into the buffer before the saturation of bentonite) and

69 later on, after the melting of the next glacial ice-sheet (i.e. after 70 000 years; Pastina & Hellä, 2006), whereas diffusion is relevant for all time frames after bentonite saturation in the buffer and backfill. Diffusion in bentonite has been thoroughly studied in conjunction with radionuclide transport. Diffusion equations for radionuclides are described in detail for example in Yu and Neretnieks (1997) and in SKB (2006). Yu and Neretnieks (1997) also give a detailed discussion of different experimental methods to quantify diffusion and how the results can be interpreted. The diffusivity data for radionuclides in compacted bentonite, compiled for SR-Can (SKB, 2006: TR-06-25), is also used in this report (see Chapter 6.2). With respect to the geosphere, advection is a process by which solutes are transported by the bulk motion of flowing groundwater. In case of sparsely fractured hard crystalline rocks such as those at Olkiluoto advection takes place predominantly in a fracture network, and is driven by differences in hydrostatic pressure (the hydraulic gradient). At a small scale like centimetres and less, however, variations in the flow field coupled with mechanical mixing and molecular diffusion can result in local water flows and velocities that are different from the larger-scale average, resulting in the spreading of solute plume (Figure 5-1).

70 Location of fluid particles: at t at t+ t Contribution of different factors in hydronynamic dispersion Main direction of flow Figure 5-1. Hydrodynamic dispersion on a microscopic scale. An additional subprocess (not shown in the figure) also contributing to the spreading of tracers is matrix diffusion, which has been identified as a key retardation process in the geosphere (see Miller & Marcos, 2007). At a large scale (tens of metres or more), flow is concentrated in a small number of flowing features, typically formed by fracture zones or fracture intersections. On a smaller scale (centimetres), flow can be thought to be directed within a fracture via paths formed by interconnected void spaces, where fracture-filling minerals can add to the complexity of flow routes. Radionuclides and other substances dissolved in groundwater can be transported at the same rate with the advecting groundwater, although the concentration of radionuclides can be reduced through dispersion and dilution processes, sorption onto exposed mineral surfaces, and diffusion. Due to the heterogeneity of the fracture network and of the groundwater flow within it, realistic modelling of radionuclide transport is difficult. According to Darcy s Law, the groundwater flow rate through rock is controlled by the hydraulic gradient and the hydraulic conductivity, which in turn is a function of porosity and permeability, and the properties of the groundwater (density, viscosity etc). The heterogeneity in the system means that the actual hydraulic conductivity of rock and, hence, the velocity and direction of flow (and thus radionuclide transport) can vary widely throughout the rock mass. Groundwater flow models have thus been developed to account for this hetero-

71 geneity, and these are often based on discrete fracture network (DFN) models (Chapter 4.2). In fractured crystalline rocks, there are in addition to the above mentioned interconnected void spaces water-filled void spaces that are not interconnected, in which water is effectively stagnant and diffusion (matrix diffusion) becomes the dominant transport mechanism. Matrix diffusion is a process by which radionuclides and other species in the flowing groundwater migrate into the stagnant pores and microfractures of the surrounding rock mass (Neretnieks, 1980; Chapter 8.3.4 in Miller & Marcos, 2007). This is an important retardation process in fractured crystalline rocks, where there is interconnected porosity in the rock mass adjacent to the flowing fractures, into which diffusion can take place. Even for non-sorbing species, matrix diffusion may provide an efficient temporal retardation and spreading process simply by removing them from the flowing groundwater for some time. The volume of pore space in rock matrix available for the process and the connections between those pores control the actual significance of matrix diffusion on radionuclide transport in the geosphere. 5.2.2 Sorption In this context, sorption is a general term describing the attachment of dissolved species onto mineral surfaces. It includes ion exchange, physical adsorption and surface complexation. Sorption can also be considered as the precursor to precipitation. Sorption is element specific and depends on radionuclide aqueous speciation, composition of solution, and the composition and surface characteristics of solid phase. Sorption is a reversible reaction, by definition, and the long-term retardation mechanism is not well established. Radionuclides, initially sorbed at solid-liquid interfaces, such as those of iron corrosion products, may be subsequently released within a shorter time than the time, during which they were accumulated. The release could take place e.g., if the redox conditions or temperature change, or if the solid phase dissolves or undergoes other mineralogical transformation. One situation, where this is deemed possible, is linked to the intrusion of oxic glacial melt water deep into the bedrock. This could lead to a rapid change in otherwise stable chemical conditions. The long-term sorption behaviour under these circumstances is not well understood. The sorption of radionuclides on bentonite buffer is assumed to be in equilibrium (i.e. kinetics are fast compared with transport processes) quantified by the distribution coefficient (K d ) between the sorbed concentration as mass per weight of solid material and the concentration in solution. K d is strongly dependent on the chemical conditions and its experimentally determined values are valid only for the particular experimental conditions. K d values of radionuclides on bentonite have been thoroughly studied by many laboratories and databases have been compiled by the waste management agencies (e.g. Bradbury & Baeyens, 2003 for NAGRA and Ochs & Talerico, 2004 for SKB).

72 Sorption is included in the modelling of radionuclide transport for bentonite buffer, tunnel backfill and rock matrix in all the calculation cases. 5.2.3 Omitted processes There are known processes that can influence the migration of radionuclides but are omitted from the analysis. The omitting of them shall be defendable on the grounds of conservatism or insignificance. Sometimes these processes can be roughly incorporated by indirect ways. For example, colloidal transport has in some analyses been handled by decreasing the K d values of some nuclides in order to take into account the possibility that colloids could enhance the migration of radionuclides 11. Sorption (Chapter 5.2.2) and (co-)precipitation are two chemical mechanisms retarding or retaining radionuclides. In the RNT-2008 near-field transport calculations precipitation is taken into account by applying solubility limits inside the disposal canisters and at the buffer-rock interface. Radionuclide concentrations are usually expected to be low in repository conditions, but it must be noted that large amounts of stable isotopes of the same elements can compete for sorption sites with corresponding radionuclides and thus weaken the sorption of radionuclides. On the other hand, the co-precipitation of a radionuclide with a stable isotope would be an advantageous process, but co-precipitation of radionuclides requires formation of new solid phases through major element reactions and this is generally difficult to quantify. In all the assessment scenarios, sorption of radionuclides on iron corrosion products in the iron insert of a canister is conservatively omitted. Also co-precipitation of radionuclides alone or with existing or new mineral phases is omitted both in the EBS (iron and copper corrosion products, minerals in the buffer and backfill) and in the geosphere (calcite, iron sulphides, iron oxides). Radionuclides will initially precipitate as microcrystalline or amorphous solid phases, which may grow to form large crystals more resistant to dissolution. The new solid phases will be the more stable the more similar the major and the minor constituents are in charge and radii. For example Ra could form a stable co-precipitate with Ba in barite, and Sr with Ca in calcite. Nonetheless, precipitation and (co-)precipitation of radionuclides in bentonite/smectite and other mineral components has been reported in connection with many natural analogue studies and also at laboratory scale (e.g. Cramer & Smellie, 1994). In disposal systems, colloids may enhance the migration of radionuclides, provided that radionuclides become fixed to colloid particles moving at similar velocities as the groundwater does. Naturally occurring colloids can usually be neglected in performance assessments, based on their very low concentrations, but colloids formed in the repository near field may need to be considered (EUR, 2005). In Posiva's near-field transport calculations colloids in bentonite are omitted, since colloids are regarded to be immo- 11 Colloids can also slow down the migration.

73 bile in the small-scale pore structure of bentonite. In the far-field analysis colloidal transport is not included. According to the results of the RETROCK project (EUR, 2005) most of the processes omitted correspond with the highest level of uncertainty and lowest level of significance associated with radionuclide transport and retardation processes in the geosphere (Figure 5-2). Figure 5-2. The uncertainties in the treatment of retention and transport processes versus their relevance for performance assessments were assessed in the RETROCK project with this kind of graph. A difficulty in placing many of these processes in a simple graph stems from their site- and concept-specificity (EUR, 2005). 5.2.4 Impact of gas In the disposal canisters significant volumes of gas can be generated through the corrosion of the iron insert of a canister. If water enters a canister, a situation may later arise where contaminated water becomes displaced by gas and thus forced out of the canister. In another process, C-14, that is generated in fuel assemblies (fuel, zircaloys, stainless steel), can migrate in gaseous form together with corrosion-based carrier gas. Gas in the far field could affect groundwater flow paths and flow rates or accelerate transport through attachment of colloids at gas-water interfaces (EUR, 2005). At the Olkiluoto site there is naturally occurring gas dissolved in the groundwater, but no signs of considerable amounts of gas have been discovered. Gas-mediated migration can be important only if gas releases and radionuclide releases happen to overlap in time. The effect of gas in the far field is not included in the RNT-2008 analysis.

74 Any radionuclides that can partition into free gas phase (bubbles) may be released from a canister as a gas (e.g. C-14). However, unless the gas pressure exceeds the confining pressure of the bentonite buffer, gas release from the canister will be diffusion limited. Gas expulsion and formation of pathways may take place, if water gets into contact with the cast iron insert in a defective canister. At the time of canister emplacement, bentonite will be unsaturated and will effectively prevent the intrusion of water onto the canister surface for years according to the present knowledge. The potential to gasinduced expulsion depends also on the location of the defect in the canister (Chapter 9 in Pastina & Hellä, 2006). If a defect locates at the weld seam of a canister lid, i.e. at the top of a canister, the gas cannot expel large quantity of water from the canister. It is regarded to be much less likely to have a defect elsewhere than at the weld seam. The closer to the canister bottom the canister defect would locate, the larger the volume of displaced water could be. The formation of pathways in the buffer due to the expulsion of gas from the canister inside is not expected to displace a significant amount of porewater or dissolved radionuclides already present in the bentonite (Harrington and Horseman, 2003). Notwithstanding its likely limited significance, the impact of gas is taken into account in radionuclide transport calculations. It is explored in the additional scenario AD-III, where it is assumed that the build-up of gas pressure will exceed the confining pressure of the bentonite buffer and lead either to the expulsion of contaminated water (containing mainly the IRF of the spent fuel) from inside the canister through the buffer or to the expulsion of gas (mostly containing C-14) generated during corrosion of the iron insert and other metal parts.

75 5.3 Mathematical formulations 5.3.1 Near field transport In the following, a mathematical formulation for radionuclide transport in the near field is presented for the base case assuming that a canister has a hole that is small compared to the buffer thickness. The transfer from the canister interior through a hole in the defective canister wall into the innermost region in the buffer block around the top of the canister is modelled by means of an analytically derived transfer coefficient. The transfer coefficient takes into account the following two resistance effects : diffusion in the hole through the canister wall and diffusion into the bentonite from the mouth of the small hole. The analytical equation for the transfer coefficient is presented in Chapter 3 of Vieno & Nordman (2000). The steady-state mass flow rate by diffusion through the small hole can be calculated straightforwardly according to Fick's first law., (5-1) where A h is the area of the hole (m 2 ) D h is the effective diffusion coefficient in the material filling the hole (m 2 /s) x is the length of the hole (m) C i is the water-phase concentration of the tracer inside the canister (Bq/m 3 ) C 1 is the water-phase concentration at the outer end of the hole (Bq/m 3 ) Q h is the equivalent flow rate through the hole (m 3 /s). From the outer end of the hole diffusing species are assumed to spread out in spherical symmetry into bentonite. The mass flow rate between two concentric spherical surfaces at radii r 1 and much larger r 2 can be described according to the equation, (5-2) where r 1 is the radius of the hole in the canister C 1 is the water-phase concentration at the mouth of the hole (Bq/m 3 ) C 2 is the water-phase concentration at the distance r 2 in bentonite (Bq/m 3 ). where Q m is the equivalent flow rate (m 3 /s) from the mouth of the hole into bentonite.

76 The mass transport resistances are combined according to the principle of resistors in series to obtain the mass flow rate from the canister interior into the bentonite block:, (5-3) where C i is the water-phase concentration inside the canister (Bq/m 3 ) C 1B1 is the water-phase concentration in the first compartment of the bentonite block B1 (Bq/m 3 ) (this corresponds to C 2 in Equation 5-2). One should note that outside the canister, most of the diffusion resistance is close to the mouth of the hole where the diffusion area is very small. The mass flow resistance may, therefore, decrease significantly if the diffusivity in the bentonite close to the hole is increased, for example, due to effects of corrosion products. This would correspond to an increase of r 1 in Equation 5-2. In the large hole cases (Lh) the mass flow rate from the canister interior into the bentonite block is practically governed by Equation 5-1 alone. In such cases the mass flow into bentonite at the mouth of the hole is modelled to take place virtually without resistance. Thus Q m in Equation 5-2 is so high that the contribution of that part of resistance to the total transport resistance in the near field becomes negligible. Transport of radionuclides by diffusion within bentonite is modelled using several compartments. From the bentonite block near the defect nuclides may also diffuse into the buffer blocks above and below. Diffusion into larger volumes of the buffer is especially important for the transient behaviour of the radionuclide releases, since relatively short-lived sorbing nuclides are retained very effectively in the buffer. The release from the buffer into the bedrock fractures intersecting the deposition hole is described by applying the transfer coefficient Q F in the outermost compartment of the bentonite block opposite to the defect of the canister (see Figure 6-5). The mass transfer from the stagnant water in bentonite into the flowing groundwater in a fracture intersecting the deposition hole is limited by the boundary layer (film) resistance (Neretnieks, 1982; Hillebrand, 1985; Nilsson et al., 1991). The applied approximation disregards local heterogeneities (varying apertures and channelling) in groundwater flow in the fracture around the buffer. Moreover, the approximation does not account for the effects of rock spalling and other physical or chemical perturbations to the buffer-rock interface, which might affect the boundary layer resistance. SR 97 (Moreno & Gylling, 1998) and TILA-99 employed the same approximation for this resistance. According to the TILA-99 notations, the equivalent flow rate is:, (5-4) where D w u 2b V is the diffusion coefficient in water (m 2 /s) is the velocity of water in the fracture (m/s). is the aperture of the fracture (m)

77 r 2 is the radius of the deposition hole (m) The backfill in the upper part of the deposition hole and in the tunnel section above the deposition hole are also included in the near-field transport model. The transfer coefficients from these sections into the geosphere are evaluated from the FEFTRA results (Chapter 4.1). The REPCOM computer code is used to model the transport of radionuclides in the near field (Appendix 2). 5.3.2 Far field transport The treatment of the geosphere is largely similar to that in TILA-99 (Vieno & Nordman, 1999). The transport of dissolved radionuclides through the geosphere occurs along complex transport paths, which are represented in the geosphere model as single paths, characterised by integrated transport resistances (Poteri, 2007). The radionuclides are retarded from the rate of pure advective transport by matrix diffusion and, for many species, by sorption on matrix pore surfaces. Releases of radionuclides from a failed canister may feed multiple paths in the far field, but for radionuclide transport modelling purposes this variability is omitted 12. This simplification means that the paths are treated as identical ones, and only transport along a single, representative and nondispersive path is calculated. Numerical exercises show that longitudinal dispersion is of minor importance (TILA-99). In this work it is assumed that the rock matrix farther than ten centimetres from a fracture wall is inaccessible to diffusing radionuclides. The assumption of a limited matrix diffusion depth is, conservative, since greater thickness could potentially further retard and spread release pulses before their arrival in the biosphere. There is evidence from, e.g. logging of formation factors by electrical methods (Löfgren & Neretnieks 2003; Liu et al., 2006), that greater matrix depths may, in reality, be accessible. However, some investigation methods suggest smaller depths in the order of a few centimetres. The available extent of matrix diffusion based on numerous sources is discussed by Jakob (2004). For the case where a constant water-phase concentration C o of a stable species prevails at the inlet of the fracture beginning at t = 0, and omitting the advective delay, the water-phase concentration at distance L into the fracture is: (5-6) where u (a 1/2 ) is a lumped parameter describing the flow and matrix material transport properties of the migration route for the given species 12 The diversity of paths was regarded in the groundwater flow and solute transport analyses. See Chapter 4.

78 (5-7) where p is the porosity of rock matrix (-), D e is the effective diffusion coefficient from the fracture into rock matrix (m 2 /s), R p is the retardation factor of the species in rock matrix (-), W is the width of the flow channel (m), L is the transport distance (m), Q is the flow rate in the channel (m 3 /a), t is time (a). The effects of the variation of the parameter u are illustrated in Appendix 1. In a semiinfinite case described above the transport parameter values of various segments of a migration path can be integrated to represent one single effective value. With the numerical model FTRANS it also is possible to describe the heterogeneities perpendicularly to advection. A basic assumption in the FTRANS calculations was a 1 cm thick altered zone next to the fracture and a non-permeable boundary at a distance of 10 cm.

79 6 CALCULATION CASES Calculation cases are set for each assessment scenario (see Chapter 5.1) following a tree methodology that provides traceability of assumptions and data used in each calculation case (e.g. Figure 6-1). The tree methodology for the formulation of calculation cases is systematic and consistent, and supports the identification of realistic calculation cases, sensitivity cases, and what if cases, and even supplementary calculation cases within each scenario. A complete list of the analysed cases and their classification as realistic cases, sensitivity cases, and what if cases is given in Appendix 3. It must be noted that the results of the most realistic case 13 of the defective canister scenario (the base case of this analysis) serve as a basis for comparison with the results of the other cases. The sensitivity cases account for uncertainties in the knowledge on the state and behaviour of the system (or parts of it) in the long term, which is reflected in the variability of the data used in the analyses. The what if cases account for uncertainties in the timing and magnitude of the events that could affect the repository system during its evolution. The assumptions made in the what if cases can go beyond the ranges of events and processes regarded possible in the repository system. The supplementary calculation cases aim to evaluate the robustness of the system against unexpected events and processes. In Posiva's safety case the so called main scenario is based on the expected evolution of the repository system, as mentioned in the beginning of Chapter 5. Among other, in the main scenario it is assumed that all the canisters will be intact at the emplacement time and they will resist any plausible processes without leakages until a very distant future (Pastina & Hellä, 2006). With such an assumption, the current radionuclide transport analysis would be meaningless and, therefore, all the calculation cases here incorporate either immediate or early canister defects. 6.1 General assumptions In the calculation cases set for each assessment scenario it is assumed that the spent fuel will start to dissolve as soon as it comes into contact with groundwater for whatever the reason. Also common to all the calculation cases is the assumption of a defective canister either due to a failure in the encapsulation process or due to events and processes able to damage canisters and other barriers in the short or long term (Defective Canister Scenarios, DCS and Additional Scenarios, AD). The calculation cases due to the possible damage of a canister due to inadvertent human intrusion (HI) will be treated within the Biosphere Assessment. 13 This refers to the most realistic case within the analysed set of calculation cases. The terming "realistic" merely refers to less conservative cases than those in the other categories of cases. Note that calculation cases are not equivalent to scenarios. Instead, they are subsets of scenarios.

80 The calculation cases defined for the safety analysis in this report take into account the size of the penetrating defect (or hole) in the case of DCS (Figure 6-1 and Figure 6-2) and the time at which water enters the canister (Figure 6-1 to Figure 6-4). At the time water enters the canister, the state of spent fuel and the barriers, together with the physico-chemical conditions in the near and far field, are defined by a few parameters (see Chapter 6.2: inventory data, IRF, K d, solubility, etc.) that may be varied 1) depending on the conditions considered and 2) in order to check the importance of a given barrier and/or parameter ( sensitivity analyses, what if cases). 6.1.1 Calculation cases in the defective canister scenario (DCS) In the defective canister scenario it is assumed that one canister has either an initial penetrating defect in the form of one of three alternative diameters (DCS-II, Figure 6-1 and Table 6-1) or that one canister fails in the long term as a consequence of corrosion (DCS-I, Figure 6-2 and Table 6-2), which may be enhanced by external events or internal processes. In this context, the defect time (t) indicates the time when water enters a canister through a defect and the dissolution of fuel matrix starts, which in DCS-II is t = 0 a for all the calculation cases (Figure 6-1). In the identifiers Sh1 stands for a small hole of 1 mm diameter, which is regarded as the maximum size of a defect that could remain undetected with non-destructive methods (Raiko, 2005). Sh1 is the most realistic case (the base case) to which the results of the other cases of RNT-2008 are principally compared. Although the input data to be used in the calculation of Sh1 is specified later on in Chapter 6.2, from Figure 6-1 it follows that a default value is used for groundwater flow, and that the selected solubility limits and distribution coefficients (K d ) for buffer and backfill corresponds to dilute/brackish groundwater. The default flow refers to the near-field flow rates and the far-field transport resistances that are selected as base values from the results of the EPM (Chapter 4.1) and DFN (Chapter 4.2) modelling.

