Waste Management of the Encapsulation Plant

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POSIVA OY

Olki luoto

FI-27160 EURAJOKI, F INLAND

Phone (02) 8372 31 (nat. ) , (+358-2-) 8372 31 ( int. )

Fax (02) 8372 3809 (nat. ) , (+358-2-) 8372 3809 ( int. )

January 2016

Working Report 2015-51

Maiju Paunonen, Ol l i Nummi, Tapani Eurajoki

January 2016

Working Reports contain information on work in progress

or pending completion.

Maiju Paunonen, Ol l i Nummi, Tapani Eurajoki

Fortum Power and Heat Oy

Working Report 2015-51

Waste Management of the Encapsulation Plant

ABSTRACT

During normal operation of the encapsulation plant about 4 m3 of radioactive liquid waste is estimated to be produced annually. Liquid waste will be dried in 200 liter drums leading to the amount of 12 drums of dried waste (less than 3 m3) during the op-erational life of about 100 years. Solid encapsulation waste will be sorted and packed into 200 liter drums and compressible waste will be compacted in the drums. Large components and metal items will be packed into metal boxes. It is estimated that about 10 solid waste drums and less than one metal box containing encapsulation waste will be produced annually during the operation of the encapsulation plant. The total amount of packed waste to be disposed of is about 520 m3 during the operating time of the en-capsulation plant and 120 m3 during the decommissioning of the plant. The waste produced in the encapsulation plant will be disposed of into the low and in-termediate level waste (LILW) repository located along the access tunnel of the spent fuel repository, at the depth of -180 meters. The drums containing dried waste will be placed in a special concrete basin, where the space between the drums will be filled with concrete. Waste packages will be transported to the repository mainly using the canister lift, but some larger components may have to be transported through the access tunnel. Transports from and to the lift will be performed by a forklift. Since the surface dose rate of the drum containing dried waste may be high (97 mSv/h), a radiation shield must be used during the transports or/and the handling must be performed remote-controlled. This can be done with the bridge cranes in the encapsulation plant's waste handling fa-cilities and in the repository. The condition of the waste packages needs to be monitored during the operating phase of the repository. Especially if the repository will be open during the whole operation time of the encapsulation plant, the humidity should be kept favorable for the waste packages to prevent damages of the waste packages prior to the closure of the repository. The waste to be disposed of in the LILW repository is the waste generated from the en-capsulation of spent fuel from LO1-2 and OL1-3 and from the decommissioning of the encapsulation plant. The activity inventory of the waste to be disposed of in the LILW repository is conservatively estimated to be 5.3·1010 Bq in 2120, which is estimated closing time of the LILW repository. Keywords: encapsulation plant, waste handling, activity inventory.

KAPSELOINTILAITOKSEN JÄTEHUOLTO

TIIVISTELMÄ

Kapselointilaitoksen normaalikäytön aikana nestemäisiä jätteitä arvioidaan kertyvän noin 4 m3 vuodessa. Nestemäiset jätteet kuivataan suoraan 200 litran tynnyriin, joita kertyy noin 12 kappaletta (alle 3 m3) koko kapselointilaitoksen käytön aikana (noin 100 vuotta). Kiinteät huoltojätteet lajitellaan ja pakataan 200 litran tynnyreihin ja kokoonpu-ristuvat jätteet puristetaan kasaan jätepuristimella. Isommat komponentit ja metallijät-teet pakataan metallilaatikoihin. Kiinteän jätteen tynnyreitä arvioidaan kertyvän noin 10 kappaletta ja metallilaatikoita alle yksi kappale vuodessa. Yhteensä loppusijoitettavaa pakattua jätettä kertyy noin 520 m3 kapselointilaitoksen käytön aikana ja noin 120 m3

käytöstäpoiston aikana. Kapselointilaitoksella syntyvät jätteet loppusijoitetaan matala- ja keskiaktiivisen jätteen loppusijoitustilaan, joka sijaitsee noin -180 metrin syvyydessä käytetyn polttoaineen loppusijoitustilaan johtavan ajotunnelin varrella. Kuivattu jäte sijoitetaan erilliseen betonikaukaloon, jossa tynnyreiden välit täytetään betonilla. Jätepakkaukset siirretään loppusijoitustasolle pääasiassa kapselihissillä, mutta isompia komponentteja voidaan joutua kuljettamaan ajotunnelia pitkin. Jätepakkaukset siirretään hissiin ja hissistä pois trukilla. Kuivatun jätteen tynnyrin pinta-annosnopeus voi nousta korkeaksi (97 mSv/h), minkä vuoksi tynnyreitä kuljetetaan säteilysuojan sisällä ja/tai kauko-ohjatusti. Siirrot voidaan toteuttaa kapselointilaitoksen jätteenkäsittelytilassa ja loppusijoitustilassa olevilla siltanostureilla. Loppusijoitustilassa jätepakkausten kuntoa on seurattava tilan käytön aikana. Erityisesti jos loppusijoitustila pidetään auki koko kapselointilaitoksen käytön ajan, on tilan kosteus syytä pitää suotuisana jätepakkauksille. Muutoin jätepakkausten kunto voi heiketä oleellisesti jo ennen loppusijoitustilan sulkemista. Matala- ja keskiaktiivisen jätteen loppusijoitustilaan loppusijoitetaan LO1-2 ja OL1-3 käytetyn ydinpolttoaineen kapseloinnista syntyvät jätteet sekä kapselointilaitoksen käytöstäpoistojätteet. Loppusijoitettavan jätteen kokonaisinventaari loppusijoitustilan sulkuhetkellä noin vuonna 2120 on konservatiivisen arvion mukaan 5,3·1010 Bq. Avainsanat: kapselointilaitos, jätteenkäsittely, aktiivisuusinventaari.