81 Time of 1 mm defect t=0 Identifier in the report Calculation TABLE 6-1 Cases in DCS-II Sh1-EPR Groundwater Sh1-VVER Groundwater flow DCS-II.1 Sh1 Sh1Fd composition default Sh1Irf dilute/brackish Sh1Q DCS-II.2 high Size of defect in DCS-II 4 mm t=0 most likely defect likely defect unlikely t=0 defect 100 mm saline dilute/brackish saline dilute/brackish saline default high default high default high default high default high DCS-II.3 DCS-II.4 DCS-II.5 DCS-II.6 DCS-II.7 DCS-II.8 DCS-II.9 DCS-II.10 DCS-II.11 DCS-II.12 Sh1Sal Sh1QSal Sh4 Sh4Q Sh4QSal LhQ LhQSal Lh Q Irf Lh B Q Lh B Q Fd Figure 6-1. The tree structure of the calculation cases in the Defective Canister Scenario DCS-II, where an initial penetrating defect is assumed. Sh1 is the base case to which the results of other cases are compared. Sh4 and Lh (Figure 6-1) stand for a small hole of 4 mm and a large hole of 100 mm diameter, respectively. Defects of these sizes can be reliably detected by non-destructive methods and it is not considered possible that a canister with such an initial penetrating defect would pass the quality control. Thus, the calculation cases derived for Sh4 and Lh do not aim to be realistic but to test the significance of the defect size. The Sh4 cases can be categorised as sensitivity cases and the Lh cases as "what if" cases. The diameter of 100 mm of the large hole is large enough to make its transport resistance negligible even, if diffusion is taken into account as the only transport mechanism. Hence transport resistance would not decrease noticeably with further enlargement of the hole. EPR and VVER stand for the fuel of the Olkiluoto 3 reactor and of the Loviisa 1-2 reactors, respectively. Q indicates high flow rates in all release routes, while the absence of Q means default flow rates. Sal stands for saline groundwater, and the lack of Sal implies dilute/brackish water. Fd stands for elevated fuel degradation rate in the cases, where the rate differs from 10-7 a -1 (the chosen variant is 10-6 a -1 ; see Chapter 6). Irf denotes an assumption, that only the radionuclides in the instant release fraction

82 (IRF) of fuel are released. B after a hole-diameter mark means that the defect is bentonite filled, while the absence of B implies that the defect is filled with water 14. Table 6-1. Identifiers and summary descriptions of each selected calculation case in the Defective Canister Scenario DCS-II that assumes an initial penetrating defect in a canister (See Figure 6-1). The defect becomes effective at t = 0 a in all cases. Calculation case Time of defect (a) Diameter of canister defect (mm) Filling of defect (W=water B=bentonite) Groundwater flow rate Q Groundwater chemistry (D = dilute B = brackish Sal = Saline) Remarks Sh1 0 1 W default D/B base case Sh1-EPR 0 1 W default D/B Sh1-VVER 0 1 W default D/B Sh1 Fd 0 1 W default D/B fuel degradation 10-6 /a Sh1 Irf 0 1 W default D/B IRF only Sh1 Q 0 1 W high D/B Sh1 Sal 0 1 W default Sal Sh1Q Sal 0 1 W high Sal Sh4 0 4 W default D/B Sh4 Q 0 4 W high D/B Sh4 Q Sal 0 4 W high Sal Large hole cases ("what if" cases) Lh Q 0 100 W high D/B Lh Q Irf 0 100 W high D/B IRF only Lh Q Sal 0 100 W high Sal Large hole cases with a hole filled with bentonite ("what if" cases) LhB Q 0 100 B high D/B LhB Q Fd 0 100 B high D/B Fuel degradation 10-6 /a In the defective canister scenario DCS-I, it is assumed that a canister becomes defective in the long term as a consequence of corrosion processes (Figure 6-2 and Table 6-2). The considered hole diameter of 100 mm represents a very unfavourable case, where the effect of sulphide-rich groundwater is focused on a small surface area of the disposal canister and where further growth of the hole would practically not decrease its transport resistance. The beginning of the transport of radionuclides through the defect is considered at two alternative times. At the end of the identifiers, t4 and t5 indicate that the defect becomes effective at 10 000 years and at 100 000 years, respectively, after the repository closure. The cases LhB Q in DCS-II and LhB Q t4 and LhB Q t5 14 Later in the report, B is also used in the beginning of an identifier for such cases in the Assessment Scenario AD-II, where bentonite in the buffer domain has weakened properties.

83 differ only in the time at which the defect becomes effective. This allows the studying of the effect of release time. The other systematically generated cases within this scenario were not analysed, since they would lead to illogical combinations. For example, a combination of enhanced corrosion rate and a large defect can only be developed in relevant time frame in case of high flow rate that is not anticipated together with the appearance of saline water. Identifier Calculation in this report Cases in DCS-I TABLE 6-2 Groundwater composition dilute/brackish Time of defect t=10 000 a Groundwater flow default high DCS-I.1 DCS-I.2 Lh B Q t4 Size of defect at time t: 100 mm DCS-I t=100 000 a saline dilute/brackish default high default high DCS-I.3 DCS-I.4 DCS-I.5 DCS-I.6 Lh B Q t5 saline default high DCS-I.7 DCS-I.8 Figure 6-2. Calculation cases in the Defective Canister Scenario DCS-I, where a large penetrating defect develops after thousands of years. Table 6-2. Identifiers and features of the analysed calculation cases in the Defective Canister Scenario DCS-I, where it is assumed that a penetrating defect (a large hole of 100 mm) develops thousands of years after canister emplacement (see Figure 6-2). Calculation case Time of defect (a) Canister defect diameter Defect filling (B = bentonite) Flow (default or high) Groundwater chemistry (saline, dilute/brackish or glacial) LhB Q t4 10 4 100 mm B high dilute/brackish LhB Q t5 10 5 100 mm B high dilute/brackish 6.1.2 Calculation cases in the rock shear/earthquake scenario (AD-I) The deposition holes will be located avoiding major fracture and deformation zones, which could form hydraulic connections up to the ground surface or to the sea and which could be reactivated as a consequence of adjustments of in situ rock stresses or of earthquakes, whether postglacial or not. Although in the Olkiluoto area no major

84 seismic events have been registered in historical times, the possibility of future strong 15 earthquakes, most likely in the period following a future glaciation, cannot be excluded and it is a requirement of the regulator to take into account the possibility of damage of one or several canisters due to rock shear, irrespective of its possible cause (STUK, 2001). Beside this requirement, it may be possible that a non-hydraulic conductive feature remains undetected during the construction and emplacement of the copper canisters in the repository. Finland in general, and Olkiluoto in particular has good tectonic stability, which is reflected by limited historical seismicity: Earthquake magnitudes in Finland have never reached 5 on the Richter scale (e.g. Mäntyniemi & Ahjos, 1990, Ahjos & Uski,1992); the latest earthquake of magnitude 4.9 dates from the 1880s (Mäntyniemi, 2005) and there are no evidences of post-glacial faults at the site (Lindberg, 2006). Moreover, La Pointe and Hermansson (2002) studied fracture displacements with respect to potential future seismicity for Olkiluoto, considering earthquakes of magnitudes (M L ) of 5.5 through 7...8, and estimated that the probability for a canister failure due to a rock displacement exceeding 10 cm could be of the order of 0.0041 % 16. Thus, the expectation value of the number of failed canisters would be around 0.12 (out of 3 000) over a 100 000 year period. In this scenario, the canister is assumed to fail as a consequence of the sudden displacement of a fracture intersecting the deposition hole. Four calculation cases were studied, according to the time when a rock shear or an earthquake could occur causing canister failures: AD-I.2 (RS1 in Figure 6-3 and Table 6-3) at 1 000 years, AD-I.6 (RS2) at 10 000 years, and AD-I.10 (R3) and AD-I.12 (RS3g) at 70 000 years after the closure of the repository. The last two cases relate to the potential occurrence of a postglacial earthquake as it would take place 70 000 years AP. According to the Weichselian-R climatic scenario defined in the context of the repository evolution (Pastina & Hellä, 2006) the next ice-sheet to cover the site will melt at that time. It is worth noting that this time would be much later (circa 176 ka AP) in the Emissions-M Scenario. The melting of the ice-sheet results in the stress release of the bedrock and this is why the timing of the melting is important. All these cases address the uncertainty in the timing of the event. As the above discussion suggests, it is reasonable to suppose that the magnitude of this class of events depends on the time of occurrence. Extensive block movements are less potential before the next glaciation than during the period starting with the recovery of the earth crust from the load of ice. As shown in Figure 6-3, all the cases to be calculated imply high flow of dilute/brackish water. The cases involving saline water are not calculated. The reason for the selection of the cases with dilute/brackish water is the assumption that a sudden displacement and opening of a fracture would allow fast downward intrusion of near-surface waters. In the case with a post-glacial earthquake at 70 000 years, the intrusion of glacial meltwater is also considered (RS3g). In this case the selected solubilities correspond to glacial meltwater composition (see Table 6-12 in Chapter 6.2), and the fuel degradation 15 "Strong" is used here as an everyday language word, not as an earthquake magnitude term. 16 The estimate is based on the simulations resulting in the canister failure probability being 0.0021 (i.e. 6 canisters out of 3 000) and the probability of such a canister-damaging earthquake being 0.02.

85 rate is also assumed to be higher than in other cases, reflecting the higher potential to oxidizing conditions in glacial waters. In the identifiers for the assessed cases, RS refers to the assumed direct consequences of rock shear/earthquake, and g for oxic glacial water that will flow down into the repository depth. Calculation Cases in AD-I Identifier RNT-2008 Groundwater Groundwater flow composition default Event time dilute/brackish t=1 000 a high* saline default high* AD-I.1 AD-I.2 AD-I.3 AD-I.4 RS1 Rock shear/ Earthquake in AD-I default dilute/brackish t=10 000 a high* saline default high* AD-I.5 AD-I.6 AD-I.7 AD-I.8 RS2 t=70 000 a glacial dilute/brackish saline default high* default high* default high* AD-I.9 AD-I.10 AD-I.11 AD-I.12 AD-I.13 AD-I.14 RS3 RS3g * The high-flow assumption differs from the other high-flow cases in that in the RS cases the high flow rate is applied only at the fracture intersecting the deposition hole (Q F). Figure 6-3. Calculation cases in the rock shear/earthquake scenario AD-I. The probability and magnitude of event vary (low in RS1 and RS2; higher in RS3), as well as the time of occurrence. RS3g is analysed in order to study the effect of intrusion of oxic glacial meltwater into the disposal depth.

86 Table 6-3. Identifiers and features of the analysed calculation cases in the Additional Scenario AD-I, where it is assumed that a canister in the deposition hole will fail due to a less significant rock shear (RS) or a more significant block movement at a postglacial earthquake (EQ). Calculation case Time of rupture (a) Event Groundwater chemistry (saline, dilute/brackish or glacial) Fuel degradation rate (1/a) Flow (default or high) RS1 10 3 RS dilute/brackish 10-7 high* RS2 10 4 RS dilute/brackish 10-7 high* RS3 7 10 4 EQ dilute/brackish 10-7 high* RS3g 7 10 4 EQ glacial 10-6 high* * See the note below Figure 6-3. 6.1.3 Calculation cases in the disrupted buffer scenario (AD-II) In this scenario it is assumed either that bentonite is poorly emplaced around an initially defective canister or that bentonite becomes ineffective with time e.g. due to the penetration of glacial melt water giving rise to chemical erosion (see process 5.2.3 in Miller & Marcos, 2007). Consequently, bentonite will not perform as expected and corrodants (e.g. sulphide) will reach the canister easier. The case is modelled assuming that sulphide-rich solutions will come into contact with the canister surface causing an initially small defect (1 mm) to enlarge by corrosion up to 400 mm in one step at 100 000 years AP. Thus the canister would lose its containment capacity almost entirely. It can be shown that further enlargement of a hole from about 400 mm would not anymore increase the activity release rates from a canister. Therefore, it is unnecessary to consider even larger openings. In this scenario, four calculation cases are derived and two calculated (Figure 6-4 and Table 6-4: Note that the D e value in this table is applied after 100 000 years). As a consequence of the buffer losing its expected barrier function, very high flow rates are expected and this is why AD-II.1 and AD-II.3 are not calculated. In the identifiers for the calculated cases, B stands for bentonite as its barrier function fails, Sh-Lh stands for a small hole enlarging in time, Q stands for high flow, and g for glacial water composition that will reach the repository depth.

87 Disruptions in the buffer AD-II Defect size at t=0... 1 mm... t=10 5 a 400 mm Groundwater composition dilute/brackish glacial Identifier Calculation in this report Cases in AD-II TABLE 6-4 Groundwater flow AD-II.1 default high default high AD-II.2 AD-II.3 AD-II.4 BSh-LhQ BSh-LhQg Figure 6-4. Calculation cases in the Additional Scenario AD-II (disruptions in the buffer due to deviations in initial conditions or later events/processes). The hole is assumed to be enlarged from 1 mm to 400 mm at 100 000 years AP. 6.1.4 Calculation cases in the gas scenario (AD-III) The scenario AD-III takes into account the formation of gas in a canister with an initial defect and the fate of the radionuclides that could be expelled from the canister as a consequence of the gas-induced displacement of contaminated water or when gaseous radionuclides are released in the gas phase. The following two calculation cases are considered (Table 6-4): GASexW, where gas expels radionuclides in the water phase, and GASexG, where radionuclides (mostly C-14) are expelled in gaseous form. As stated in the process report (Miller & Marcos, 2007), the major uncertainty in the consideration of water-gas systems in a canister is the location and size of the defect. If a perforation was located at the top of a canister, only small gas volume could accumulate in the canister headspace and thus cause only minor expulsion of water out of the canister. There are significant conceptual and numerical uncertainties related to the detailed description and modelling of the behaviour of the system including gas transport through the bentonite buffer. In GASexW the conditions under which gas-induced displacement of radionuclide contaminated water from the canister interior through the defect are highly hypothetical. It would require both of the following conditions to make GASexW a safety-relevant case: A penetrating defect well below the canister lid: the canister pre-fabrication tests give no potential for such defects elsewhere than at the weld (see Raiko, 2005; Koivula, 2005). A penetrating defect at the bottom is regarded possible practically only, if the canister is emplaced upside down in the deposition hole. The possibility of emplacing the canister in the deposition hole upside down is minimized during the transfer of the canister from the encapsulation plan to the repository through a shaft and emplacement in the deposition holes (see Tanskanen & Palmu, 2004);

88 Water inflow into the canister: the water inflow would be strongly reduced due to the sealing of the hole by bentonite and corrosion products (see Chapter 9 in Pastina & Hellä, 2006). Nonetheless GASexW is included in the analysed cases (Table 6-4). It is assumed that a gas-driven water pulse, beginning at 2800 years after the emplacement and lasting for a further 1300 years, propels water from the canister interior through the buffer to a fracture in the bedrock. For comparison, the timing of the event is the same as in the case PD-EXPELL in the radionuclide transport report of KBS-3H (Smith et al., 2007). It is well-reasoned to assume in the analysis that gas expels only the IRF (instant release fraction) from the canister. Only radionuclides that can partition into the gas in the canister headspace may be transported further as gases (e.g. C-14). However, unless the gas pressure exceeds the confining pressure of bentonite buffer, gas release from the canister will still be diffusion limited. A number of gas migration experiments have been performed in compacted bentonite over the last 20 years. Gas flow through saturated bentonite is discussed e.g. in Harrington & Horseman (2003), SKB (2006, 2006a), Rodwell (2005) and Pastina & Hellä (2006). There is strong evidence that gas can flow in bentonite through a network of pressureinduced pathways. When gas generation ceases, or if the gas generation rates are low enough, the transport pathways in bentonite have been found to close. After their closure, gas migrates solely by diffusion. In GASexG, where the expulsion of C-14 as gas is assumed, pathways for advective gas flow are assumed to form and to allow at least half of the C-14 originated in the IRF to be released instantaneously from the near field to the far field and further to the biosphere. It is assumed that the gas pressure inside the canister reaches the gas breakthrough pressure of bentonite at 900 years after repository closure. The chosen timing is based on the same reasoning as in the case PD- VOL-1, calculated within the radionuclide transport analysis of KBS-3H (Smith et al., 2007).

89 Table 6-4. Identifiers and features of the analysed calculation cases in the Additional Scenario AD-II, where it is assumed that the bentonite buffer will become disrupted, and in the Additional Scenario AD- III that considers gas-induced migration of radionuclides. Calculation case* Time span for defect enlargement (a) Flow (default or high) Groundwater chemistry (saline, dilute/brackish or glacial) Fuel degradation rate or release mode (1/a) Remarks Additional Scenario AD-II B Sh Lh Q 0 10 5 10 6 high dilute/brackish 10-7 Bentonite: D e 8.6 10-10 m 2 /s and Backfill: D e 4 10-10 m 2 /s Both values from 10 5 a on B Sh Lh Q g 0 10 5 10 6 high glacial 10-7 Same as above Additional Scenario AD-III GASexW 2800 4100 default dilute/brackish only IRF GASexG 900 default dilute/brackish only 50 % of C-14 in IRF * B in the beginning of a case identifier indicates disrupted bentonite buffer. GASex is a general identifier for the calculation cases with gas-induced transport of radionuclides. W at the end of identifier means that contaminated water is expelled from the canister, and similarly G means expel of gaseous radionuclides. 6.1.5 Supplementary cases to appraise the robustness/behaviour of the system The purpose of the supplementary cases is to appraise the robustness and behaviour of the system. Most of them would belong to the defective canister scenario (DCS), but the reasoning in the selection of data is beyond any scientifically and technically realistic considerations as to the processes involved in the evolution of the system and the current understanding of its behaviour. Thus, these belong to so called what if cases. Varying systematically the equivalent flow rates for the near-field routes depicted in Figure 6-5, several calculation cases were defined to evaluate, how different flow patterns influence the release rates of different kinds of radionuclides (see Chapter 7.4). The safety function of the far field was studied in a special set of calculations. The barrier effect of the geosphere as a function of a lumped parameter, u, is illustrated in Appendix 1. The far-field transport is determined there with the following six parameters: p, D e, R p, W, L, and Q. Any combination of these parameters resulting in the same value of u leads to the same transport behaviour. In this sense geosphere transport depends on the single parameter u.

90 6.2 Data 6.2.1 Near-field flow and geometrical model The near-field modelling geometry for all the calculation cases of the DCS scenarios, where a penetrating defect is assumed to occur or develop at the top of a canister, is described in Figure 6-5. The concept includes three near-field release routes featuring the route-specific equivalent flow rates Q F, Q DZ, and Q TDZ, also depicted in the figure. Figure 6-5. The conceptual near-field transport model. Diffusion through bentonite blocks B1 B3 and tunnel backfill is included in the model. The concept includes the following equivalent flow rates for three release routes to the geosphere: Q F from bentonite to a fracture in the bedrock, Q DZ from the backfill in the deposition hole to the bedrock, and Q TDZ from the tunnel to the bedrock. The conceptual model is the same as presented in Nordman & Vieno (2003). The diameter of the deposition hole is assumed to be 1.76 metres. That is, the thickness of compacted bentonite between the canister and the rock is 35 cm. The 0.5 metres of bentonite at the bottom of the deposition hole (Saanio et al., 2006) has conservatively been ignored in the model. The total amount of buffer is thus 11.0 m 3, representing 90 % of the buffer in the actual KBS-3V design. The transport of radionuclides is assumed to take place by diffusion within the bentonite blocks B1, B2 and B3. It is modelled using several compartments. The release from the buffer into rock fractures intersecting the deposition hole/drift is concentrated in block B1, where the transfer coefficient Q F is applied in the outermost compartment. From block B1, radionuclides may also diffuse into buffer blocks B2 and B3, above and below of block B1, respectively.

91 The backfill in the upper part of the deposition hole and in the tunnel section above the deposition hole are included in the near-field transport model. The transfer coefficients from these sections into the geosphere are evaluated and selected from the FEFTRA results (Chapter 4.1). The transfer coefficients from the near field to the geosphere (Q F, Q DZ and Q TDZ ) and the WL/Q values are presented in Table 6-5. The geometrical data for the near field are presented in Table 6-6. For the cases in the rock shear/earthquake scenario (AD-I: calculation cases RS1, RS2, RS3 and RS3g) the conceptual model is almost similar to that in the penetrating defect scenarios (DCS). It differs in the thickness of bentonite that is assumed to be reduced to 10 cm because of the remold of the buffer caused by a rock shear displacement. Thus the outer radius of bentonite is set to 0.63 m instead of 0.88 m (Table 6-6). The canister is assumed to be severely damaged over a height of 35 cm from the top that is in contact with block B1. In the cases RS3 and RS3g the water in the route of Q F is assumed to get fully mixed with the entire water volume of the canister. The disrupted buffer scenario (AD-II) is analysed with the cases BSh LhQ and BSh LhQg. Their conceptual model is as in the cases for the DCS, except that the defect size is assumed to enlarge from 1 mm to 400 mm at 100 000 years AP (see Section 6.1.3). Also the diffusion coefficients in the buffer and tunnel backfill are increased to reflect their weakened performance. Table 6-5. Default and high equivalent flow rates in all the cases Sh, Lh (DCS-II and DCS-I; see the calculation cases in Figure 6-1, Figure 6-2, Table 6-1 and Table 6-2), and BSh LhQ (AD-II; see the calculation cases in Figure 6-4 and Table 6-4) and RS (AD-I, see calculation cases in Figure 6-3 and Table 6-3). Identifier Q F (L/a) Q DZ (L/a) Q TDZ (L/a) Transport resistance WL/Q (a/m) default 0.2 2 10 50 000 high 2 20 100 5 000 RS 2* 2 10 5 000 * Note that in this type of cases high flow applies only at the route from bentonite to a fracture (Q F), and that for RS3 and RS3g the flow gets mixed with the canister volume.

92 Table 6-6. Geometrical data for modelling of cases. Canister interior water volume: BWR 0.7 m 3, EPR 0.5 m 3, Loviisa VVER 0.45 m 3 amount of fuel: BWR 2.14 tu, EPR 2.13 tu, Loviisa VVER 1.44 tu transfer into the innermost compartment of buffer block B1 Penetrating defect diameter: 1 mm in Sh1 cases, 4 mm in Sh4 cases, and 100 mm in Lh cases. In B Sh Lh Q cases the defect increases from 1 mm diameter to 400 mm at 100 000 years. RS1 RS3 cases: diffusion contact number of compartments: 1 in the radial and axial directions Buffer block B1 height: 0.35 m inner radius: 0.53 m outer radius: 0.88 m (0.63 m in the RS cases) transfer into the geosphere from the outermost compartment: Q F L/a number of compartments in the radial direction: 11 number of compartments in the axial direction: 1 Buffer block B2 height: 1.5 m radius: 0.88 m (0.63 m in the RS cases) number of compartments in the radial direction: 12 number of compartments in the axial direction: 4 Buffer block B3 height: BWR 4.45 m, EPR 4.90 m, Loviisa VVER 3.25 m inner radius: 0.53 m outer radius: 0.88 m (0.63 m in RS cases) number of compartments in the radial direction: 11 number of compartments in the axial direction: 4 Backfill in the upper part of the KBS-3V deposition hole height: 1.0 m radius 0.88 m transfer into the geosphere : Q DZ L/a number of compartments in the radial direction: 1 number of compartments in the axial direction: 2 Backfill in the tunnel section above the KBS-3V deposition hole volume 100 m 3 transfer into the geosphere: Q TDZ L/a number of compartments: 1 in the radial and axial directions no diffusion resistance in the tunnel 6.2.2 Radionuclide inventory The radionuclide inventories vary between fuel types, as described in Anttila (2005). In addition to the OL1 and OL2 fuel types, other fuels were considered assuming in extra calculation cases that an initial penetrating defect is in a canister of either VVER-440 fuel (case Sh1-VVER in Figure 6-1) or EPR fuel (case Sh1-EPR). The VVER-440 (PWR) fuel in a canister in case Sh1-VVER contains 1.44 tonnes of U metal, and the corresponding quantity for EPR fuel in case Sh1-EPR is 2.13 tonnes. In all other cases, the defect was assumed to affect a single canister containing BWR type fuel with 2.14 tu.

93 In RNT-2008 the applied instant release fractions of the relevant elements in fuel matrix are equal to the conservative values presented in the SR-Can data report (SKB, 2006b) (Table 6-7). These values are used for the partitioning between the IRF and fuel matrix, the result of which for the different fuel types is included in Table 6-8 through Table 6-10. Table 6-7. Instant release fractions of fuel matrix in all cases. Element C Cl Se Sr Tc Pd Sn I Cs Conservative instant release fraction (%) 10 10 0.1 1 1 1 0.01 5 5

94 Table 6-8. The radionuclide inventory of a single canister containing BWR fuel from the OL1 and OL2 reactors after 30 years of cooling, with the half-lives and partitioning between fuel matrix, instant release fraction (IRF), Zircaloy and other metallic parts. The assumed burnup is 40 MWd/kgU and enrichment 4.2 %. (Anttila, 2005) Radionuclide Half-life (a) Activity inventory after 30 years of cooling (GBq/tU) Fuel matrix without IRF Partitioning (%) IRF Zircaloy Other metal parts Trace elements C-14 5.7 10 3 2.78 10 1 30 3 33 33 Cl-36 3.0 10 5 1.04 10 0 45 5 50 0 Mo-93 4.0 10 3 1.13 10 1 0 0 0 100 Actinides and fission products Se-79 2.95 10 5 3.30 10 0 99.9 0.1 0 0 Sr-90 2.9 10 1 1.70 10 6 99 1 0 0 Zr-93* 1.5 10 6 8.26 10 1 100 0 0 0 Tc-99 2.1 10 5 6.12 10 2 99 1 0 0 Pd-107 6.5 10 6 4.29 10 0 99 1 0 0 Sn-126 1.0 10 5 2.18 10 1 99.99 0.01 0 0 I-129 1.6 10 7 1.14 10 0 95 5 0 0 Cs-135 2.3 10 6 2.15 10 1 95 5 0 0 Cs-137 3.0 10 1 2.36 10 6 95 5 0 0 Sm-151 9.0 10 1 1.10 10 4 100 0 0 0 Ra-226 1.6 10 3 Th-229 7.3 10 3 Th-230 7.7 10 4 Pa-231 3.2 10 4 U-233 1.6 10 5 2.41 10-3 100 0 0 0 U-234 2.4 10 5 5.59 10 1 100 0 0 0 U-235 7.0 10 8 7.45 10-1 100 0 0 0 U-236 2.3 10 7 1.30 10 1 100 0 0 0 U-238 4.5 10 9 1.16 10 1 100 0 0 0 Np-237 2.1 10 6 1.30 10 1 100 0 0 0 Pu-238 8.8 10 1 8.71 10 4 100 0 0 0 Pu-239 2.4 10 4 1.05 10 4 100 0 0 0 Pu-240 6.5 10 3 1.98 10 4 100 0 0 0 Pu-241 1.4 10 1 9.84 10 5 100 0 0 0 Pu-242 3.8 10 5 7.61 10 1 100 0 0 0 Am-241 4.3 10 2 1.08 10 5 100 0 0 0 Am-243 7.4 10 3 7.62 10 2 100 0 0 0 Cm-245 8.5 10 3 6.16 10 0 100 0 0 0 Cm-246 4.7 10 3 1.19 10 0 100 0 0 0 Zircaloy and other metal parts Ni-59 8.0 10 4 1.32 10 2 0 0 0 100 Ni-63 9.6 10 1 1.41 10 4 0 0 0 100 Zr-93 1.5 10 6 8.16 10 0 0 0 100 0 Nb-94 2.0 10 4 3.00 10 1 0 0 0 100 * Originating in fuel. Zr-93 in Zircaloy is listed separately under Zircaloy and other metal parts.

95 Table 6-9. The radionuclide inventory of a single canister containing VVER-440 (PWR) fuel (case Sh1-VVER), with the half-lives and partitioning between fuel matrix, instant release fraction (IRF), Zircaloy and other metal parts. (Anttila, 2005) Radionuclide Half-life (a) Activity inventory (GBq/tU) Fuel matrix without IRF Partitioning (%) IRF Zircaloy Other metal parts Trace elements C-14 5.7 10 3 3.83 10 1 45 5 0 50 Cl-36 3.0 10 5 3.88 10-1 72 8 20 0 Mo-93 4.0 10 3 1.80 10 0 0 0 0 100 Actinides and fission products Se-79 2.95 10 5 3.24 10 1 99.9 0.1 0 0 Sr-90 2.9 10 1 1.62 10 6 99 1 0 0 Zr-93* 1.5 10 6 7.93 10 1 100 0 0 0 Tc-99 2.1 10 5 5.96 10 2 99 1 0 0 Pd-107 6.5 10 6 4.92 10 0 99 1 0 0 Sn-126 1.0 10 5 2.32 10 1 99.99 0.01 0 0 I-129 1.6 10 7 1.19 10 0 95 5 0 0 Cs-135 2.3 10 6 2.06 10 1 95 5 0 0 Cs-137 3.0 10 1 2.37 10 6 95 5 0 0 Sm-151 9.0 10 1 1.53 10 4 100 0 0 0 Ra-226 1.6 10 3 Th-229 7.3 10 3 Th-230 7.7 10 4 Pa-231 3.2 10 4 U-233 1.6 10 5 2.88 10-3 100 0 0 0 U-234 2.4 10 5 5.32 10 1 100 0 0 0 U-235 7.0 10 8 8.36 10-1 100 0 0 0 U-236 2.3 10 7 1.21 10 1 100 0 0 0 U-238 4.5 10 9 1.16 10 1 100 0 0 0 Np-237 2.1 10 6 1.59 10 1 100 0 0 0 Pu-238 8.8 10 1 1.09 10 5 100 0 0 0 Pu-239 2.4 10 4 1.51 10 4 100 0 0 0 Pu-240 6.5 10 3 2.13 10 4 100 0 0 0 Pu-241 1.4 10 1 1.44 10 6 100 0 0 0 Pu-242 3.8 10 5 8.71 10 1 100 0 0 0 Am-241 4.3 10 2 1.57 10 5 100 0 0 0 Am-243 7.4 10 3 1.03 10 3 100 0 0 0 Cm-245 8.5 10 3 1.32 10 1 100 0 0 0 Cm-246 4.7 10 3 2.03 10 0 100 0 0 0 Zircaloy and other metal parts Ni-59 8.0 10 4 3.28 10 2 0 0 0 100 Ni-63 9.6 10 1 3.48 10 4 0 0 0 100 Zr-93 1.5 10 6 8.78 10 0 0 0 100 0 Nb-94 2.0 10 4 5.03 10 2 0 0 100 0 * Originating in fuel. Zr-93 in Zircaloy is listed separately under Zircaloy and other metal parts.

96 Table 6-10. The radionuclide inventory of a single canister containing EPR fuel (case Sh1-EPR), with the half-lives and partitioning between fuel matrix, instant release fraction (IRF), Zircaloy and other metal parts. (Anttila, 2005) Radionuclide Half-life (a) Activity inventory (GBq/tU) Fuel matrix without IRF Partitioning (%) IRF Zircaloy Other metal parts Trace elements C-14 5.7 10 3 1.90 10 1 45 5 0 50 Cl-36 3.0 10 5 7.49 10-1 72 8 20 0 Mo-93 4.0 10 3 1.34 10 1 0 0 0 100 Actinides and fission products Se-79 2.95 10 5 3.25 10 0 99.9 0.1 0 0 Sr-90 2.9 10 1 1.62 10 6 99 1 0 0 Zr-93* 1.5 10 6 7.98 10 1 100 0 0 0 Tc-99 2.1 10 5 5.98 10 2 99 1 0 0 Pd-107 6.5 10 6 4.89 10 0 99 1 0 0 Sn-126 1.0 10 5 2.31 10 1 99.99 0.01 0 0 I-129 1.6 10 7 1.19 10 0 95 5 0 0 Cs-135 2.3 10 6 2.22 10 1 95 5 0 0 Cs-137 3.0 10 1 2.37 10 6 95 5 0 0 Sm-151 9.0 10 1 1.50 10 4 100 0 0 0 Ra-226 1.6 10 3 Th-229 7.3 10 3 Th-230 7.7 10 4 Pa-231 3.2 10 4 U-233 1.6 10 5 2.90 10-3 100 0 0 0 U-234 2.4 10 5 5.36 10 1 100 0 0 0 U-235 7.0 10 8 8.15 10-1 100 0 0 0 U-236 2.3 10 7 1.22 10 1 100 0 0 0 U-238 4.5 10 9 1.16 10 1 100 0 0 0 Np-237 2.1 10 6 1.58 10 1 100 0 0 0 Pu-238 8.8 10 1 1.13 10 5 100 0 0 0 Pu-239 2.4 10 4 1.41 10 4 100 0 0 0 Pu-240 6.5 10 3 2.13 10 4 100 0 0 0 Pu-241 1.4 10 1 1.38 10 6 100 0 0 0 Pu-242 3.8 10 5 8.78 10 1 100 0 0 0 Am-241 4.3 10 2 1.51 10 5 100 0 0 0 Am-243 7.4 10 3 1.03 10 3 100 0 0 0 Cm-245 8.5 10 3 1.30 10 1 100 0 0 0 Cm-246 4.7 10 3 2.08 10 0 100 0 0 0 Zircaloy and other metal parts Ni-59 8.0 10 4 1.27 10 2 0 0 0 100 Ni-63 9.6 10 1 1.36 10 4 0 0 0 100 Zr-93 1.5 10 6 5.45 10 0 0 0 100 0 Nb-94 2.0 10 4 2.64 10 2 0 0 0 100 * Originating in fuel. Zr-93 in Zircaloy is listed separately under Zircaloy and other metal parts.

97 6.2.3 Release rate from fuel element In addition to the instant release fractions, the applied other activity release rates from the different parts of the fuel element are shown in Table 6-11. For fuel matrix degradation, the rate of 10-7 a -1 has been adopted (see Chapter 5), which means constant release rate over ten million years. For Zircaloy and other metal parts, the data of TILA- 99 have been used, except in the cases of the gas-related scenario AD-III (see below). Table 6-11. Degradation rates from the different parts of a fuel element. Parameter Unit Value Source / comments Fuel matrix fractional degradation rate Zircaloy fractional corrosion rate Fractional corrosion rate for other metal parts a -1 10-7 variant 10-6 a -1, see cases Sh1 Fd, LhB Q Fd, RS3g a -1 10-4 TILA-99 a -1 10-3 TILA-99 It is assumed that the inventory of activation products in Zircaloy and other metal parts is released congruently with the corrosion of the metal. A more pessimistic approach was taken in SR-Can, where no credit was taken for the delay due to the limited rate of metal corrosion. The assumed fractional corrosion rate of Zircaloy of 10-4 per year, taken from TILA-99 (p. 101 of Vieno & Nordman, 1999), is somewhat higher than the expected rate of corrosion in (SKB2006a), where the lower limit for Zircaloy lifetime is set to 100 000 years. About the corrosion of metal parts The other metal parts in a BWR fuel element are mostly stainless steel (53 kg/tu) and Inconel (2.9 kg/tu), which contains most of the Nb-94 and C-14 inventory. In TILA-99 it was assumed that the fractional corrosion rate of these metal parts is 10-3 per year, which has also been used in this work for most of the calculation cases. The corrosion of stainless steel can start only after water will first have entered the canister through a penetrating hole and, secondly, stainless steel will not anymore be electrochemically protected by cast iron of the canister insert. Initially, during the transient phase with oxygen possibly present, the corrosion rate of the cast iron insert is assumed to be fast (up to 30 m per year), but it will decrease rapidly as a surface-protective film of magnetite develops. An experimental corrosion study performed in compacted bentonite (Smart et al., 2004) for a period of about one year gave corrosion rates of 1 2 m per year. The cast iron insert has a thickness of about 60 mm against the copper inner surface. This would mean that the corrosion of the entire cast iron insert ( 13 tonnes) would take about 30 000 to 60 000 years. Iron is lower in the galvanic series than stainless steel, and thus the stainless steel parts would be protected against corrosion. In addition, the stainless steel parts are typically at least 1 cm thick in the upper part of the fuel element. A rough estimate is that in similar

98 conditions the corrosion rate of stainless steel is 10 % of the corrosion rate of cast iron, and thus the corrosion of 10 mm of stainless steel would take at least 50 000 years. Taking these corrosion rates into account it can be concluded that iron corrodes electrochemically preferentially to stainless steel, which will accordingly start to corrode after at least 30 000 years, and the corrosion rate of stainless steel will be much lower than 10-3 a -1, which is the conservative value used in this work for most of the calculation cases. Consequently, at the period of time when gas would be able to effectively expel either contaminated water or gaseous C-14 out of a canister, the metal parts would be only slightly corroded due to the slow corrosion rates. Moreover, the radionuclide releases from spent fuel matrix, Zircaloy, and other metal parts would be insignificant at this time. For these reasons it is assumed for simplification in the Additional Scenario AD- III, which takes into consideration gas formation and expelling due to the corrosion of the metal parts of fuel assemblies and that only the IRF of the nuclide inventory is released to the geosphere (see GASexW and GASexG in Table 6-4). 6.2.4 Solubility limits Solubility limits for the near field have been estimated by Grivé et al. (2007) for a range of groundwater types (dilute/brackish, saline, glacial). The applied solubility limits are presented in Table 6-12. It must be noted that in evaluating the solubility limits the amount of stable nuclides for several elements are taken into account (Table 6-13).

99 Table 6-12. Solubility limits. Element Solubility (M) dilute/brackish water saline glacial melt water C High * High High Cl High High High Ni 4.3 10-3 7.4 10-4 4.1 10-7 Se 3.1 10-10 5.2 10-6 4.5 10-11 Sr 9.1 10-5 3.7 10-4 1.1 10-5 Mo 2.6 10-8 9.2 10-9 1.5 10-4 Zr 1.7 10-8 1.6 10-8 1.8 10-8 Nb 3.8 10-5 6.2 10-5 2.9 10-3 Tc 4.2 10-9 4.0 10-9 4.5 10-9 Pd 2.5 10-6 2.8 10-6 2.7 10-6 Sn 1.2 10-7 1.0 10-7 2.7 10-6 I High High High Cs High High High Sm 7.5 10-8 3.8 10-8 1.8 10-9 Ra 2.2 10-8 5.9 10-8 8.8 10-7 Th 6.3 10-9 6.8 10-10 8.1 10-10 Pa 3.0 10-7 2.8 10-7 3.2 10-7 U 9.5 10-10 6.2 10-10 2.3 10-9 Pu 1.1 10-6 2.9 10-8 1.3 10-10 Np 1.1 10-9 7.2 10-10 8.2 10-10 Am 4.0 10-7 9.2 10-7 5.6 10-8 Cm 4.0 10-7 9.2 10-7 5.6 10-8 * High indicates that no solubility limit is applied in radionuclide release and transport calculations.

100 Table 6-13. Inventory of stable nuclides taken into account in estimating whether solubility limits are exceeded. (Anttila, 2005). Element Amount (mol/tu) BWR VVER-440 EPR Ni 1.20 10 2 3.45 10 2 1.31 10 2 Se 7.70 10-1 7.50 10-1 7.56 10-1 Sr 5.00 4.68 4.72 Zr 4.93 10 3 5.37 10 3 3.11 10 3 Zr (fuel matrix only) 5.30 10 1 5.00 10 1 5.10 10 1 Nb 4.91 6.80 10 1 3.50 10 1 Pd 1.20 10 1 1.50 10 1 1.50 10 1 Sn 2.90 10-1 3.08 10-1 3.08 10-1 Mo 1.77 10 1 3.60 1.70 10 1 6.2.5 Diffusion and retardation data in the near field. The diffusion coefficients and porosities for buffer and backfill in Table 6-14 are the central values of SR-Can (SKB, 2006b). They are applied in RNT-2008, except in the cases where alternate figures are separately mentioned. The used effective diffusivity D e in bentonite is at its highest value for Cs (cation), the second highest for neutral species and the lowest for anions (Cl, Se, Mo and I). The effective porosity in bentonite is for anions only 40 % of the value applied for neutral species and Cs. The same ratios of D e and between neutral species, anions and Cs are used for tunnel backfill as for bentonite. The cases BSh LhQ and BSh LhQg (AD-II scenario) were constructed for studying situations, where bentonite does not perform as expected. The weakened performance is simulated by increasing the diffusion coefficients in both cases: in BSh LhQ at the time of disposal to address the misplacement of bentonite blocks, and in BSh LhQg at 10 5 years to address buffer erosion by glacial meltwater. In these cases, equal D e were applied for all species, both in bentonite and tunnel backfill. The chosen effective diffusivities approach the product of porosity and the self-diffusivity of water and thus represent the maximum possible diffusivities in these materials.

101 Table 6-14. Data for bentonite buffer and tunnel backfill. Speciation Neutral Anion* Cs Buffer (bentonite) D e (m 2 /s) Grain density (kg/m 3 ) Backfill (bentonite/ballast) D e (m 2 /s) Grain density (kg/m 3 ) BSh LhQ (data applied from t = 0 a), and BSh LhQg (data applied after t = 10-5 a) Buffer D e (m 2 /s) Backfill D e (m 2 /s) 0.43 1.2 10-10 2700 0.23 5 10-11 2700 0.17 1 10-11 2700 0.092 4.2 10-12 2700 0.43 3 10-10 2700 0.23 1.3 10-10 2700 8.6 10-10 8.6 10-10 8.6 10-10 4.0 10-10 4.0 10-10 4.0 10-10 * Cl, Se, Mo and I are assumed to behave as anions in all cases, and Sn, Sm, Am and Cm only in glacial melt water cases. The K d values, which are mostly selected conservatively from the lower limit values given in Table A-12 of the SR-Can Data Report (SKB, 2006b) for saline and non-saline groundwaters, are listed in Table 6-15. An exception to the use of SR-Can sorption data is for Mo. No data for Mo are given in SR-Can and, in Table 6-15, the K d value for Mo, which is assumed to be present predominantly in anionic form (see below), is conservatively set to zero. Simulated salinities in the tunnel backfill at different times are presented in Table 4-1 in page 34.

102 Table 6-15. Sorption coefficients (K d values) in various groundwater types for buffer and backfill. Element K d (m 3 /kg) in bentonite buffer K d (m 3 /kg) in backfill (bentonite plus ballast) Dilute/ brackish Saline Glacial Saline & brackish Glacial melt water C 0 s s 0 s Cl 0 s s 0 s Ni 0.03 0.01 s 0.0096 s Se 0 s s 0 s Sr 0.0009 0.0002 s 0.00028 s Mo 0 s s 0 s Zr 0.1 0.2 s 0.065 s Nb 0.2 s s 0.095 s Tc 2.3 1.4 s 0.73 s Pd 0.3 s s 0.09 s Sn 2.3 1.4 0 0.69 0 I 0 s s 0 s Cs 0.018 0.006 s 0.0061 s Sm 0.8 0.5 0 0.31 0 Ra 0.001 0.0002 s 0.007 s Th 6 4 s 1.9 s Pa 0.2 s s 0.095 s U 0.5 s s 0.15 s Pu 4 2 s 1.3 s Np 4 2 s 1.2 s Am 10 4 0 3.2 0 Cm 10 4 0 3.1 0 s = same as for brackish/dilute water 6.2.6 Far-field data In TILA-99, C, Cl, Se, Pd, Sn, and I were assumed to exist as anions in the geosphere in all cases. According to the speciation calculations by Grivé et al. (2007), however, Sn will predominantly take the form of neutral hydroxide complexes. Mo, which was not among the safety-relevant elements considered in TILA-99, is also assumed to exist in anionic form in the present base case, Sh1. Carbon is conservatively assumed to be predominantly in methanic form and Pd is assumed to be dominated by neutral complexes (Grivé et al., 2007). Thus, in the present study only Cl, Se, Mo and I are assumed to be present as anions in the geosphere in all the groundwater types. In glacial meltwater also Tc, Sm, U, Np, and Pu are assumed to be present as anions (Table 6-16).

103 Table 6-16. Matrix porosities and effective diffusion coefficients in the rock matrix with different groundwaters (from Table 11-10 in Vieno & Nordman 1999). Parameter Porosity Effective diffusion coefficient D e Distance from fracture 0 1 cm 1 10 cm 0 1 cm 1 10 cm Species Brackish/ dilute and glacial groundwaters Saline groundwater Anions 0.1 % 0.2 % Neutral and cationic species 0.5 % 0.5 % Anions 0.02 % 0.04 % Neutral and cationic species 0.1 % 0.1 % Anions 10-14 m 2 s -1 5 10-14 m 2 s -1 Neutral and cationic species 10-13 m 2 s -1 10-13 m 2 s -1 Anions 10-15 m 2 s -1 5 10-15 m 2 s -1 Neutral and cationic species 10-14 m 2 s -1 10-14 m 2 s -1 Cl, Se, Mo, and I are assumed to be present as anions in all cases. Tc, Sm, U, Np and Pu as anions only in glacial water cases. The sorption data (K d ) for the far field (Table 6-17) are taken, for the most part, from the conservative data for non-saline and reducing conditions in TILA-99 (Table 11-9 in Vieno & Nordman, 1999). Exceptions to the use of TILA-99 values are the K d values for C and Mo. Carbon, which, as noted above, is assumed to be predominantly in methanic form and thus non-sorbing. There are no data available in the literature on which to base a K d value for Mo in crystalline rocks. The value given in Table 6-17 has been chosen by expert judgement, based on a study of sorption on illite by Motta and Miranda (1989), a comparison of the cation exchange capacities of illite and rock, and the known ph-dependency of K d values for sorption on kaolinite. 6.2.7 Dose conversion factors The analysis of calculation cases results in activity release rates, Bq/a as the unit. To show the impact of the releases in the biosphere, the release rates are converted to indicative dose rates using a set of dose conversion factors that are listed in Table 6-18. The dose conversion factors have been derived using a simple model of an indicative stylised drinking water well ("WELL-2008"), where ingestion of water is the only pathway considered. In a separate biosphere assessment, more elaborate models will be used, including the consideration of e.g. the accumulation of radioactivity in the biosphere, the groundwater flow rates that are orders of magnitude higher in the soil and above ground than in sparsely fractured bedrock, and other cases much more complex than the use of drinking water from a stylised kind of well. The multiplication of the WELL-2008 dose conversion factors with the annual activity release rates from the geosphere to the biosphere results in an estimate of the committed effective dose (ICRP, 1991) to an adult, obtained from ingestion of well water during one year.

104 Table 6-17. Distribution coefficients (K d ) for the rock matrix with various groundwater types (m 3 /kg). Element Dilute / brackish Saline Glacial C 0 s s Cl 0 s s Ni 0.1 0.005 s Se 0.0005 0.0001 s Sr 0.005 0.0001 s Zr 0.2 s s Mo 0.0005 0.0001 s Nb 0.02 s s Tc 0.05 s 0 Pd 0.001 0.0001 s Sn 0.001 0.0001 s I 0 s s Cs 0.05 0.01 s Sm 0.02 s 0 Ra 0.2 0.02 s Th 0.2 s s Pa 0.05 s s U 0.1 s 0.001 Pu 0.5 s 0.2 Np 0.2 s 0.002 Am 0.04 s s Cm 0.04 s s s = same as for brackish/dilute water

105 Table 6-18. Dose conversion factors in WELL-2008, based on the assumptions that the annual releases from the repository into the biosphere are diluted in 100 000 m 3 of water, and that an individual drinks 600 litres of contaminated water annually. In the case of decay chains, the dose conversion factors are for the parent radionuclide and the decay products are assumed to be in secular equilibrium. Secular equilibrium is a situation in which the quantity of a radioactive isotope remains constant, because its production rate, due e.g. to decay of a parent isotope, is equal to its decay rate. Radionuclide / decay products WELL-2008 dose conv. factor (Sv/Bq) C-14 3.48E-15 Cl-36 Ni-59 Ni-63 Se-79 Mo-93 Sr-90 5.58E-15 3.78E-16 9.00E-16 1.74E-14 1.86E-14 1.68E-13 Zr-93 Nb-93m 1.56E-14 Nb-94 Tc-99 Pd-107 1.02E-14 3.84E-15 2.22E-16 Sn-126 Sb-126 4.26E-14 I-129 6.60E-13 Cs-135 Cs-137 Sm-151 1.20E-14 7.80E-14 5.88E-16 Ra-226 Rn-222 Pb-210 Bi-210 Po-210 1.30E-11 Th-229 Ra-225 Ac-225 3.68E-12 Th-230 1.26E-12 Pa-231 Ac-227 Th-227 Ra-223 1.15E-11 U-233 3.06E-13 U-234 2.94E-13 U-235 Th-231 2.84E-13 U-236 2.82E-13 Np-237 Pa-233 6.65E-13 U-238 Th-234 2.90E-13 Pu-239 Am-241 Pu-240 Pu-242 1.50E-12 1.20E-12 1.50E-12 1.44E-12 Am-243 Np-239 1.20E-12 Cm-245 Pu-241 1.26E-12 Cm-246 1.26E-12

106

107 7 RESULTS 7.1 General The analysis results of the calculation cases are presented in this chapter in the following order: the calculation cases within the defective canister scenario (DCS) in Chapter 7.2, the additional scenario (AD) cases in Chapter 7.3, and supplementary cases addressing the behaviour of the repository system in Chapter 7.4. 7.2 Defective-canister calculation cases 7.2.1 Activity release rates from the near and far field Sh1 the base case The release rates of radionuclides from near and far field in case Sh1, that serves as the base case for comparison purposes, are presented in Table 7-1 and Figure 7-1. It can be observed that in case Sh1 the activity release rates are reduced below 0.001 Bq/a for all actinides except Pa-231, when the transport resistance in the far field is relatively high (WL/Q = 50 000 a/m). Actinides are typically strongly sorbing in the geosphere.

108 Table 7-1. Maximum activity release rates from the near and far field in the Sh1 case. The time of the peak release rate is denoted by t max, and it assumes values within the time window of 0 10 6 years. The expression 1.0E+06 implies that the cut-off limit of the calculations was reached before the exact peak release rate. In those cases the release rates are likely to continue to ascend after the time limit 17. Radionuclide Near field Far field Chain nuclides*** t max (a) Release rate (Bq/a) t max (a) Release rate (Bq/a) C-14 5.9E+03 1.5E+04 6.6E+03 1.5E+04 Cl-36 2.7E+04 7.6E+02 2.8E+04 7.6E+02 Ni-59 1.1E+05 4.0E+02 4.9E+05 2.8E+00 Ni-63 2.8E+02 2.2E+01 1.7E+02 6.1E-13 Se-79 Sr-90 4.0E+01 1.6E+04 8.7E+00 8.8E-11 Mo-93 2.5E+03 3.2E-03 4.6E+03 1.7E-03 Zr-93 Zr-93p 1.0E+06 6.0E-03 1.0E+06 3.4E-04 Nb-94 2.5E+04 5.2E-01 1.1E+05 6.9E-03 Tc-99 1.0E+06 1.6E-03 1.0E+06 3.0E-04 Pd-107 1.0E+06 1.5E-01 1.0E+06 1.4E-01 Sn-126 1.3E+05 7.9E-03 1.6E+05 6.5E-03 I-129 1.0E+06 1.5E+02 1.0E+06 1.5E+02 Cs-135 6.4E+05 3.0E+03 1.0E+06 1.6E+03 Cs-137 7.7E+01 1.2E+04 4.7E+01 6.1E-10 4N Pu-240 3.0E+04 9.6E-03 1.6E+04 3.8E-20 U-236 4N+1* See note 4N+2 Cm-246 Pu-242 1.0E+06 1.9E-01 1.0E+06 1.9E-06 U-isot** Th-230 7.9E+05 1.0E-03 1.0E+06 5.1E-07 Ra-226 1.0E+06 1.1E+01 1.0E+06 5.2E-07 4N+3 Am-243 5.0E+04 1.5E-03 1.5E+05 2.5E-09 Pu-239 8.5E+04 1.4E-01 8.6E+05 6.1E-16 U-235 Pa-231 6.6E+05 1.4E+00 9.3E+05 6.4E-03 * The release rates for the nuclides of chain 4N+1 (Cm-245, Am-241, Np-237, U-233 and Th-229 all remained under 1E-3 Bq/a. ** The release rates for the uranium isotopes U-238 and U-234 remained under 1E-3 Bq/a. *** 4N: Thorium Series, 4N+1: Neptunium Series, 4N+2: Uranium Series, 4N+3: Actinium Series. 17 The issue of the values ascending at the cut-off time of calculation is discussed further in Chapter 7.2.2.

109 Figure 7-1. Release rates from the near field (left) and far field (right) in Sh1 for two sets of radionuclides in case Sh1. Effect of defect size The defect diameter is 1 mm in Sh1, 4 mm in Sh4 and 100 mm in the Lh cases. The release rates taking into account different defect sizes with high flow are presented in Table 7-2. The chosen high flow case decreases the role of the far field, emphasizing the significance of the defect size.

110 Table 7-2. The effect of defect diameter on the near-field release rate maxima in cases with dilute/brackish water and high flow (DCS-II; see Figure 6-1). Radionuclide Sh1 Q (hole 1 mm) Sh4 Q (hole 4 mm) Lh Q (hole 100 mm) Chain nuclides t max (a) Release rate (Bq/a) t max (a) Release rate (Bq/a) t max (a) Release rate (Bq/a) C-14 3.1E+03 2.2E+04 3.1E+03 2.8E+05 1.0E+03 1.3E+07 Cl-36 1.6E+04 7.9E+02 1.3E+04 4.9E+03 1.0E+04 1.0E+05 Ni-59 5.9E+04 1.5E+03 5.9E+04 1.9E+04 1.0E+04 3.8E+06 Ni-63 2.8E+02 2.2E+02 2.8E+02 2.9E+03 2.8E+02 2.1E+06 Se-79 6.8E+03 1.3E+00 Sr-90 4.0E+01 1.5E+05 4.0E+01 2.0E+06 4.0E+01 1.3E+09 Mo-93 1.2E+03 5.5E-03 1.2E+03 3.5E-02 7.4E+02 1.4E+01 Zr-93 3.4E+05 7.5E-01 Zr-93p 6.0E+05 1.3E-02 5.9E+05 1.7E-01 3.7E+05 5.5E+01 Nb-94 2.3E+04 4.9E+00 2.3E+04 6.3E+01 1.9E+04 3.8E+04 Tc-99 1.0E+06 1.4E-02 1.0E+06 1.8E-01 1.0E+06 1.0E+02 Pd-107 1.0E+06 4.5E-01 1.0E+06 5.8E+00 1.0E+06 6.0E+02 Sn-126 1.2E+05 7.6E-02 1.2E+05 9.9E-01 1.0E+05 6.0E+01 I-129 1.0E+06 1.5E+02 1.1E+04 5.3E+02 2.4E+02 8.2E+04 Cs-135 3.3E+05 3.2E+03 3.9E+04 2.1E+04 1.7E+02 8.1E+04 Cs-137 7.7E+01 1.1E+05 7.7E+01 1.7E+06 6.5E+01 8.4E+08 4N Pu-240 3.0E+04 9.5E-02 3.0E+04 1.2E+00 2.6E+04 4.3E+02 U-236 1.0E+06 8.4E-03 4N+1 Cm-245 4.5E+04 1.8E-02 Am-241 4.5E+04 1.9E-02 Np-237 1.0E+06 4.2E+00 U-233 1.0E+06 3.3E-03 1.0E+06 4.3E-02 1.0E+06 1.9E+01 Th-229 1.0E+06 3.2E+00 4N+2 Cm-246 Pu-242 1.0E+06 1.8E+00 1.0E+06 2.3E+01 4.3E+05 3.7E+02 U-238 1.0E+06 7.2E-03 U-234 3.3E+05 1.5E-02 Th-230 1.0E+06 9.9E-03 7.7E+05 1.3E-01 1.0E+06 8.6E+01 Ra-226 8.6E+05 7.0E+01 9.0E+05 9.1E+02 3.0E+05 4.1E+05 4N+3 Am-243 5.0E+04 1.5E-02 5.0E+04 1.8E-01 4.0E+04 1.3E+00 Pu-239 8.5E+04 1.4E+00 8.5E+04 1.8E+01 5.5E+04 3.4E+03 U-235 Pa-231 6.1E+05 1.1E+01 8.6E+05 1.4E+02 1.0E+06 5.0E+02 The variation of the defect size from 1 mm to 4 mm resulted in increase of near-field release rates of most of the nuclides by a factor of 10...15, which is only somewhat less than the ratio of the defect areas (16). A further increase of the defect area by a factor of 625 leads to a smaller relative increase of release rates compared with the ratio of areas.

111 In the large hole case, the resistance of the hole would then become negligible and the other near-field processes would control the release rates. Effect of flow rate and geosphere transport resistance The effect of flow rate on near-field release rates and the effect of transport resistance on far-field release rates is presented in Table 7-3 and Figure 7-2 for the default-flow case Sh4 and the high-flow case Sh4 Q. The near-field flow rates are assumed to be in Sh4 Q ten times higher than in Sh4, and the transport resistance in Sh4 Q one tenth of that in Sh4 (Table 6-5, page 91).

112 Table 7-3. The effect of flow rate on near-field and far-field release rate maxima for Sh4 cases involving dilute/brackish water (DCS-II; see Figure 6-1). The defect diameter is 4 mm. Radionuclide Chain nuclides Near-field release rate maxima (Bq/a) Far-field release rate maxima (Bq/a) Sh4 Sh4 Q Sh4 Sh4 Q C-14 1.9E+05 2.8E+05 1.8E+05 2.8E+05 Cl-36 4.6E+03 4.9E+03 4.6E+03 4.9E+03 Ni-59 5.3E+03 1.9E+04 3.6E+01 1.1E+04 Ni-63 2.9E+02 2.9E+03 Se-79 3.6E-03 3.7E-03 3.2E-03 3.7E-03 Sr-90 2.1E+05 2.0E+06 2.7E+01 Mo-93 2.0E-02 3.5E-02 1.1E-02 3.3E-02 Zr-93 1.0E-03 2.3E-03 Zr-93p 7.7E-02 1.7E-01 4.4E-03 1.5E-01 Nb-94 6.8E+00 6.3E+01 9.0E-02 4.0E+01 Tc-99 2.1E-02 1.8E-01 3.9E-03 1.5E-01 Pd-107 1.9E+00 5.8E+00 1.8E+00 5.8E+00 Sn-126 1.0E-01 9.9E-01 8.5E-02 9.7E-01 I-129 5.2E+02 5.3E+02 5.2E+02 5.3E+02 Cs-135 9.2E+03 2.1E+04 2.9E+03 1.6E+04 Cs-137 1.7E+05 1.7E+06 4N Pu-240 1.3E-01 1.2E+00 U-236 4N+1* U-233 7.2E-03 4.3E-02 4N+2 Cm-246 Pu-242 2.5E+00 2.3E+01 U-isot** Th-230 1.3E-02 1.3E-01 Ra-226 1.4E+02 9.1E+02 4N+3 Am-243 1.8E-02 1.8E-01 Pu-239 1.8E+00 1.8E+01 U-235 Pa-231 1.8E+01 1.4E+02 8.3E-02 8.1E+01 * The release rates for the nuclides of chain 4N+1 (Cm-245, Am-241, Np-237, and Th-229 remained under 1E-3 Bq/a. ** The release rates for the uranium isotopes U-238 and U-234 remained under 1E- 3 Bq/a.

113 Figure 7-2. Release rates from the near (left) and far field (right) in Sh4 (default flow) and Sh4 Q (high flow). It can be observed in Figure 7-2 that the release rates of Ni-59 and Cs-135 are sensitive to variations in flow rate. For weakly and non-sorbing nuclides like I-129 the effect on the releases is minor. The high flow rate increases the release rates of sorbing nuclides from the near field up to a factor of 10. The effect on the far-field releases of relatively short-lived sorbing nuclides is most notable. Role of instant release from fuel matrix There is uncertainty associated to the instant release fractions of nuclides from fuel matrix and their influence on release rates. A case taking into account the IRF as the only release mode from fuel was calculated separately for the Sh1 case and another for the Lh Q case. See the selections for Sh1 Irf and Lh Q Irf in Table 6-1, page 82. The results are presented in Table 7-4 and Table 7-5 as well as in Figure 7-3 that shows the release rates of C-14 and I-129. The corresponding indicative dose rates are found in Table 7-7 and Table 7-8. In a situation of a large hole combined with high flow rates, the release rate is sensitive to IRF. In the Lh Q case the IRF dominates the maximum release rates from the near field more than in the Sh1 case, since nuclides can be released quickly to the geosphere

114 via the larger hole and with reduced film resistance between the stagnant water in bentonite and the water flowing in a fracture of the bedrock. In Figure 7-3, the four sources of C-14 can be clearly identified either from the three different C-14 curves or from the curve shapes: IRF, corrosion of Zircaloy ending at 10 000 a, corrosion of other metal parts ending at 1 000 a, and fuel matrix degrading at a rate of 10-7 a -1. Due to the short half-life of C-14 (5 730 years), also the effect of this source term fades away soon compared to the whole degradation time of fuel matrix. In the case of I-129, which typically dominates the WELL-2008 doses, the early dose rates are practically directly proportional to the IRF of I-129. In Figure 7-3 it can be observed that the IRF of I-129, which constitutes 5 % of the total inventory, dominates its maximum release rate in case of a large hole and high flow rates. The rest of its inventory in the fuel matrix is leached congruently at a constant rate of 10-7 a -1. Table 7-4. The role of the instant release from fuel matrix in near-field release rates, studied with two small-hole cases including either all release modes (Sh1) or only instant release fraction (Sh1 Irf), and assuming default flow (DCS-II, Figure 6-1). Radionuclide Sh1 Sh1 Irf t max (a) Release rate (Bq/a) t max (a) Release rate (Bq/a) C-14 5.9E+03 1.5E+04 4.4E+03 1.1E+03 Cl-36 2.7E+04 7.6E+02 2.2E+04 7.0E+01 Se-79 Sr-90 4.0E+01 1.6E+04 4.0E+01 1.6E+04 Tc-99 1.0E+06 1.6E-03 1.0E+06 1.6E-03 Pd-107 1.0E+06 1.5E-01 1.0E+06 1.5E-01 Sn-126 1.3E+05 7.9E-03 1.1E+05 7.4E-03 I-129 1.0E+06 1.5E+02 3.0E+04 8.4E+01 Cs-135 6.4E+05 3.0E+03 2.9E+05 1.9E+03 Cs-137 7.7E+01 1.2E+04 7.7E+01 1.2E+04

115 Table 7-5. The role of the instant release in near-field release rates, studied with two large-hole cases with high flow rates (DCS-II, Figure 6-1). Radionuclide Lh Q Lh Q Irf t max (a) Release rate (Bq/a) t max (a) Release rate (Bq/a) C-14 1.0E+03 1.3E+07 2.0E+02 2.0E+06 Cl-36 1.0E+04 1.0E+05 2.4E+02 7.2E+04 Se-79 6.8E+03 1.3E+00 7.1E+03 1.3E+00 Sr-90 4.0E+01 1.3E+09 4.0E+01 1.3E+09 Tc-99 1.0E+06 1.0E+02 9.8E+05 1.0E+02 Pd-107 1.0E+06 6.0E+02 1.6E+04 1.7E+02 Sn-126 1.0E+05 6.0E+01 2.2E+04 1.5E+00 I-129 2.4E+02 8.2E+04 2.4E+02 8.2E+04 Cs-135 1.7E+02 8.1E+04 1.7E+02 8.1E+04 Cs-137 6.5E+01 8.4E+08 6.5E+01 8.4E+08 Figure 7-3. The near-field release rates of C-14 and I-129 in case Lh Q Irf, where only the IRF inventory is handled, compared with the total release rates in Lh Q. Also the near-field release rate of C-14 from fuel matrix degradation can be observed. Effect of salinity The effect of groundwater salinity on release rates is presented in Table 7-6 and Figure 7-4. The near-field release rates were calculated for the cases Sh1 and Sh1 Sal (default flow) as well as for Lh Q and Lh Q Sal (high flow). As for the geosphere transport, the release rates were calculated only for the large-hole high-flow cases Lh Q and Lh Q Sal to study the combined effect of high flow rate and reduced sorption due to saline groundwater. Figure 7-4 presents the near-field and far-field release rates of Sr-90 and Ra-226 in cases Sh4 Q and Sh4 Q Sal. As can be seen, Sr and Ra are quite sensitive to salinity-related variations in groundwater chemistry, simulated here through the

116 variations of K d, diffusivity and porosity (Table 6-15 and Table 6-16). The release rates of short-lived sorbing radionuclides from the far field would be limited even in the high-flow case, but together with unfavourable saline chemistry the retardation of such radionuclides is thought to diminish substantially.

117 Table 7-6. The effect of salinity on near-field and far-field release rates. Both smalland large-hole cases were calculated for the near field and only large-hole cases for the far field (scenario DCS-II, see Figure 6-1). Radionuclide Chain nuclides Near-field release rate (Bq/a) Far-field release rate (Bq/a) Sh1 Sh1 Sal Lh Q Lh Q Sal Lh Q Lh Q Sal C-14 1.5E+04 1.5E+04 1.3E+07 1.3E+07 1.3E+07 1.3E+07 Cl-36 7.6E+02 7.6E+02 1.0E+05 1.0E+05 1.0E+05 1.0E+05 Ni-59 4.0E+02 1.2E+02 3.8E+06 1.7E+06 1.1E+06 1.5E+06 Ni-63 2.2E+01 2.3E+01 2.1E+06 2.1E+06 Se-79 1.3E+00 4.8E+03 1.2E+00 4.4E+03 Sr-90 1.6E+04 8.0E+04 1.3E+09 5.4E+09 1.7E+04 1.4E+09 Mo-93 3.2E-03 1.1E-03 1.4E+01 4.8E+00 1.2E+01 3.8E+00 Zr-93 7.5E-01 6.3E-01 6.2E-01 5.4E-01 Zr-93p 6.0E-03 4.1E-03 5.5E+01 4.7E+01 4.6E+01 4.0E+01 Nb-94 5.2E-01 8.5E-01 3.8E+04 6.2E+04 2.4E+04 3.8E+04 Tc-99 1.6E-03 2.4E-03 1.0E+02 1.4E+02 8.7E+01 1.2E+02 Pd-107 1.5E-01 1.6E-01 6.0E+02 6.0E+02 6.0E+02 6.0E+02 Sn-126 7.9E-03 9.6E-03 6.0E+01 8.5E+01 5.8E+01 8.4E+01 I-129 1.5E+02 1.5E+02 8.2E+04 8.2E+04 8.2E+04 8.2E+04 Cs-135 3.0E+03 3.1E+03 8.1E+04 1.7E+05 3.0E+04 8.8E+04 Cs-137 1.2E+04 5.5E+04 8.4E+08 4.2E+09 4N Pu-240 9.6E-03 1.3E-03 4.3E+02 1.2E+02 3.0E-01 8.5E-02 U-236 8.4E-03 5.5E-03 7.7E-03 5.1E-03 4N+1 Cm-245 1.8E-02 2.7E-01 4.9E-03 6.9E-02 Am-241 1.9E-02 2.8E-01 5.2E-03 7.3E-02 Np-237 4.2E+00 4.7E+00 3.4E+00 3.8E+00 U-233 1.9E+01 1.2E+01 1.5E+01 1.0E+01 Th-229 3.2E+00 1.6E+00 7.4E+00 4.9E+00 4N+2 Cm-246 Pu-242 1.9E-01 9.8E-03 3.7E+02 5.7E+02 2.2E+02 3.3E+02 U-238 7.2E-03 4.7E-03 6.7E-03 4.3E-03 U-234 1.5E-02 1.0E-02 1.3E-02 8.2E-03 Th-230 1.0E-03 1.6E-04 8.6E+01 1.3E+01 4.0E+01 6.2E+00 Ra-226 1.1E+01 4.0E+01 4.1E+05 7.5E+05 1.1E+02 7.9E+04 4N+3 Am-243 1.5E-03 2.6E-02 1.3E+00 2.3E+01 3.2E-01 5.3E+00 Pu-239 1.4E-01 8.9E-03 3.4E+03 7.6E+02 9.7E+01 2.6E+01 U-235 Pa-231 1.4E+00 1.3E+00 5.0E+02 5.0E+02 2.8E+02 2.8E+02

118 Note: "ns" refers to non-saline and "sal" to saline. Figure 7-4. The effect of salinity on Ra-226 and Sr-90 release rates from near field (left) and far field (right) in cases Sh4 Q and Sh4 Q Sal. The Ra-226 release rate from the far field falls in the non-saline case below the scale, 0.23 Bq/a at maximum. 7.2.2 Dose rates The indicative dose rates are calculated with the WELL-2008 dose conversion factors (see Chapter 6.2.7) which take into account only the dose from drinking of well water. They are obtained simply by multiplying the far-field release rate of each radionuclide with its dose conversion factor. The total dose rates in the base case Sh1 and for the most important radionuclides are presented in Figure 7-5. Due to the nature of the dose conversion factor approach, the dose rate curves follow the form of the far-field release rate curves. This can be seen by comparing Figure 7-5 to Figure 7-1. Figure 7-5. The total dose rate and the most important radionuclides contributing to it in the base case Sh1.

119 The cut-off limit of the calculations, one million of years in this work, is also the largest possible t max in the tables. In the cases, where t max is displayed as 10 6 a, the dose rate would as a rule continue to rise, if calculation was continued to a later point of time. The maximum values beyond one million years would, however, not essentially contribute to the conclusions of RNT-2008 analysis. Nevertheless, for confidence building it could be meaningful at least in some calculation cases of upcoming analyses to move the end of calculation forward in order to distinctly pass the maxima. In RNT-2008, the adopted degradation time of spent fuel matrix is usually ten million years and in a few exceptions one million years. From those values it can be inferred that the release rates and dose rates from principally matrix-bound long-lived nuclides can rise until the end of the period of either 10 7 or 10 6 years. A longer degradation time naturally corresponds with a lower release rate, since the source inventory to be finally released does not change in other ways than by radioactive decay. The storage capacities and retarding processes in the disposal system give additional length to the release pulse. I-129 (t 1/2 = 1.6 10 7 a) and Cs-135 (t 1/2 = 2.3 10 6 a) are examples of this scheme. While the dose rate of Cs-135 in Figure 7-5 has a rather steep slope still at 1 Ma, Figure 7-1 reveals that the maximum near-field release rate of Cs-135 is reached at about one million years and due to geosphere retention the maximum far-field release rate is reached somewhat later. The in-growth of daughter nuclides forms another source of late-appearing peak dose rates, as can be seen from Pa-231, a daughter of U- 235. The U-235 precipitate in a defective fuel canister produces Pa-231, the release rate and the dose rate of which continues to increase at one million years e.g. in case Lh Q. Besides extending release and transport calculations beyond one million years, it can be reasonable to use other kind of safety assessment approaches for exceedingly long time scales, if further elaboration of the issue altogether is regarded worthwhile. According to Guide YVL 8.4 (STUK, 2001) this could be carried out with complementary considerations "that may include e.g. bounding analyses by simplified methods, comparisons with natural analogues or observations of the geological history of the disposal site. The significance of such considerations grows as the assessment period of interest increases, and the judgement of safety beyond one million years can mainly be based on the complementary considerations." In Table 7-7, the dose rates for the most realistic and sensitivity cases of the defective canister scenario (DCS) are compiled, and Table 7-8 presents the dose rates for the what if cases of the scenarios DCS-I and DCS-II. It must be noted that the time at which the dose rate maximum is obtained, does not coincide with the time at which the individual nuclides achieve their peak dose rates as seen e.g. in Figure 7-5.

120 Table 7-7. The maximum total dose rate, obtained using the WELL-2008 dose conversion factors, and the three most important radionuclides contributing to the dose rate in the most realistic and sensitivity cases of the DCS-II scenario. In the first column there is the case identifier, in the second one the point of time at which the maximum total dose rate is obtained, in the third column the maximum total dose rate and then three most prominent radionuclides with their individual dose rates. Calculation case t max (a) Dose rate max (Sv/a) 1 st nuclide Small hole, defect diameter 1 mm from t = 0 a Dose rate (Sv/a) 2 nd nuclide Dose rate (Sv/a) 3 rd nuclide Dose rate (Sv/a) Sh1 10 6 1.2 10-10 I-129 1.0 10-10 C-14 5.2 10-11 Cs-135 1.9 10-11 Sh1-EPR 10 6 1.4 10-10 I-129 1.2 10-10 C-14 6.5 10-11 Cs-135 2.2 10-11 Sh1-VVER 7.9 10 3 1.1 10-10 I-129 8.5 10-11 C-14 5.6 10-11 Cs-135 1.6 10-11 Sh1 Fd 10 6 8.9 10-10 I-129 7.7 10-10 Cs-135 1.1 10-10 C-14 5.2 10-11 Sh1 Irf 2.3 10 4 5.6 10-11 I-129 5.6 10-11 Cs-135 9.7 10-12 C-14 3.8 10-12 Sh1 Q 10 6 2.0 10-10 I-129 1.0 10-10 C-14 7.7 10-11 Pa-231 7.2 10-11 Sh1 Sal 10 6 1.3 10-10 I-129 1.0 10-10 C-14 5.2 10-11 Cs-135 3.2 10-11 Sh1 Q Sal 6.8 10 1 3.7 10-8 Sr-90 3.7 10-8 Ra-226 2.9 10-10 I-129 1.0 10-10 Small hole, defect diameter 4 mm from t = 0 a Sh4 6.5 10 3 9.4 10-10 C-14 6.4 10-10 I-129 3.4 10-10 Cs-135 3.5 10-11 Sh4 Q 3.2 10 3 1.3 10-9 C-14 9.7 10-10 Pa-231 9.3 10-10 I-129 3.5 10-10 Sh4 Q Sal 6.8 10 1 4.9 10-7 Sr-90 4.9 10-7 Ra-226 3.7 10-9 C-14 9.7 10-10 (see the "what if" cases in Table 7-8) Table 7-8. The maximum total dose rate, obtained using the WELL-2008 dose conversion factors, and three most important nuclides contributing to the dose rate in the what if cases of the scenarios DCS-I and DCS-II. Calculation case t max (a) Dose rate max (Sv/a) 1 st nuclide Dose rate (Sv/a) 2 nd nuclide Large hole, defect diameter 100 mm from t = 10 000 and 100 000 a Dose rate (Sv/a) 3 rd nuclide Dose rate (Sv/a) LhB Q t4 1.1 10 4 1.1 10-8 C-14 7.1 10-9 I-129 4.4 10-9 Pa-231 3.2 10-9 LhB Q t5 1.0 10 5 4.5 10-9 I-129 4.4 10-9 Pa-231 3.2 10-9 Cs-135 3.5 10-10 Large hole*, defect diameter 100 mm from t = 0 a Lh Q 4.3 10 2 8.3 10-8 I-129 5.4 10-8 C-14 4.5 10-8 Pa-231 3.3 10-9 Lh Q Irf 2.5 10 2 6.4 10-8 I-129 5.4 10-8 C-14 6.8 10-9 Sr-90 2.8 10-9 Lh Q Sal 5.7 10 1 2.4 10-4 Sr-90 2.4 10-4 Ra-226 1.0 10-6 I-129 5.4 10-8 LhB Q 1.3 10 3 2.8 10-8 C-14 2.4 10-8 I-129 4.4 10-9 Pa-231 3.2 10-9 LhB Q Fd 10 6 3.5 10-8 Pa-231 3.2 10-8 C-14 2.4 10-8 I-129 4.5 10-9 * A canister with such a large initial defect would not pass the quality control. Thus these cases are highly improbable.

121 I-129, C-14 and Cs-135 constitute together the group of three most important radionuclides in the most realistic cases. Sr-90 and Ra-226 are the dominant radionuclides in cases with a combination of saline groundwater and high flow rate (the cases Sh1 Q Sal, Sh4 Q Sal in Table 7-7, and Lh Q Sal in Table 7-8). Among these results, Sr-90 is the only nuclide exceeding the regulatory limit 0.1 msv/a. However, its case is a fully hypothetical one, and also its dose rate would fall below the limit during an assumed 200 years' administrational post-control period. In addition, it should be noticed that the analysis does not employ an initial release delay that results from the saturation of bentonite buffer and allows sufficient decay time for the most shortlived nuclides. Before the saturation of bentonite, there will be no release process to speak of. Moreover, the mentioned regulatory limit is an expectation value, i.e. a product of the deterministic annual dose and the probability of its occurrence. Thus the deterministic values would translate to lower expectation values, if probabilities of calculation cases were considered. Probabilities are not evaluated in RNT-2008, and the presented results are deterministic ones. Pa-231 becomes the dominant nuclide in LhB Q Fd, the case of a large hole filled with bentonite, high flow rate and high fuel degradation rate. In the Sh1 cases with the defect diameter of 1 mm the time of dose maximum is typically at 10 6 years (the end-time of the calculations) and dominated by I-129. In the Lh (defect diameter 100 mm) cases the I-129 maximum comes earlier as the IRF is released quickly to the geosphere (because of the large hole). When the defect in Lh cases is filled with bentonite (LhB), C-14 becomes the dominant radionuclide, as it does not behave as an anion like I-129 (see low D e for anions in bentonite). In Figure 7-6, the influence of the defect size on the total dose rate is illustrated for all the cases with high flow and dilute/brackish water (i.e. non-saline water). The maxima before 10 000 years are dominated by C-14 and I-129. Figure 7-6. The effect of defect size on dose rates in cases with high flow and dilute/ brackish groundwater chemistry. The cases differ from each other only by the defect diameters: Sh1 1 mm, Sh4 4 mm, and Lh 100 mm.

122 7.3 Additional calculation cases 7.3.1 Cases in the AD-I scenario (Rock-shear/Earthquake) Table 7-9 shows the maximum dose rates in cases RS1, RS2 and RS3/RS3g with increased fracture flow and decreased transport resistance (case RS in Table 6-5) and in the base case Sh1 for comparison. In RS3 and RS3g it is assumed that the water in the adjacent fracture flows via the canister interior, the flow rate being 2 La -1. RS3g represents, in addition, glacial melt water conditions with a high fuel degradation rate of 10-6 a -1 (degradation rates in Table 6-3). The dose rate curves of RS1 and RS3g are presented in Figure 7-7. Table 7-9. The maximum total dose rate, obtained using the WELL-2008 dose conversion factors, and the most important radionuclides for the cases of the AD-I scenario (rock shear/earthquake). The results of Sh1 are shown for comparison. Calculation case t max (a) Dose rate max (Sv/a) 1 st nuclide Dose rate (Sv/a) 2 nd nuclide Dose rate (Sv/a) 3 rd nuclide Dose rate (Sv/a) Sh1 10 6 1.2 10-10 I-129 1.0 10-10 C-14 5.2 10-11 Cs-135 1.9 10-11 RS1 1.0 10 3 1.5 10-7 I-129 1.3 10-7 C-14 3.8 10-8 Ra-226 1.1 10-8 RS2 1.0 10 4 1.4 10-7 I-129 1.3 10-7 C-14 1.3 10-8 Ra-226 1.1 10-8 RS3 7.0 10 4 2.2 10-7 I-129 2.2 10-7 Ra-226 1.4 10-8 Pa-231 8.7 10-9 RS3g 7.0 10 4 2.2 10-7 I-129 2.2 10-7 Pa-231 8.7 10-8 Ra-226 1.7 10-8 Although the fuel degradation rate is ten times higher in RS3g than in RS3, the Ra-226 dose rate does not increase as much as that from Pa-231. This is due to the fact that the share of radium, which reaches the well, originates mostly from the Th-230 parent in the far field instead of coming directly from the fuel. In addition, Th will reach its solubility limit although rather late. 7.3.3 Cases in the AD-II and AD-III scenarios Table 7-10 shows the maximum dose rates in the AD-II scenario, where two timedependent cases are defined (Figure 6-4) and in the AD-III scenario, where gas expels radionuclides from the canister in water phase or gaseous form. See also the time dependence of dose rates in cases B Sh Lh Q and GASexW in Figure 7-7.

123 Table 7-10. The maximum total dose rate, obtained using the WELL-2008 dose conversion factors, and the most important radionuclides for four cases in the AD-II and AD-III scenarios. Calculation case t max (a) Dose rate max (Sv/a) 1 st nuclide Dose rate (Sv/a) 2 nd nuclide Dose rate (Sv/a) 3 rd nuclide Dose rate (Sv/a) Sh1 10 6 1.2 10-10 I-129 1.0 10-10 C-14 5.2 10-11 Cs-135 1.9 10-11 B Sh Lh Q 1.0 10 5 4.0 10-8 I-129 3.8 10-8 Pa-231 4.4 10-9 Ra-226 3.3 10-9 B Sh Lh Q g 1.0 10 5 4.6 10-8 I-129 4.0 10-8 Sn-126 5.5 10-9 Pa-231 4.4 10-9 GASexW 3.0 10 3 6.6 10-8 I-129 6.2 10-8 C-14 3.7 10-9 Cl-36 4.6 10-10 GASexG 9.0 10 2 6.3 10-7 C-14 6.3 10-7 The about eight times lower solubility limit of Th-230 in glacial water (Table 6-12) results in the B Sh Lh Q g case in a lower release rate of Ra-226 to the geosphere than in the B Sh Lh Q case. Note that the applied fuel matrix degradation rate is 10-7 a -1 in both cases. Pa-231 is not solubility limited in these cases and thus the dose rate would rise, if the fuel degradation rate in B Sh Lh Q g was increased in this analysis. The dose rate of Sn-126 becomes higher, as it appears as a non-sorbing anion in the near field in glacial water and is assumed to become more soluble. The near-field pulse of Ra-226 reaches the geosphere without its parent nuclide Th-230 only in saline groundwater conditions. In the GASexW case only the IRF of fission products is assumed to be released between the years 2 800 and 4 100 AP (Chapter 6.1.4). The second local maximum at 200 000 years in the figure results from Tc-99 and Cs-135, that have more advantageous farfield retention parameters. In case GASexG half of the total C-14 in the IRF, 8.9 10 8 Bq, is released to the geosphere instantaneously at 900 years. The release is assumed to have a duration of one year. In the geosphere, according to calculations with an analytical model (see Appendix C in Smith et al., 2007) the flux in the pulse decreases by a factor of 5, so the release rate at the well is 1.8 10 8 Bq/a. With the corresponding Well-2008 dose conversion factor (3.48 10-15 Sv/Bq in Table 6-18) the equivalent dose rate is 6.3 10-7 Sv/a.

124 Figure 7-7. The total dose rates four selected cases of the scenarios AD-I (rock shear/earthquake: RS1 and RS3g), AD-II (small hole enlarging in time due to weakened performance of bentonite: B Sh Lh Q), and AD-III (contaminated water expelled by gas from a canister: GASexW). 7.4 Supplementary cases illustrating the behaviour of system 7.4.1 Effect of variation in the near-field flow distribution The behaviour of the system regarding radionuclide transport was studied by varying the near-field flow rates. These variations are tied to 1) the existence of a fracture intersecting the deposition hole (Q F ), 2) the uncertainties in the performance of the backfill at the top of the deposition hole (Q DZ ), and 3) the uncertainties in the extent of the EDZ in the tunnel section as a connector of flow paths with the far field (Q TDZ ). A set of supplementary calculation cases was formed in order to study the impact of the changes (Table 7-11). The flow rates and transport resistance were varied typically by one order of magnitude and some flow rates by two. As mentioned in Chapter 6.1.5, these cases would belong to the defective canister scenario DCS-II, where an initial penetrating defect is considered. In these cases the effect of variation in the flow distribution is studied with defects of 1 mm diameter (Sh1 cases) and of 100 mm diameter (Lh cases), considering dilute/brackish groundwater and a spent fuel degradation rate of 10-7 a -1.

125 Table 7-11. Flow data for the near field in the supplementary calculation cases. Data for Lh, Sh1 and LhQ are the same as in Table 6-5. VhQ refers to very high values of Q DZ and Q TDZ. NoF refers to a situation in which there is no fracture intersecting with the deposition hole. Identifier Q F (L/a) Q DZ (L/a) Q TDZ (L/a) Transport resistance WL/Q (a/m) Lh, Sh1 0.2 2 10 50 000 Sh1 VhQ 0.2 200 1000 5 000 Lh Q 2 20 100 5 000 Lh NoF 0 2 10 50 000 Lh Q NoF 0 20 100 5000 Lh VhQ NoF 0 200 1000 5 000 In the analysis of these cases solubility, sorption, and data for buffer and backfill (Table 6-12, Table 6-14, and Table 6-15) for four radionuclides with two neutral species (Ra-226, C-14), one anion (Cl-36) and one cation (Cs-135), and for two non-sorbing generic nuclides (very long-lived, neutral and anion) are used. A unit (delta pulse) release of 1 GBq to the canister interior is assumed for the generic neutral and anionic radionuclides.

126 Table 7-12. Maximum release rates from the near field and the percentage of the three release routes corresponding to Q F, Q DZ, and Q TDZ (Figure 6-5). N.B the percentages are at the time of the maximum release rate (t max ). In the cases of the fictive neutral and anionic species, a unit release of 1 GBq was assumed. Neutral Ra-226 C-14 Anion Cl-36 Cs-135 Sh1 t max (a) maximum (Bq/a) - bentonite 1 (%) - backfill in hole 2 (%) - backfill in tunnel 3 (%) 1.7 10 4 2.7 10 3 8 30 62 1.0 10 6 1.1 10 1 45 53 2 5.9 10 3 1.5 10 4 10 33 57 3.1 10 4 1.5 10 3 41 49 10 2.7 10 4 7.6 10 2 41 49 10 6.5 10 5 3.0 10 3 5 23 73 Sh1 VhQ t max (a) maximum (Bq/a) - bentonite 1 (%) - backfill in hole 2 (%) - backfill in tunnel 3 (%) 9.0 10 3 2.8 10 3 5 94 1 1.0 10 6 7.9 10 1 6 94 0 3.5 10 3 2.1 10 4 5 94 1 2.7 10 4 1.5 10 3 36 64 0 2.5 10 4 7.7 10 2 36 64 0 3.4 10 5 3.2 10 3 2 94 4 Lh t max (a) maximum (Bq/a) - bentonite 1 (%) - backfill in hole 2 (%) - backfill in tunnel 3 (%) 4.6 10 2 4.3 10 5 17 56 27 2.6 10 5 5.6 10 4 47 51 2 1.0 10 3 4.1 10 6 14 46 39 8.7 10 2 4.3 10 5 44 52 4 1.0 10 4 9.0 10 4 41 49 10 5.4 10 4 1.1 10 4 5 24 71 Lh Q t max (a) maximum (Bq/a) - bentonite 1 (%) - backfill in hole 2 (%) - backfill in tunnel 3 (%) 2.0 10 2 2.2 10 6 41 54 5 3.0 10 5 4.1 10 5 57 42 1 1.0 10 3 1.4 10 7 39 50 1 2.4 10 2 1.4 10 6 92 8 0 1.0 10 4 1.0 10 5 81 19 0 1.7 10 2 8.1 10 4 100 0 0 Lh NoF t max (a) maximum (Bq/a) - backfill in hole 2 (%) - backfill in tunnel 3 (%) 1.0 10 3 3.8 10 5 42 58 2.4 10 5 3.0 10 4 96 4 1.4 10 3 3.6 10 6 48 52 1.3 10 3 2.7 10 5 89 12 1.0 10 4 7.5 10 4 83 17 5.9 10 4 1.1 10 4 25 75 Lh Q NoF t max (a) maximum (Bq/a) - Q DZ hole (%) - Q TDZ tunnel (%) 2.8 10 2 1.5 10 6 87 13 2.6 10 5 1.9 10 5 97 3 1.1 10 3 1.0 10 7 81 19 8.7 10 2 3.7 10 5 99 1 1.0 10 4 8.2 10 4 99 1 5.2 10 3 5.5 10 4 83 17 Lh VhQ NoF t max (a) maximum (Bq/a) - backfill in hole 2 (%) - backfill in tunnel 3 (%) 2.0 10 2 2.1 10 5 99 1 2.7 10 5 4.1 10 5 100 0 1.0 10 3 1.2 10 7 99 1 8.7 10 2 3.8 10 5 100 0 1.0 10 4 8.3 10 4 100 0 3.1 10 3 1.3 10 5 97 3 1) Route as with the equivalent flow Q F, from the bentonite in the deposition hole into rock fissures; 2) As with Q DZ, from the backfill in the top of the deposition hole into damaged rock zone; 3) As with Q TDZ, from the tunnel section above the deposition hole to the geosphere.

127 Some observations from Table 7-12 are provided below: In the Sh1 cases, the hole is the limiting factor. The increase of Q DZ and Q TDZ (Sh1 VhQ) compared to the base case affects only the Ra-226 release rate (79 Bq/a versus 11 Bq/a) and the release distribution along these routes. Ra-226 is solubility limited and has a low K d value in bentonite, so the concentration profile reaches the upper part of the deposition hole and the backfill in the hole by diffusion and is released from the backfill in the hole along route Q DZ to the geosphere. In the Lh cases, the maximum release rate of Ra-226 occurs approximately, when it reaches the solubility limit. The two other non-anionic species, Neutral and C-14, are assumed to be nonsorbing and behave in a different way from Ra-226, as they are not solubility limited. The differences are biggest in the Sh1 case. The maximum release rate of C-14 occurs, when corrosion of metal parts ends or somewhat later in the Sh1 case. The IRF of Cs-135 governs the early behaviour of the release and can be transported fast into the fracture, if the flow rate Q F is high and the release from the canister is not hindered by a small hole (Lh Q). Releases along the two other routes take more time. When the defect is smaller, the IRF release from the canister is smoothed out more effectively. To get even better understanding on the relative release rates, the equivalent flow rates from the near field were calculated for three generic virtually stable (very long-lived i.e. radioactive decay is not accounted for) elements: Neutral, Anion and Cation (as Cs in Table 6-14). Diffusion data from Table 6-14 are applied here, and because sorption does not play a role in the steady state situation, the K d was set to zero. Inside the canister, a constant solubility limit (or concentration) of 1 unit per litre is assumed. The calculated activity or mass flow rate (release rate) from the near field to the geosphere is thus the equivalent flow rate multiplied by the concentration in canister interior. Cases Sh1, Sh4 and LhB Q with defect diameters of 1 mm, 4 mm and 100 mm were chosen for this numerical experiment and the results are presented in Table 7-13. Table 7-13. The behaviour of constant concentration inputs of virtually stable isotopes. The equivalent flow rates from the canister interior to the geosphere (= Q ekv ) and time points when the output rates reach 90 % of the maximum rates. Modelled case t 90max (a) Neutral Anion Cation Q ekv (L/a) t 90max (a) Q ekv (L/a) t 90max (a) Q ekv (L/a) Sh1 8.3 10 3 0.00091 1.2 10 4 0.00050 7.1 10 3 0.00096 Sh4 8.3 10 3 0.012 1.2 10 4 0.0032 7.1 10 3 0.014 Lh Q 1.0 10 3 3.2 8.7 10 2 1.1 8.7 10 2 4.6 LhB Q 2.3 10 3 0.53 2.3 10 3 0.048 1.4 10 3 1.3

128 In cases Sh1 and Sh4, the defect size limits transport resistance effectively. For Anion, bentonite plays a more important role as a barrier due to low diffusivity. Thus the increase in the release rate for anions is not as big as for Neutral and Cation, when the hole becomes larger.

129 8 OVERVIEW OF RESULTS AND DISCUSSION 8.1 General The results of the RNT-2008 analysis essentially represent the highest plausible radionuclides release rates. Additionally, the results illustrate the functions and roles of the individual barriers of the multi-barrier system. The outcome of the analysis is mainly presented as activity release rates at the following two boundaries: 1) from the near field to the geosphere and 2) from the geosphere to the biosphere. Beside release rates, indicative dose rates are presented for the most important calculation cases. They were calculated by assuming the use of a stylised drinking water well in the vicinity of the repository, by applying the WELL-2008 dose conversion factors. The principal dose rate results for Posiva's safety case will be produced using the release rates from this analysis in Posiva's upcoming revision of the biosphere assessment. The assessed calculation cases, including sensitivity and "what if" cases, show that the disposal system is robust and would tolerate large deviations from the expected conditions without drastic effects on the releases into the biosphere. The key barriers of the planned disposal system are the ceramic and metallic waste forms, the copper-iron canister, the bentonite buffer, and the bedrock. The key phenomena contributing to the performance of the disposal system are regarded to be low flow rates of groundwater, slow corrosion of copper, low dissolution rates of spent fuel (reduced by reducing chemical conditions) and fuel assembly materials, low solubilities of several of the most hazardous radionuclides, slow transport of radionuclides in bentonite buffer (advection and colloidal transport avoided; sorption provided), and slow transport in the host rock particularly adjacent to the near field (limited groundwater flow, diffusion into the rock matrix, sorption in the rock matrix).

130 The particularly conservative features of modelling and data include the assumptions of the near-field transport model: o an initially open hole through the copper overpack o no delay of release provided by the cast iron insert o no sorption or co-precipitation of radionuclides with corrosion products of iron o high flow rates through the backfill at the top of the deposition hole and in the tunnel low transport resistance (WL/Q) in the far-field transport. matrix diffusion is the only phenomenon assumed to cause retardation and dispersion in the far-field transport. Sorption on fracture fillings and diffusion into stagnant water pools in the fractures are not taken into consideration. pessimistic K d values. The following subchapters discuss the most important factors affecting the results, the uncertainties associated with the assessment, and the regulatory compliance. In addition, RNT-2008 is compared with its predecessor TILA-99 and the recent KBS-3H transport analysis. 8.2 The most important factors affecting the results The release rates of radionuclides to the biosphere depend on many parameters. The main factors affecting the results are presented in Table 8-1 with cases where the relevant factors are pointed out together with explanatory comments.

131 Table 8-1. Main factors, relevant cases and comments concerning factors affecting the analysis results. Factor See cases Comment Defect size Table 7-2, Table 7-7 and Table 7-8 IRF Fuel degradation rate Fuel type Salinity Table 7-4 and Table 7-5 Table 7-7 and Table 7-8: Sh1 Fd LhB Q Fd Table 7-7 and Table 7-8: Sh1 EPR Sh1 VVER Table 7-6 Figure 7-4 Flow rate Table 7-3 Figure 7-2 WL/Q Onset of solubility limit In cases where the diameter is 4 mm, the release rates of Cs and neutral species are about 10 times higher than in cases, where the diameter is 1 mm. The ratio of the areas of these defects is 16, which shows that other barriers have a bigger role in the 4 mm case. In cases where the diameter is 100 mm, the release rates are naturally even higher, but the roles of other barriers also even more significant. When the defect is large, the IRF is more dominant, see e.g. I-129 compared with long-lasting releases. Uncertainties in IRF are reflected directly proportionally to the maximum release rates. The release rate of Pa-231 depends linearly on the fuel degradation rate in the LhB Q Fd case, but not in Sh1 Fd case as Pa-231 is solubility limited in this case and because the small size of the defect causes accumulation of Pa in the canister's water volume. The smaller inner volume relative to the amount of fuel in the EPR case (see also Table 6-6) causes more rapid increase of concentrations and thus higher release rates of I-129. In the VVER case the smaller ratio of fuel to inner volume leads to lower release rates. The strongest effect is the higher release rates of Ra-226 and Sr-90 from the far field because of reduced retardation in saline high-flow conditions. The effect is most notable for short-lived sorbing nuclides in the far-field transport. Release rates of cations from the near field are typically one order of magnitude higher with high flow rate. The effect is much lower with anions, because the diffusion resistance in the near field dominates overall transport. Appendix 1 shows the effect of far-field parameter variations. The values of the lumped parameter u can be interpreted as variations of WL/Q. Depends on the degradation rate of fuel matrix, type of radionuclide, potential parent nuclide, other isotopes of the element in question and on equivalent flow rate from canister interior to bentonite. See more details below, in Table 8-2. The importance of solubility limits depends strongly on the radionuclide and the assumptions for the calculation case. Further information on the onset of solubility limits of elements is presented in Table 8-2.

132 Table 8-2. Onset of solubility limits. Sh stands for small defects, Lh for large defects and RS for rock shear cases. Nuclide Solubility limit reached Comment on cases U, Zr, Np Always All cases and assumptions. Tc, Se Nearly always Se-79 not limited in case Lh Q Sal. Tc-99 not anymore limited after 900 000 years in Lh Q Irf (only IRF included). Ra Mostly Ra-226 becomes solubility limited between 30 000 300 000 years depending on the case. In B Sh Lh after the enlargement of the defect it is no longer solubility limited. Pu Mostly In all Sh cases solubility limited. In Lh cases Pu is limited for hundreds of thousands of years only in LhB Q cases when the defect is filled with bentonite and in the Lh Q Sal case. In RS cases and B Sh Lh cases Pu is solubility limited only in the RS3g case with low solubility in glacial water. Th Mostly In the B Sh Lh case after the enlargement of the defect it is no longer solubility limited. Pa Am In Sh cases only In Sh cases only Pa is solubility limited later in time due to in-growth and release from the fuel. The earliest occurrences are in the cases of high fuel degradation rate. In the Sh1 Fd case Pa becomes solubility limited at 55 000 years. Am becomes solubility limited after a few thousand of years, but it is not solubility limited at much later times, e.g. in Sh1 Fd with high fuel degradation rate Am is not solubility limited after 60 000 years. Am isotopes have short half-lives in these timescales. Ni Mostly In Lh and RS cases with non-glacial chemistry, Ni is solubility limited until a time range of 7000 100 000 a, depending on the case. Pd Mostly Not solubility limited in Lh cases, when the defect is filled with water and RS cases late in time, e.g. 5000 20 000 a. Mo Mostly Solubility limited in glacial chemistry (cases RS3g and B Sh Lh Q g) only for a few thousands of years. Sr Mostly Not solubility limited in saline chemistry and Lh cases after 90 500 years. Sr-90 was calculated only over a period of 2000 years, due to its relatively short half-life. Sn Mostly In Lh and RS cases Sn is not solubility limited after some time which varies depending on the case. Not solubility limited in RS3. Nb Mostly Not solubility limited in the RS3g case and at later times in Lh cases. In general, the most important nuclides in most cases are I-129, C-14, Cs-135 and Pa- 231, in this order, calculated with the dose conversion factors of WELL-2008. Principal aspects affecting these nuclides, in addition to their relatively long half lives, are: I-129: C-14: high IRF and weak retardation large inventory in metal parts is released quickly; weak retardation Cs-135: high IRF plus fairly weak retardation Pa-231: fuel degradation rate and solubility limit important only in cases with high flow. The in-growth from its parent U-235 is not a major factor.

133 8.3 Uncertainties Sufficient confidence on safety is a prerequisite for the licensing of a nuclear waste repository and calls for management of uncertainty. "Uncertainty treatment" is the term for a subset of uncertainty management, and represents what is needed in assessing the repository performance. Uncertainty arises from imperfect knowledge of the system to be assessed and its evolution, and is inherent to all safety cases. Hence, adequate uncertainty treatment is a cornerstone of any safety case and shall be considered ubiquitously within the programmes. The complexity of the phenomena of concern and the temporal and spatial scales under consideration make the management of uncertainties exceptionally demanding in this context. In addition, some key parameters applied in the transport analysis cannot in practice be directly determined in natural conditions and in the long term required. RNT-2008 is a deterministic analysis, where the data for each repository calculation case are individually specified, i.e. not sampled from probabilistic distributions. However, some of the chosen input data, as the WL/Q values, were produced based on probabilistic methods. In Finland the regulatory criteria are not based on risk criteria and do not as such require the use of probabilistic distributions. The uncertainties are scoped in RNT-2008 firstly with wide-range modelling assumptions and secondly with parameter variations. The assumptions are purposely conservative, meaning that they shall ensure that the results, with high degree of certainty, overestimate radioactive releases and radiation exposures. The analysis includes calculation cases that demonstrate the robustness of the repository design, illustrate the relative importance of the components of the multi-barrier system, examine the sensitivity to variations of key parameters, and also explore the system's behaviour in extreme situations. Such sensitivity cases and hypothetical "what if" cases form an essential part of the assessment, and are also called scoping and bounding calculation cases. Cases giving a realistic best estimate of the performance of the waste management system were not calculated in this analysis. Their results would go below the activity release rates of the present calculated cases. The formed calculation cases give a quantitative, albeit not especially precise picture on uncertainties related to various processes and events, and the features (FEPs). The uncertainties stemming from the formulation of scenarios and calculation cases are closely tied up with the understanding of the system. In RNT-2008, the number of calculation cases is kept small by allowing wide variability of assumptions and parameters in order to cover all potential evolutional FEPs of the repository. Thus the calculation cases also cover consequences of FEPs that are not explicitly referred to in connection with the calculation case definitions. An important category of uncertainty is bound to conceptual modelling. Conceptual uncertainty relates to the understanding of the nature of processes involved in repository evolution. This concerns not only the mechanistic understanding, but also how well they are represented in simplified mathematical models. The real processes have often to be considerably simplified for the numerical analysis by forming conceptual models representing the real processes and to allow their mathematical formulation. Good

134 understanding of the most important subprocesses is a central prerequisite for satisfactory treatment of this type of uncertainty. In RNT-2008 the conceptualizations within both near and far field were kept at a quite plain level, emphasizing transparency. The uncertainties associated to numerical models can be reduced by using models that are fit-for-purpose, highly validated, and have qualified input data within reach. The near-field transport model REPCOM and the far-field transport model FTRANS used in this work have gained trust during their long history of usage and participation in many benchmarking exercises. The consequences of uncertainties connected to parameter values (input data) used in the calculations are handled by selections aimed to be conservative. The exchange of information between many organisations on this area provides a rather large reference data base for such central parameters like sorption factors and diffusivities. Uncertainties caused by using generic data can be reduced by further increasing the number of site-specific determinations. Examples of current data with high uncertainties are the solubility limits of some relevant radionuclide-containing chemical species. The numerical testing of the role of the various parts of the repository system not only builds understanding of the system's behaviour but can also be utilized in a graded approach, where the efforts in the research and technical development programmes are focused to the actions effectively serving the progress. The focus can be directed elsewhere, if the uncertainty in a particular process can be proved irrelevant, i.e. uncertainty is not important to safety, because, for example, safety is controlled by other processes. 8.4 Regulatory compliance Results of a selected set of calculation cases are presented and compared with the regulatory constraints of Guide YVL 8.4 (Table 2-1). The occurrence time of the maximum total activity release rate and the ratio of the maximum to the regulatory limit are given first. The regulatory guide gives dose rates as constraints for an initial time period of several thousands of years and, for longer times, activity release rates of specific radionuclides 18. When comparing dose rates to the limit, the calculated release rates are converted to dose rates using the WELL-2008 dose conversion factors. Dose conversion factors over multiple pathways will be used in the biosphere assessment, that takes the radionuclide release rates from this transport analysis as an input. In the table, dose or release rate ratios are also given for the three most important nuclides at times of their own individual maxima. 18 For Mo, an activity release constraint of 3 GBq/a is assigned as presented in Chapter 2.

135 Table 8-3. Maximum total activity release rates (moving averages over one thousand years) from the geosphere to the biosphere compared with the release rate constraints of YVL 8.4 and the corresponding ratios of the three most important nuclides at their individual maxima. See also the results of the "what if" cases in Table 8-4. Note that these sets of cases do not include all the calculation cases presented in this report. Calculation case t max (a) Max ratio* of release rates 1 st nuclide Ratio** of release rates 2 nd nuclide Ratio of release rates 3 rd nuclide Ratio of release rates Sh1 6.5 10 3 5.1 10-5 C-14 4.9 10-5 Cs-135 5.3 10-6 Cl-36 2.5 10-6 Sh1 Fd 6.5 10 3 5.1 10-5 C-14 4.9 10-5 Cs-135 3.1 10-5 I-129 1.2 10-6 Sh1 Q 3.2 10 3 7.5 10-5 C-14 7.3 10-5 Cs-135 1.0 10-5 Cl-36 2.6 10-5 Sh1 Sal 6.5 10 3 5.1 10-5 C-14 4.9 10-5 Cs-135 8.9 10-6 Cl-36 2.5 10-6 Sh1 Q Sal 3.8 10 3 7.5 10-5 C-14 7.3 10-5 Cs-135 1.1 10-5 Cl-36 2.6 10-6 Sh4 6.5 10 3 6.1 10-4 C-14 6.1 10-4 Cl-36 1.5 10-5 Cs-135 9.7 10-5 Sh4 Q 3.2 10 3 9.2 10-4 C-14 9.2 10-4 Cs-135 5.3 10-5 Cl-36 1.6 10-5 Sh4 Q Sal 3.2 10 3 9.4 10-4 C-14 9.2 10-4 Cs-135 7.9 10-5 Cl-36 1.6 10-5 RS1 1.7 10 3 3.1 10-2 C-14 3.1 10-2 I-129 9.5 10-4 Cl-36 4.6 10-4 RS2 1.1 10 4 1.1 10-2 C-14 1.0 10-2 I-129 9.4 10-4 Cl-36 4.5 10-4 RS3g 7.1 10 4 1.5 10-3 I-129 1.1 10-3 Cl-36 4.5 10-4 Pa-231 2.5 10-4 B Sh Lh Q 1.0 10 5 1.5 10-3 Cl-36 1.0 10-3 I-129 4.6 10-4 Cs-135 1.3 10-4 B Sh Lh Q g 1.0 10 5 1.6 10-3 Cl-36 1.1 10-3 I-129 4.9 10-4 Cs-135 1.3 10-4 (see the "what if" cases in Table 8-4) * The maximum of the total activity release rate of all nuclides / release rate constraint (= calculated release rate / regulatory limit). ** The maximum of the activity release rate of an individual nuclide / release rate constraint. Table 8-4. An extension to Table 8-3 with the what if cases of the scenarios DCS-I and DCS-II. Calculation case t max (a) Max ratio of release rates 1 st nuclide Ratio** of release rates 2 nd nuclide Ratio of release rates 3 rd nuclide Ratio of release rates LhB Q t4 1.2 10 4 6.4 10-3 C-14 6.3 10-3 Cl-36 1.6 10-4 Cs-135 9.8 10-5 LhB Q t5 1.1 10 5 2.3 10-4 Cl-36 1.3 10-4 Cs-135 9.6 10-5 I-129 6.6 10-5 Large hole*, defect diameter 100 mm since t = 0 a Lh Q 8.8 10 2 3.8 10-2 C-14 3.7 10-2 I-129 6.6 10-4 Cl-36 3.4 10-4 Lh Q Sal 8.8 10 2 3.8 10-2 C-14 3.7 10-2 Ra-226 2.6 10-3 I-129 6.6 10-4 * A canister with such a large initial defect would not pass the planned quality control, and thus these cases are highly improbable.

136 Figure 8-1. Maximum total activity release rate from the geosphere to the biosphere in relation to the release rate constraints of YVL 8.4 for a selection of calculation cases (Ratio = calculated release rate / regulatory limit). Case Lh Q is a very hypothetical "what if" case, assuming a 100 mm hole in the canister at the time of the repository closure. 8.5 Comparison to past safety assessments and KBS-3H Compared with TILA -99, there are a number of changes in the near-field modelling described in this report. The major changes in the modelling data are listed below, although the effects are not further analysed in this work: The solubility limits are lower in RNT-2008 than in TILA-99; K d values in buffer and backfill in RNT-2008 are conservative data from SR- Can, and mostly higher than in TILA-99; The effective diffusion coefficients in buffer are two times higher for anions and about ten times lower for cations (Cs). The K d values of Cs are lower. For neutral species the effective diffusion coefficients are 20 % higher; Fuel degradation rate is lower than in TILA-99; C-14 is not assumed to exist as an anion as in TILA-99; The radionuclide inventories have been updated, but the changes are minor. The flow rate Q TDZ in the tunnel is by a factor of 10 lower than in TILA-99; The diameter of the smallest hole or penetrating defect was 2.5 mm in TILA-99 in contrast to the values of 1 mm or 4 mm in the small hole cases of this work; In the small hole cases of TILA-99, only 1 % of bentonite buffer and no backfill was included in the conceptual model (see Fig. 11-4 in TILA-99), whereas in this work 90 % of bentonite and also the backfill is included. The computer modelling is true two dimensional in this work; The far-field model is basically similar to that in TILA-99.

137 8.5.1 Comparison with the small hole case of TILA-99 The dose rates from the small hole cases of TILA-99 and the most similar cases (analogical assumptions) of this work are compared in Table 8-5. Table 8-5. The maximum total dose rate, obtained using the WELL-2008 dose conversion factors, and the most important nuclides in the defective canister cases. Calculation case t max (a) Dose rate max (Sv/a) 1 st nuclide Dose rate (Sv/a) 2 nd nuclide Dose rate (Sv/a) 3 rd nuclide Dose rate (Sv/a) Sh1 10 6 1.2 10-10 I-129 1.0 10-10 C-14 5.2 10-11 Cs-135 1.9 10-11 Sh1 Fd 10 6 8.9 10-10 I-129 7.7 10-10 Cs-135 1.1 10-10 C-14 5.2 10-11 SH50ns TILA-99 8.4 10 5 1.6 10-9 I-129 1.3 10-9 Cs-135 2.7 10-10 C-14 1.1 10-10 Sh4 6.5 10 3 9.4 10-10 C-14 6.4 10-10 I-129 3.4 10-10 Cs-135 3.5 10-11 The cases Sh1 Fd in this work and SH50ns in TILA-99 have about the same release rates from fuel elements. For Cs-135, the K d value is lower in this work, but D e is higher. C-14 is not assumed to be an anion in this work. For the anion I-129, the D e and porosity in the buffer are higher in this work. In addition, the size of the defect in Sh1 is about 6 times smaller than in TILA-99. These factors explain the differences in the dose rates between TILA-99 and this work. 8.5.2 Comparison of the disappearing canister case of TILA-99 and PD-BC case of KBS-3H analyses In TILA-99, it was assumed that the canister will fully lose its isolation capability at 10 000 years in the case DC-50ns (median flow and non-saline chemistry). The base case of this work, Sh1, is quite different. Thus, it is reasonable to select for comparison the base case PD-BC of KBS-3H (Smith et al., 2007), which is close to DC-50ns, and to modify case Sh1 of RNT-2008 to make it more analogous to PD-BC. In particular, in PD-BC of KBS-3H it was assumed that: at 1 000 years, a defect of 1 mm in diameter initiates a pathway; at 10 000 years, the defect is enlarged and loses its resistance completely; the canister interior is in contact with bentonite by diffusion over a height of 35 cm at the top of the canister; otherwise the data are as in case Sh1 of this work. N.B.: In PD-BC of the KBS-3H analyses there was only one release route to geosphere, namely Q F to the fracture intersecting the hole. In calculating a corresponding PD-BC for KBS-3V, the release routes at the upper part of the deposition hole, Q DZ, and in the tunnel, Q TDZ, are included (see Figure 6-5).

138 The results from Sh1, after its adaptation to correspond with PD-BC, from DC-ns50 of TILA-99, and from PD-BC of KBS-3H are presented in Table 8-6 and Figure 8-2. Discussion on the results and their differences: The reason for the second nuclide Sn-126 to appear in the results of TILA-99 is that it was modelled as an anion. The diffusion data for I-129 differs from TILA-99. C-14 was assumed to behave as an anion in TILA-99. The maximum dose rate in KBS-3H PD-BC is smaller than in Sh1 as PD- BC, since as mentioned above the KBS-3H model did include neither Q DZ nor Q TDZ and in the RNT-2008 modelling of KBS-3V the effect of excavation damaged zone (EDZ) implicitly affects the chosen near-field flow rate for the tunnel (Q TDZ ). Towards the end of calculation time, the dose rate in all the cases is dominated by I-129. The release rate of I-129 slowly reaches the degradation rate of fuel as the IRF is dissipated. In TILA-99, the fuel degradation is completed at 960 000 years, which can be seen from the curve of Figure 8-2. Table 8-6. The maximum total dose rate, obtained using the WELL-2008 dose conversion factors, and the most important nuclides in rather corresponding cases of different performance assessments. Case t max (a) Dose rate max (Sv/a) 1 st nuclide Dose rate (Sv/a) 2 nd nuclide Dose rate (Sv/a) 3 rd nuclide RNT-2008: Sh1, adapted to correspond to PD-BC of KBS-3H analyses (size of the defect increases at 10 000 years) Dose rate (Sv/a) Sh1 as PD-BC 1.0 10 4 2.4 10-8 I-129 1.4 10-8 C-14 7.9 10-9 Cl-36 1.1 10-9 TILA 99: DC-ns50 1.1 10 4 3.8 10-8 I-129 3.5 10-8 Sn-126 9.7 10-9 C-14 5.3 10-9 KBS-3H: PD-BC 1.0 10 4 9.6 10-9 I-129 7.9 10-9 C-14 1.2 10-9 Cl-36 5.9 10-10

139 Figure 8-2. Dose rates in the comparison cases KBS-3H PD-BC, TILA-99 DCns50 and Sh1 as PD-BC that represent different performance assessments. 8.5.3 The rock shear/earthquake case RS3g of RNT-2008 and the postglacial fault (DC-pgf) scenario of TILA-99 The maximum WELL 2008 doses in the pgf scenario of TILA-99 and RS3g of RNT- 2008 are shown in Table 8-7. Table 8-7. The maximum total dose rate, obtained using the WELL-2008 dose conversion factors, and the most important nuclides in two cases of RNT-2008 and TILA-99 for comparison. Calculation case RNT-2008: t max (a) Dose rate max (Sv/a) 1 st nuclide Dose rate (Sv/a) 2 nd nuclide Dose rate (Sv/a) 3 rd nuclide Dose rate (Sv/a) RS3g 7.0 10 4 2.2 10-7 I-129 2.2 10-7 Pa-231 8.7 10-8 Ra-226 1.7 10-8 TILA-99: DC-pgf scenario 7.1 10 4 3.6 10-5 Pu-239 2.8 10-9 I-129 1.1 10-6 Ra-226 9.3 10-6 Reasons for the differences: The fault is assumed to occur at 30 000 years in TILA-99 and at 70 000 years in this work. This has a moderate effect on Pu-239; The fuel degradation time was 10 000 years in TILA-99 compared with 1 000 000 years in this work. This affects the release rate of Pu-239 from the fuel;

140 In TILA-99, the damaged canister was modelled as a mixing tank of bentonite with volume of 7 000 litres and a through flow of 200 L/a, whereas in this work a flow through the canister interior of 2 L/a is assumed with a volume of 700 litres; Because of the higher solubility limit of thorium in TILA-99, the concentration of Th-230 was about 1 000 times higher in the inner volume of canister; thus also the release rate of the daughter nuclide Ra-226 was higher.

141 9 SUMMARY This report discusses the radionuclide release and transport analysis RNT-2008 forming a part of Posiva's safety case report portfolio. The analysis was conducted for the spent fuel repository of the KBS-3V type that is planned to be constructed in the sparsely fractured bedrock of the Olkiluoto site. RNT-2008 serves as an evolutionary release and transport analysis on the way towards the licensing analyses of the repository. The work deals with the release of radionuclides from spent nuclear fuel and their transport out of the disposal canister, through the surrounding bentonite buffer, and in the fracture network of the host rock up to their release to the biosphere. The main quantitative results are expressed as activity release rates and indicative dose rates. RNT-2008 gathers information from the following other reports of Posiva's portfolio: the site description from Andersson et al. (2007), characteristics of spent fuel from Anttila (2005), canister design from Raiko (2005), repository design from Saanio et al. (2006), processes from Miller & Marcos (2007), and evolution of site and repository from Pastina & Hellä (2006). Additionally, there is a number of other sources more or less directly utilized in the analysis. The groundwater flow analysis and the solute transport analysis by Löfman & Poteri (2008) provided the near-field flow rates from the FEFTRA simulations and far-field transport resistances from the ConnectFlow simulations. The FEFTRA results also delivered the hydraulic gradient, the salinity and the temperature in the tunnel backfill. The rates of congruent release of radionuclides from fuel and the corrosion rates of metallic parts of the fuel assemblies are based on the data used in the TILA-99 safety assessment (Vieno & Nordman, 1999), where the previous release and transport analysis for Posiva's KBS-3V concept was included. The applied instant release fractions of the nuclides in nuclear fuel originate in (SKB, 2006b) and the solubility data in Grivé et al. (2007). The sorption and diffusion coefficients plus porosities for the near field were selected and modified from the central values of the Sr-Can analysis (SKB, 2006), and for the far field from the TILA- 99 data with numerous adjustments. The near-field geometrical model closely resembles the model applied in the REPCOM calculations in TILA-99. The far-field transport model is basically similar to that applied with the FTRANS code in TILA-99. A tree-like structure of calculation cases was composed from the scenarios described in the process report (Miller & Marcos, 2007), and the actually assessed cases were chosen from the complete set of defined cases. The branches of the structure represent different diameters and timings of canister holes, salinities of groundwater, the groundwater flow rates in the near field, and hydrodynamic control of retention in the geosphere. The base case includes a penetrating pinhole failure that would exist initially in a canister with a diameter of 1 mm. It serves as a reference for the other calculation cases. The main scenario of Posiva's safety case was not analysed, since it assumes scientifically reasonable and realistic, extremely slow corrosion of fuel canisters, in which case a radionuclide transport analysis would be meaningless. The number of calculation cases was reduced from the TILA-99 analysis, but they are deemed to also cover consequences of features, events and processes that are not explicitly referred to in the calculation case definitions. Many of the TILA-99 cases and their outcomes can still be used to complement understanding on the repository type in question. In addition to the

142 so called realistic calculation cases, complementary sensitivity cases and hypothetical "what if" cases were analysed as a means to demonstrate the robustness of the repository design and illustrate the relative importance of the components of the multibarrier system. The realistic cases cannot be regarded truly realistic but merely as less conservative ones than the other classes of cases. All the cases assume singlecanister failures, which means that direct conclusions on multiple-canister failures cannot be drawn from this analysis without further efforts. The following release and transport processes were included in the models: Release from the fuel elements o Congruent release from the degrading ceramic fuel matrix o Instant release of the radionuclides gathered in fuel rod gaps and the grain boundaries of fuel pellets o Corrosion of activated metal components Near-field transport o Solubility limits inside the canister and at the buffer-rock interface o Release through an initial or delayed hole in the canister wall o Gas-induced release of contaminated water or gaseous radionuclides from a canister o as-induced release from the canister o Advection, diffusion and sorption (K d ) Far-field transport o Advection in the fracture network of bedrock o Diffusion and sorption in rock matrix The safety indicators used in this analysis are radionuclide release rates and indicative dose rates. The release rates are presented at the following two boundaries: from the near field to the geosphere, and from the geosphere to the biosphere. The indicative dose rates were calculated by multiplying the nuclide-specific release rates at the geosphere-biosphere interface by the WELL-2008 dose conversion factors. The release rates will be fed as an input to a separate biosphere assessment that will deal with more complex exposure pathways within the first several thousands of years after the repository closure and end up in so called landscape doses (see e.g. Broed, 2007). The time scale of the current analysis is one million years enveloping the essential span of all the meaningful long-lived nuclides. RNT-2008 is a deterministic analysis, where the data for each repository calculation case are individually specified, i.e. not sampled from probabilistic distributions. The uncertainties are firstly scoped with wide-range modelling assumptions and secondly with parameter variation. The assumptions are purposely conservative, meaning that they shall ensure that the results, with high degree of certainty, overestimate the radiation exposures or radioactive releases. The conceptualizations within both near and far field were kept at a quite plain level, emphasizing transparency.

143 The results suggest that the assessed repository system well complies with the regulatory criteria. None of the calculated nuclide-specific or total activity release rates exceeds the corresponding regulatory constraints. The presented indicative dose rates with the WELL-2008 dose conversion factors remain below the regulatory limit of the annual effective dose, 0.1 msv, in all the cases except an early release peak of Sr-90 in an entirely hypothetical "what if" case (Lh Q Sal). It is essential to take into account that the dose rates in this report represent deterministic radiation impacts, i.e. the probabilities of the cases are not incorporated. Their incorporation would give true expectation values of dose rates or effective dose rates, which would mean lower numerical values. Considering unlikely events like those in the "what if" cases, Guide YVL 8.4 states that, if the resulting individual dose might imply deterministic radiation impacts of 0.5 Sv or above, the order of magnitude estimate for its annual probability of occurrence shall be 10-6 at the most. Based on the dose rates calculated using the biosphere dose conversion factors of WELL-2008, the most significant radionuclides in most cases are I-129, C-14, Cs-135, Pa-231 and Cl-36, in this order. Besides their inventories and relatively long half-lives, the following aspects contribute to their importance: I-129: high instant release fraction (IRF) in the release from fuel; no sorption accounted; rather high dose conversion factor C-14: rapid release from the large inventory fraction in activated metal parts; no sorption accounted Cs-135: high IRF; fairly weak retention Pa-231: continuously produced as a daughter nuclide of U-235; relatively strong sorption both in the near field and far field; important only with high flow rate and at late times; one of the highest dose conversion factors Cl-36 high IRF; no sorption. Thus, besides having relatively long half-lives, the dominating nuclides typically have high release rates from fuel elements and are modelled either as non-sorbing or weakly sorbing species. The results of the analysis emphasize the importance of disposal canisters and other system parts in the immediate vicinity of canisters. The engineered barriers together with the sparsely fractured host rock around the near field dominate the capacity of the repository to retain the radionuclides and retard their movements. The central role of the bedrock farther from the canisters is to provide stable and favourable chemical and physical conditions for the engineered barrier system, to hinder inadvertent human intrusion into the repository, and to extend the migration pathways of radioactive substances. Complementing the above mentioned, the conclusions on the regulatory compliance regarding the landscape doses will be presented in the biosphere assessment reporting.

144

145 REFERENCES Ahjos T., Uski M., 1992. Earthquakes in northern Europe in 1375-1989. Tectonophysics, 207, 1 23. Ahokas H., Vaittinen T., Tammisto E., Nummela J., 2007. Modelling of Hydro-Zones for Layout Planning and Numerical Flow Model in 2006. Posiva Oy. Working Report 2007-01. Andersson J., Ahokas H., Hudson J., Koskinen L., Luukkonen A., Löfman J., Keto V., Pitkänen P., Mattila J., Ikonen A. & Ylä-Mella M., 2007. Olkiluoto site description 2006. Posiva Oy, Olkiluoto, Finland. Report POSIVA 2007-03. Andersson J., Hermansson J., Elert M. Gylling B., Moreno L. & Selroos J-O., 1998. Derivation and treatment of the flow wetted surface and other geosphere parameters in the transport models FARF31 and COMP23 for use in safety assessment. Swedish Nuclear Fuel and Waste Management Co (SKB), Stockholm, Sweden. SKB R-98-60. Anttila M., 2005. Radioactive Characteristics of the Spent Nuclear Fuel of the Finnish Nuclear Power Plants. Posiva Oy, Olkiluoto, Finland. Posiva Working Report 2005-71. Bradbury M. & Baeyens B., 2003. Near-field sorption databases for compacted MX-80 bentonite for performance assessment of a high-level radioactive waste repository in Opalinus Clay host rock. Nagra, Wettingen, Switzerland. Technical Report 02-18. Broed R., 2007. Landscape model configuration for biosphere analysis of selected cases in TILA-99 and in KBS-3H safety evaluation. Posiva Working report 2007-108. Council of State, 1991. Section 8 in STATE 398/91 (14.2.1991). The Council of State of Finland. <http://www.edilex.fi/stuklex/en/lainsaadanto/19910398> Cramer J.J. & Smellie J.A.T., 1994. Final report of the AECL/SKB Cigar Lake analog study. AECL Technical Report, AECL-10851; SKB Technical Report, TR 94-04. Cvetkovic, V., Selroos, J. O., Cheng, H. 1999. Transport of reactive tracers in rock fractures. Journal of Fluid Mechanics, vol. 378. pp. 335-356. Duro L., Grivé M., Cera E., Gaona X., Domènech C. & Bruno J., 2006. Determination and assessment of the concentration limits to be used in SR-Can. Swedish Nuclear Fuel and Waste Management Co. (SKB), Stockholm, Sweden. SKB TR 06-32. EUR, 2005. Treatment of Radionuclide Transport in Geosphere within Safety Assessments. (Retrock). Final report June 2005. Contract No. FIKW-CT-2001-20201. EUR 21230 EN. Grivé M., Montoya V. & Duro L., 2007. Assessment of the concentration limits for radionuclides for Posiva. Posiva Oy, Olkiluoto, Finland. Posiva Working Report 2007-103.

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147 Liu J., Löfgren M. & Neretnieks I., 2006. SR-Can Data and uncertainty assessment. Matrix diffusivity and porosity in situ. Swedish Nuclear Fuel and Waste Management Co. (SKB), Stockholm, Sweden. SKB Report R-06-111. Löfgren M. & Neretnieks I., 2003. Formation factor logging by electrical methods: Comparison of formation factor logs obtained in situ and in the laboratory. J. Contaminant Hydrology, 61, 107-115. Löfman J., 2005. Simulation of Hydraulic Disturbances Caused by the Decay Heat of the Repository in Olkiluoto. Posiva Oy, Olkiluoto, Finland. Report POSIVA 2005-07. Löfman J., Keto V. & Mészáros F., 2007. FEFTRA Verification. VTT Research Notes 2385. Edita Prima Oy. Helsinki. http://www.vtt.fi/inf/pdf/tiedotteet/2007/t2385.pdf Löfman J. & Poteri A., 2008. Groundwater flow and transport simulations in support of RNT-2008 analysis. Posiva Oy, Eurajoki, Finland. Posiva Working Report 2008-52. Miller B. & Marcos N., 2007. Process report- FEPs and scenarios for a spent fuel repository at Olkiluoto. Posiva Oy, Eurajoki, Finland. POSIVA 2007-12. Moreno L. & Gylling B., 1998. Equivalent flow rate concept in near field transport model COMP23. SKB R-98-53. Svensk Kärnbränslehantering AB. Motta M.M. & Miranda C.F., 1989. Molybdate adsorption on kaolinite, montmorillonite and illite: constant capacitance modelling. Soil Sci. Soc. Am. J., 53, 380-385. Mäntyniemi P. & Ahjos T., 1990. A catalog of Finnish earthquakes in 1610-1990. Geophysica, 26(2), 17-35. Mäntyniemi P., 2005. A Tale of two earthquakes in the Gulf of Bothnia, Northern Europe in 1880s. Geophysica, 41(1-2), 73-91. Neretnieks I., 1980. Diffusion in the rock matrix: an important factor in radionuclide migration? Journal of Geophysical Research, 85, 4379-4397. Neretnieks I., 1982. Leach rates of high level waste and spent fuel Limiting rates as determined by backfill and bedrock conditions. New York, Elsevier Science Publishing company Inc., Scientific Basis for Nuclear Waste Management, 557-568. Neretnieks I., 2002. A stochastic multi-channel model for solute transport analysis of tracer tests in fractured rock. Journal of Contaminant Hydrology 55 (2002) 175 211. Nilsson L., Moreno L., Neretnieks I. & Romero L., 1991. A resistance network model for radionuclide transport into the near field surrounding a repository for nuclear waste (SKB, Near Field Model 91). Swedish Nuclear Fuel and Waste Management Co. (SKB), Stockholm, Sweden. SKB Technical Report TR-91-30.

148 Nordman H. & Vieno T., 1994. Near-field model REPCOM. Nuclear Waste Commission of Finnish Power Companies (YJT), Helsinki, Finland. Report YJT-94-12. Nordman H. & Vieno T., 2003. Modelling of near-field transport in KBS-3V/H type repositories with PORFLOW and REPCOM codes. Posiva Oy, Olkiluoto, Finland. Posiva Working Report 2003-07. Ochs M. & Talerico C., 2004. SR-Can Data and uncertainty assessment Migration parameters for the bentonite buffer in the KBS-3 concept. Swedish Nuclear Fuel and Waste Management Co. (SKB), Stockholm, Sweden. SKB Technical Report TR-04-18. Pastina B. & Hellä P. (Eds.), 2006. Expected Evolution of the Spent Nuclear Fuel Repository at Olkiluoto. Posiva Oy, Olkiluoto, Finland. Report POSIVA 2006-05. Pereira A., 2006. Three dimensional modelling of a KBS-3 canister for spent nuclear fuel - some migration studies. SKI, Stockholm, Sweden. Swedish Nuclear Power Inspectorate (SKI) Report 2006:17. Poinssot C., Ferry C., Grambow B., Kelm M., Spahiu K., Martinez A., Johnson L., Cera E., de Pablo J., Quinones J., Wegen D., Lemmens K., McMenamin T., 2006. Mechanisms governing the release of radionuclides from spent nuclear fuel in geological repository: Major outcomes of the European Project SFS. Scientific Basis for Nuclear Waste Management XXIX, Symposium held in Ghent Belgium September 12 16, 2005, Materials Research Society Symposium Proceedings Volume 932. Posiva, 2006. TKS-2006 Nuclear waste management of the Olkiluoto and Loviisa power plants: Programme for research, development and technical design for 2007 2009. Posiva Oy, Olkiluoto, Finland. Posiva, 2008. Safety Case Plan. Posiva Report 2008-05. Posiva Oy. July 2008. Poteri A., 2007. A concept for radionuclide transport modelling. Oy, Olkiluoto, Finland. Working Report POSIVA 2007-24. Påsse T., 1996. A mathematical model of the Shorelevel displacement in Fennoscandia. Technical Report 96-24. SKB, Stockholm. Raiko H., 2005. Disposal Canister for Spent Nuclear Fuel Design Report. Posiva Oy, Olkiluoto, Finland. Report POSIVA 2005-02. Rodwell W.R., 2005. Summary of a GAMBIT Club workshop on gas migration in bentonite, Madrid 29 30 October, 2003. Swedish Nuclear Fuel and Waste Management Co (SKB); Stockholm, Sweden. SKB Technical Report TR-05-13. Ruokola E., 2002. Consideration of timescales in the Finnish safety regulations for spent fuel disposal. Proc. of Workshop on the handling of timescales in assessing postclosure safety of deep geological repositories on 16 18 April 2002 in Paris, France. Organisation for Economic Co-operation and Development, Nuclear Energy Agency.

149 Saanio T., Kirkkomäki T., Keto P., Kukkola T. & Raiko H., 2006. Preliminary design of the Repository. Stage 2. Posiva Oy, Olkiluoto, Finland. Posiva Working Report 2006-94. Saanio T. & Raiko H., 1999. Retrievability of spent fuel canisters (in Finnish). Posiva Working Report 99-21. Serco Assurance, 2005. CONNECTFLOW Release 9.0 Technical Summary Document, Serco Assurance Report SA/ENV/CONNECTFLOW/15. SKB, 1999. SR 97 - Processes in the repository evolution. Swedish Nuclear Fuel and Waste Management Co. (SKB), Stockholm, Sweden. SKB Technical Report TR-99-07. SKB, 2006. Long-term safety for KBS-3 repositories at Forsmark and Laxemar a first evaluation. Main report of the SR-Can project. Swedish Nuclear Fuel and Waste Management Co (SKB); Stockholm, Sweden. SKB Technical Report TR-06-09. SKB, 2006a. Fuel and canister process report for the safety assessment SR-Can. Swedish Nuclear Fuel and Waste Management Co (SKB); Stockholm, Sweden. SKB Technical Report TR-06-22. SKB, 2006b. Data report for the safety assessment SR-Can. Swedish Nuclear Fuel and Waste Management Co (SKB); Stockholm, Sweden. SKB Technical Report TR-06-25. SKBF/KBS, 1983. Kärnbränslecykelns slutsteg. Använt kärnbränsle KBS-3. Del I IV. Svensk Kärnbränsleförsörjning AB. Smart N.R., Fennell P.A.H., Rance A.P. & Werme L., 2004. Galvanic corrosion of copper-cast iron couples in relation to the Swedish radioactive waste canister concept. In Prediction of Long term Corrosion Behaviour in Nuclear Waste Systems, Proceedings of the 2nd International Workshop, Nice September 2004, Eurocorr 2004, edited by ANDRA, France, 52-60. Smith P., Nordman H., Pastina B., Snellman M., Hjerpe T. & Johnson L., 2007. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto - Radionuclide transport report. POSIVA 2007-07 and SKB R-08-38. Posiva Oy, Olkiluoto, Finland and Swedish Nuclear Fuel and Waste Management Co (SKB), Stockholm, Sweden. (in press) STUK, 1999. Government decision on the safety of disposal of spent nuclear fuel (478/1999). Radiation and Nuclear Safety Authority (STUK) Report STUK-B-YTO 195. STUK, Helsinki, Finland. STUK, 2001. Long-term safety of disposal of spent nuclear fuel. Radiation and Nuclear Safety Authority (STUK). Guide YVL 8.4. Tanskanen J. & Palmu M., 2004. Facility Description 2003. Posiva Oy, Eurajoki, Finland. Posiva Working Report 2004-26.

150 Vieno T., Hautojärvi A., Koskinen L. & Nordman H., 1992. TVO-92 safety analysis of spent fuel disposal. Helsinki, Nuclear Waste Commission of Finnish Power Companies, Report YJT-92-33E. Vieno T. & Ikonen A.T.K., 2005. Plan for Safety Case of spent fuel repository at Olkiluoto. POSIVA 2005-01. Posiva Oy, Olkiluoto, Finland. Vieno T. & Nordman H., 1996. Interim report on safety assessment of spent fuel disposal TILA-96. Posiva Oy, Helsinki, Finland. Report POSIVA 96-17. Vieno T. & Nordman H., 1999. Safety assessment of spent fuel disposal in Hästholmen, Kivetty, Olkiluoto and Romuvaara, TILA-99. Posiva Oy, Helsinki, Finland. Report POSIVA 99-07. Vieno T. & Nordman H., 2000. Updated compartment model for near-field transport in a KBS-3 type repository. Posiva Oy, Helsinki, Finland. Posiva Working Report 2000-41. Werme L.O., Johnson L.H., Oversby V.M., King F., Spahiu K., Grambow B. & Shoesmith D.W., 2004. Spent fuel performance under repository conditions: A model for use in SR-Can. Swedish Nuclear Fuel and Waste Management Co. (SKB), Stockholm, Sweden. SKB Technical Report TR-04-19. Yu J.-W. & Neretnieks I., 1997. Diffusion and sorption properties of radionuclides in compacted bentonite. Swedish Nuclear Fuel and Waste Management Co (SKB) Stockholm, Sweden. SKB Technical Report TR-97-12.

151 Appendix 1: ILLUSTRATION OF GEOSPHERE RETENTION The geosphere retention depends on many parameters, and to get an overview and understanding on each of these parameters it is useful to look at a simplified case which can be described analytically. The main processes in the geosphere transport are advection and diffusion from the flowing water into the pore space of the rock matrix. In the geosphere transport the delay consists of two components: groundwater travel time and the delay time distribution due to the diffusional processes between the flowing water and rock matrix. In the time perspective of the long-term safety analysis the delay due to the groundwater travel time is insignificant. In the following mathematical formulation this is omitted, but it can be simply added as a time shift into the results. The value 25 years was used in the indicative calculations, the results of which are shown in this appendix. As mentioned in Chapter 5.3 (A1-1) where u [T 1/2 ] is a parameter describing the transport properties of the migration route for the given species. (A1-2) p is the porosity of the rock matrix [ ], D e is the effective diffusion coefficient from the fracture into the rock matrix [L 2 T -1 ], R p is the retardation factor of the species in the rock matrix [ ], W is the width of the flow channel i.e. the width, over which the flow is measured [L], L is the transport distance [L], Q is the flow rate in the channel or over the given width [L 3 T -1 ], t is the time [T]. The first factor of the u parameter is related to diffusion and sorption in the rock matrix. The retardation factor, R p, depends on the rock properties and the volume-based distribution coefficient, K d (m 3 /kg) in the following way (A1-3) where s is the solid density of the rock (2700 kg/m 3 ). For non-sorbing species (i.e. K d = 0 m 3 /kg) R p = 1. For moderately or strongly sorbing species R p K d s / p even for quite low values of K d (e.g. K d 10-5 m 3 /kg when p 0.01). The u parameter is thus reduced to:

152 (A1-4) (A1-5) The second factor (WL/Q) of the u parameter can be expressed also in terms of the groundwater transit time t w and the volume aperture of the flow channel 2b v : (A1-6) where v is the advection velocity of the groundwater in the channel (m/a). The flowrelated transport parameter is represented in the literature also by the so-called F factor, F = 2 WL/Q (Chapter 4.2.3). The derivation of WL/Q for the radionuclide release and transport calculations in the present study is described at length in Chapter 4.2. The effect of the u parameter on retention is illustrated for a set of six K d values for a selected set of other transport parameters given in Table A1-1. Because the lumped parameter u alone determines the retention behaviour, any set of parameter values resulting in a specific value of the u parameter value gives the same retention characteristics. Some alternative sets are given in the third column in Table A1-1. The porosity contributes significantly to the u parameter value only in the first alternative example. In other cases, the u parameter is solely determined by sorption and flowrelated parameters together with the effective diffusion coefficient. Table A1-1. Lumped parameter u values with p = 0.001, D e = 10-14 m 2 /s, and WL/Q = 50000 a/m with six K d values from 10-5 to 1 m 3 /kg. Alternative set of rock matrix properties is shown in the third column; WL/Q is assumed again to 50 000 a/m. u (a 1/2 ) K d (m 3 /kg) Alternative set of rock matrix parameters to obtain same u parameter value. WL/Q still 50 000 a/m. 4.70 10-5 p* = 0.005, D e = 5 10-14 m 2 /s, K d = 2.3 10-7 m 3 /kg 14.6 10-4 p = 0.0001, D e = 10-15 m 2 /s, K d = 10-3 m 3 /kg 46.1 10-3 p = 0.005, D e = 5 10-14 m 2 /s, K d = 2.0 10-4 m 3 /kg 146 10-2 p = 0.0001, D e = 10-15 m 2 /s, K d = 10-1 m 3 /kg 461 10-1 p = 0.005, D e = 5 10-14 m 2 /s, K d = 2.0 10-2 m 3 /kg 1460 10 0 p = 0.0001, D e = 10-15 m 2 /s, K d = 10 1 m 3 /kg * Note that porosity p alone has minor importance in obtaining u (a 1/2 ) and that any suitable combination of parameters gives the same value. The release pulses from the geosphere with u parameter values from Table A1-1 and groundwater transit time of 25 years are shown in Figure A1-1 in log-log scale. A closeup of the five first release pulses up to 100 000 years is presented in Figure A1-2 in a linear scale.

153 For very long-lived nuclides the release rates will ultimately approach the inflow rates. It can be concluded that without radioactive decay everything that gets into the geosphere, would be released into the biosphere. The timing of releases depends on the values of retention parameters. Figure A1-1. Release rates from geosphere with different u parameter values and a constant inflow rate of 1 unit per year (e.g. mol/a for stable species or Bq/a for very long-lived nuclides). Groundwater transit time of 25 years has been added into the results. The unit of u is square root of year. Figure A1-2. Release rates from geosphere with different u parameter values and a constant inflow rate of 1 unit per year (e.g. mol/a for stable species or Bq/a for very long-lived nuclides). Groundwater transit time of 25 years has been added into the results. The unit of u is a 1/2.

154

155 Appendix 2: DESCRIPTION OF CODES A.2.1 The REPCOM code Near-field analyses have been performed with the REPCOM code. REPCOM has been developed by the Technical Research Centre of Finland (VTT) for radionuclide transport analyses in the near field of repositories for low and intermediate level waste or spent fuel. The phenomena that can be modelled using REPCOM are: release from the waste - several waste types, each with different release functions, can be included; advective and / or diffusive transport within a system of engineered barriers; sorption on solid surfaces; solubility limitation of concentrations; and radionuclide decay and in-growth. Detailed descriptions of how these phenomena are treated, including the governing equations solved by REPCOM, are given in Nordman & Vieno (1994). In order to simulate the migration of radionuclides through a system of engineered barriers and their release to the geosphere, the repository is discretised into small volumes termed compartments. In setting up a problem to be solved using REPCOM, compartment sizes must be chosen that are sufficiently small that instantaneous mixing can be assumed in each compartment 19. REPCOM was developed to a two-dimensional model before this analysis. Stepwise changes in the properties and geometry of the compartments can be accommodated, and different boundary conditions at the interface with the geosphere can be applied by means of user-specified mass transfer coefficients 20. The derivation of the transfer coefficients is presented in detail in Chapter 11.6 of TILA-99 (Vieno & Nordman, 1999), and discussed further in Vieno & Nordman (2000), where the transfer coefficients used in TILA-99 are compared with those used in the Swedish SR 97 assessment (SKB, 1999). REPCOM has been verified against analytical models (Nordman, 1986; Vieno et al., 1992; Nordman & Vieno, 1994). It has also been verified against the commercial PORFLOW code, a computational fluid dynamics tool used by nuclear waste management and research organisations in several countries. This exercise included preliminary near-field transport analyses for KBS-3H, as well as KBS-3V (Nordman & Vieno, 2003). 19 A concept for the discretisation of the near field of KBS-3V type repository for use with REPCOM is described in Vieno & Nordman (2000). This represents an update of the discretisation used in TILA-99 and earlier assessments. 20 The applicability of the transfer coefficient approach has been assessed using transient threedimensional modelling of the repository near field in Pereira (2006).

156 A.2.2 The FTRANS code Geosphere analyses have been performed with the FTRANS code (FTRANS, 1983; Nordman & Vieno, 1994). FTRANS is a dual-porosity model for flow and transport. In the flowing porosity domain, conceptualized here as a single fracture, phenomena that can be modelled with FTRANS are: groundwater flow; advective radionuclide transport; and longitudinal dispersion. In the matrix porosity domain, phenomena that can be modelled are: diffusion; and sorption on solid surfaces Radioactive decay and in-growth are represented in both domains, and transfer of radionuclides across the boundary between the domains takes place by diffusion. Flow porosity in the analyses carried out for KBS-3V takes the form of a representative planar fracture of width W [m], length L [m] and flow rate Q [m 3 a -1 ]. FTRANS input parameters are chosen in such a way as to give the required value for the lumped parameter WL/Q, which, as discussed for example in Chapter 11.5 of TILA-99 (Vieno & Nordman, 1999), represents the transport resistance of the geosphere. WL/Q = groundwater travel time / fracture aperture; thus by choosing groundwater velocity and fracture aperture for the FTRANS code the desired WL/Q is obtained (see Eq. A1-6). Matrix porosity in the rock adjacent to the fracture can be subdivided into different subdomains, each with different transport properties. This can be used, for example, to differentiate between mineralogically altered rock immediately adjacent to the fracture, and more distant, unaltered rock. FTRANS has been verified using test cases as part of the INTRACOIN and INTRAVAL projects (INTRACOIN, 1984; Rasilainen, 1989). Additional verification tests against an analytical model have been presented in TVO-92 (Vieno et al., 1992). A2.3 Auxiliary computer codes Besides the main codes described above, the auxiliary codes Apu1a and Apu2a have been used. These are, respectively, pre- and post-processing routines which are used, among other things, to convert REPCOM output into FTRANS input format and to convert the results from FTRANS into a more user-friendly format.

157 Pre-processor Apu1a FTRANS uses four main input files: Files 24, 35, 20 and 17. Apu1a reads these files as follows. Data from File 24: The first line gives the number of materials (typically two different layers of rock matrix) and number of nuclides in the file. The second line gives the porosity and D e values for the two layers of rock used, one layer (e.g. 1 cm thick in file 35) represents the rock matrix close to the fracture surface and the second (e.g. 9 cm thick in file 35) represents the rock matrix further away from the surface. The entire rock matrix thickness would thus be 10 cm. The first four values are for neutral and cationic species and next four values for anions. The next lines give the names and K d values of nuclides. Negative K d values indicated that the nuclide is an anion but the absolute value is used in calculations. With Apu1a, it is possible to assign different porosities and D e values (in the rock matrix) for chain elements thereby adding flexibility to the model. The following lines give the names and half-lives of nuclides. Example of file 24. Additional data, which are not used, are in the file because of historical reasons. These data are shaded in the example below. 2 2 5.00E-03 1.00E-13 1.00E-03 1.00E-14 1.00E-03 1.00E-14 2.00E-04 1.00E-15 C-14 0.00E-04 2.00E-09 0.00E-01 2.00E-10 5.00E-02 1.00E-12 1.00E-01 2.00E-14 2.50E-02 2.00E-12 Cl-36-1.00E-99 2.00E-09 0.00E-01 2.00E-10 5.00E-02 1.00E-12 1.00E-01 2.00E-14 2.50E-02 2.00E-12 C-14 5.70E+03 2.90E-15 1.00E+33 1.28E+10 1.44e+09 1.44e+08 1.09E+10 7.84e+09 0.00E+00 Cl-36 3.00E+05 4.70E-15 1.00E+33 0.00E+00 0.00E+05 0.00E+00 6.72e+08 0.00e+00 0.00e+00 Data from File 35 The first line gives the velocity of water (e.g. 24 m/year), the half aperture of the fracture (e.g. 2.5 10-4 m), the length of the migration route (e.g. 600 m), the number of nodes in the migration direction (e.g. 20) and the increase factor of the distance between nodes in the flow direction (e.g. 1.02 for a 2 % increase) The second line gives the number of rock material layers (2 layers in our case), the thickness of the first layer of rock matrix (e.g. 0.01 m near the fracture)), the number of nodes of the first layer (e.g. 7) and the factor by which the distance between nodes is increased (e.g. 1.5 for a 50 % increase between successive nodes). Next on the second line is the thickness of the second layer (e.g. 0.09 m), the number of nodes in the second layer (e.g. 12) and the factor by which the

158 distance between successive nodes is increased. The three last variables on this second line are read if a two-layer system is used, as is the case in this report. The next two lines do not contain variables and provide guidance for the mathematical solution to be used in calculating the results. Therefore it is important that these lines remain unchanged if the results are to be reproduced. Further information on the lines that do not contain any variable parameter values, but rather information on the mathematical models is provided in the FTRANS manual (FTRANS, 1983). In the fifth line, the third variable gives the number of time steps that may be changed (e.g. 120). The rest of the line must remain unchanged for the same reasons discussed above. On the sixth line, there are five variables, four of which may be changed by the user: the length of first time step (e.g. 0.1), start time of calculation (always zero), increase factor of time step (e.g. 1.2), the maximum length of a time step (e.g. 20 000 years) and the end time of calculation (e.g. 106 years). The next lines in the file must remain unchanged for the same reasons discussed above. Example of File 35 (the contents of this file are described in the FTRANS manual) 24. 2.5e-4 600. 20 1.02 2 0.01 7 1.5 0.09 12 1.5 1 FTRANS TEST: LASKENTATAPAUS HENKKA (korjaus 1) VARIAATIO 1 861 800 120 1 2 8 1 6 0 40 1 0 1.0E-1 0.0E+4 1.2E+0 2.0E+4 1.0E+6 0.0 0-1 10 1 1 1 0 0 0 0 0 0 0 0.0E00 0.0E00 0.00E00 3.155E-9 5.0E-9 3.0E-0 0.0E+0 2.0E+4 1.00e+00 1.000e+00 1.000e+00 0.0E+2 0.0E00 1.0E00 2.5E-4 0.00E+4 0.0E+0 1.00e+00 1.000e+00 1.000e+00 0 2.888E-5 1.000E+0 1.000E+0 1.000E-7 1.0E+5 Data from Files 17 and 20 File 20 and file 17 are the output files from REPCOM. They include the release pulses from the near field. Files 17 and 20 must be in the same directory in which the calculation is carried out. The nuclide names must be the same as in file 24. In file 20, the start time is zero, but in file 17 the real start time, i.e. the time of the event triggering the release is specified (e.g. 1000 years, the time for a defective canister to become penetrated).

159 Example of File 17 Output file for C-14 in the penetrating defect-base case (Sh1) C-14 299 EM= 6.7675E+05 T= 1.01E+04 A= 1.96E-09 2.21E+09 1.000E+03 9.779E-10 1.000E+03 3.818E-06[ ] 2.000E+03 6.096E+02[ ] 3.087E+03 1.227E+03[ ] 5.521E+03 2.109E+03[ ] 1.000E+04 2.687E+03[ ] 5.469E+04 4.397E+02[ ] 1.047E+05 2.939E-01[ ] 2.047E+05 1.534E-07[ ] 3.047E+05 1.998E-13[ ] 4.047E+05 7.834E-19[ ] 5.047E+05 3.985E-24[ ] 6.047E+05 2.081E-29[ ] 7.047E+05 1.089E-34[ ] 8.047E+05 5.697E-40[ ] 9.047E+05 2.982E-45[ ] 1.001E+06 2.444E-50 Example of File 20 Carbon-14 data for Sh1 case In the first line, the number 1 means that nuclide number one in a chain is considered. The number 299 indicates that there are 299 time steps in the release pulse. Following lines: time (years from emplacement) and release (Bq) 1 299 C-14 0.000E+00 9.779E-10 2.749E-01 3.818E-06 4.804E-01 2.690E-04 7.218E-01 3.478E-03 1.005E+00 1.936E-02 1.730E+01 6.840E+00[ ] 1.250E+02 4.470E+01[ ] 1.000E+03 6.096E+02[ ] 1.000E+04 2.745E+05[ ] 5.369E+04 4.397E+02[ ] 1.037E+05 2.939E-01[ ] 5.037E+05 3.985E-24[ ] 9.037E+05 2.982E-45[ ] 1.000E+06 2.444E-50

160 From files 24, 35, 17 and 20 data, Apu1a produces the input file for FTRANS (File 25, see example below, the contents of which is described in the FTRANS manual, FTRANS 1983) and the release pulse from the near field to File 11. FTRANS then is run using File 25. The results from FTRANS are stored into File 10 (the output pulse). Example of File 25 FTRANS input for C-14 in Sh1 (penetrating defect-base case) 1 FTRANS TEST: LASKENTATAPAUS HENKKA (korjaus 1) VARIAATIO 1 420 380 120 1 2 8 1 6 0 20 1 0 1.0E-1 0.0E+4 1.2E+0 2.0E+4 1.0E+6 0.0 0-1 10 1 1 1 0 0 0 0 0 0 0 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 0.000E+00 0.000E+00 0.000E+00 1.000E+00 1.000E+00 3.000E+00 0.000E+00 2.000E+04 1.000E+00 3.156E-06 5.000E-03 0.000E+00 0.000E+00 0.000E+00 1.000E+00 1.000E+00 3.000E+00 0.000E+00 2.000E+04 1.000E+00 3.156E-07 1.000E-03 0.000E+00 0.000E+00 1.000E+00 2.500E-04 0.000E+00 0.000E+00 1.000E+00 0 1.216E-04 1.000E+00 1.000E+00 1.000E-07 1.000E+05 20 21 600.000 600.000 1 0.000 24.694 49.882 75.574 101.779 128.509 155.773 183.582 211.948 240.881 270.393 300.495 331.199 362.517 394.461 427.044 460.279 494.179 528.756 564.025 600.000 0.000E+00 3.108E-04 7.771E-04 1.476E-03 2.525E-03 4.099E-03 6.459E-03 1.000E-02 1.035E-02 1.087E-02 1.166E-02 1.284E-02 1.461E-02 1.726E-02 2.124E-02 2.722E-02 3.617E-02 4.961E-02 6.977E-02 1.000E-01 1 1 21 1 2.500E-04 2 21 41 1 2.500E-04 3 41 61 1 2.500E-04 4 61 81 1 2.500E-04 5 81 101 1 2.500E-04 6 101 121 1 2.500E-04 7 121 141 1 2.500E-04 8 141 161 1 2.500E-04

161 9 161 181 1 2.500E-04 10 181 201 1 2.500E-04 11 201 221 1 2.500E-04 12 221 241 1 2.500E-04 13 241 261 1 2.500E-04 14 261 281 1 2.500E-04 15 281 301 1 2.500E-04 16 301 321 1 2.500E-04 17 321 341 1 2.500E-04 18 341 361 1 2.500E-04 19 361 381 1 2.500E-04 20 381 401 1 2.500E-04 0 1 1 1 1 1 1.000E+00 6.000E-03 0.000 0.000 0.000 0.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 24.000 1 0 401 0 0.600E-02

162

163 Appendix 3: LIST OF CALCULATION CASES Summary of assessment scenarios and corresponding calculation cases. Type of case: RC (realistic case 21 ), SC (sensitivity case), WIC (what if case). Identifier Time of defect (a) Diameter of canister defect (mm) Filling of defect W=water B=bentonite Groundwater flow rate Q Groundwater chemistry D = dilute B = brackish Sal = Saline DEFECTIVE CANISTER SCENARIO DCS-II Fuel degradation rate (1/a) Type of case Sh1 0 1 W default D/B 10-7 RC Sh1-EPR 0 1 W default D/B 10-7 RC Sh1- VVER 0 1 W default D/B 10-7 RC Sh1 Fd 0 1 W default D/B 10-6 SC Sh1 Irf 0 1 W default D/B 10-7 SC Sh1 Q 0 1 W high D/B 10-7 SC Sh1 Sal 0 1 W default Sal 10-7 SC Sh1 Q Sal 0 1 W high Sal 10-7 WIC Sh4 0 4 W default D/B 10-7 SC Sh4 Q 0 4 W high D/B 10-7 SC Sh4 Q Sal 0 4 W high Sal 10-7 WIC Lh Q 0 100 W high D/B 10-7 WIC Lh Q Irf 0 100 W high D/B 10-7 WIC Lh Q Sal 0 100 W high Sal 10-7 WIC LhB Q 0 100 B high D/B 10-7 WIC LhB Q Fd 0 100 B high D/B 10-6 WIC DEFECTIVE CANISTER SCENARIO DCS-I LhB Q t4 10 4 100 mm B high D/B 10-7 WIC LhB Q t5 10 5 100 mm B high D/B 10-7 WIC 21 Less conservative cases than in the other categories of cases.

164 Identifier ROCK SHEAR/EARTHQUAKE SCENARIO (AD-I) Time of rupture (a) Event Groundwater chemistry Fuel degradation rate (1/a) Type of case RS1 10 3 RS D/B 10-7 WIC RS2 10 4 RS D/B 10-7 WIC RS3 7 10 4 EQ D/B 10-7 RC RS3g 7 10 4 EQ glacial 10-6 SC Identifier Time span for defect enlargement (a) DISRUPTED BUFFER SCENARIO (AD-II) Flow Groundwater chemistry Fuel degradation rate (1/a) Type of case B Sh-Lh Q 0-10 5-10 6 high D/B 10-7 WIC B Sh- Lh Q g 0-10 5-10 6 high glacial 10-7 WIC Identifier Time span for defect enlargement (a) GAS SCENARIO (AD-III) Flow Groundwater chemistry Remarks Type of case GASexW 2800-4100 default D/B Only IRF WIC GASexG 900 default D/B Only C-14 in IRF WIC

1 (1) LIST OF REPORTS POSIVA-REPORTS 2008 POSIVA 2008-01 KBS-3H Design Description 2007 Jorma Autio, Erik Johansson, Annika Hagros, Saanio & Riekkola Oy Lennart Börgesson, Torbjörn Sandén, Clay Technology AB Pekka Anttila, Paul-Erik Rönnqvist, Fortum Nuclear Services Ltd Magnus Eriksson, Bo Halvarsson, Vattenfall AB Jarno Berghäll, Raimo Kotola, Ilpo Parkkinen Finnmap Oy ISBN 978-951-652-160-5 POSIVA 2008-02 Microbiology of Olkiluoto Groundwater, 2004 2006 Karsten Pedersen, Microbial Analytics Sweden AB February 2008 ISBN 978-951-652-161-2 POSIVA 2008-03 Horizontal deposition of canisters for spent nuclear fuel - Summary of the KBS-3H Project 2004 2007 ISBN 978-951-652-163-6 POSIVA 2008-04 Seismicity in the Olkiluoto Area Jouni Saari ÅF-Enprima March 2008 ISBN 978-951-652-164-3 POSIVA 2008-05 Safety Case Plan 2008 Posiva Oy ISBN 978-951-652-165-0 July 2008 POSIVA 2008-06 Radionuclide Release and Transport - RNT-2008 Mikko Nykyri, Safram Oy Henrik Nordman, VTT Nuria Marcos, Saanio & Riekkola Oy Jari Löfman, VTT Antti Poteri, VTT Aimo Hautojärvi, Posiva Oy ISBN 978-951-652-166-3 December 2008