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TABLE OF CONTENTS

ABSTRACT TIIVISTELMÄ 1  INTRODUCTION .................................................................................................... 3 2  AMOUNTS OF WASTE .......................................................................................... 5 

2.1  Initial data ...................................................................................................... 5 2.2  Liquid waste .................................................................................................. 5 2.3  Solid waste .................................................................................................... 7 2.4  Decommissioning waste ............................................................................... 9 2.5  Materials ending up in the LILW repository ................................................. 10 

3  WASTE HANDLING SYSTEMS FOR LIQUID WASTE ........................................ 11 3.1  Collection and handling system of drainage water of the controlled

area ............................................................................................................. 11 3.2  Drying system of radioactive drainage water .............................................. 13 

3.2.1  Provision for using drying systems in OL3 unit ............................... 14 3.3  Handling of other liquid waste ..................................................................... 15 

4  HANDLING OF SOLID WASTE ............................................................................ 17 4.1  Sorting of waste .......................................................................................... 17 4.2  Compaction of compressible waste ............................................................ 17 4.3  Handling of metal waste and large components ......................................... 18 4.4  Activity measurement and records management of packaged waste ......... 19 

4.4.1  Activity measurement of drums ....................................................... 19 4.4.2  Activity measurement of metal boxes and large components ......... 20 4.4.3  Records management of waste ...................................................... 20 

5  FINAL REPOSITORY FOR ENCAPSULATION WASTE ..................................... 23 5.1  Waste handling systems in the repository ................................................... 23 5.2  Challenges related to the long operational lifetime of the repository .......... 23 

6  LIMITING THE AMOUNTS OF WASTE ............................................................... 27 7  CLEARANCE OF WASTE .................................................................................... 29 

7.1  Regulations regarding clearance ................................................................ 29 7.2  Clearance procedures at the encapsulation plant ....................................... 29 

7.2.1  Clearance of drums containing solid waste..................................... 29 7.2.2  Clearance of metal waste ............................................................... 29 7.2.3  Criteria for water released from encapsulation plant....................... 30 

7.3  Other nuclear use items .............................................................................. 30 8  LOGISTICS ........................................................................................................... 33 9  ACTIVITY INVENTORY OF ENCAPSULATION WASTE ..................................... 35 

9.1  Release of fission products from the spent fuel .......................................... 35 9.1.1  Fuel defects in the encapsulation process ...................................... 35 9.1.2  Frequency of fuel defects ................................................................ 37 9.1.3  Fuel defects at the reactor units ...................................................... 39 9.1.4  Release fractions of fission products .............................................. 40 9.1.5  Activity inventory of spent fuel......................................................... 41 9.1.6  Total activity of fission products at the LILW hall closure ................ 43 

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9.2  CRUD .......................................................................................................... 45 9.2.1  Handling of crud at the encapsulation plant .................................... 45 9.2.2  Activity inventory in crud ................................................................. 46 9.2.3  Amount and elemental composition of crud .................................... 46 9.2.4  Activation rate ................................................................................. 48 9.2.5  Residence time in reactor ............................................................... 49 9.2.6  Activity concentration in crud immediately after activation .............. 50 9.2.7  Activity inventory of crud ................................................................. 50 

9.3  Summary of activity inventory in encapsulation waste ................................ 52 9.4  Activity inventory in the dried liquid waste drums ........................................ 56 9.5  Maximum activity concentration of drainage water ..................................... 58 9.6  Uncertainties in the activity inventory .......................................................... 59 

10  SUMMARY ........................................................................................................... 61 REFERENCES ............................................................................................................. 63 

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1 INTRODUCTION

This report presents Posiva's waste management plans for the low and intermediate lev-el (LILW) operational and decommissioning waste from the spent fuel encapsulation plant. The waste to be disposed of in the LILW repository is the waste generated from the encapsulation of spent fuel from LO1-2 and OL1-3 and from the decommissioning of the encapsulation plant. The report covers the waste management from sorting of the waste to the final disposal of the waste packages. The plans are based on an assumption that the encapsulation plant is operating as an independent facility. Until the end of the operation of the Olkiluoto nuclear power plant, it may be possible to utilize the waste management facilities of the existing power plant units. Still, after the final shutdown of the power plant, the encapsulation plant has to operate as an independent facility, and therefore at least a preliminary design and corresponding layout provisions should be made already at the design stage of the encapsulation plant. In addition to the waste management systems, this report discusses possibilities to limit the accumulation of waste as well as clearance of radioactive waste. Operational experi-ence from the waste management and final disposal at the Olkiluoto and Loviisa power plants has served as an important input to the proposed design. Previously, the processes generating radioactive waste at the encapsulation plant, as well as the waste volumes and activities have been assessed by Paunonen et al. (2012). The assessment of the ac-tivity inventory and waste volumes has been updated for this report. In the encapsulation plant, the waste management systems are mainly located at the ele-vation +10.30. According to the present design, liquid waste is dried in drums, and the required facilities for the in-drum drying and consequent interim storage are significant-ly smaller than the earlier provision for the concrete-based solidification. Several systems with different layout requirements are involved in the radioactive waste management and shall be taken into account in the design of the encapsulation plant. These systems have the following functions:

sorting and packing of the waste

in-drum compression of compressible waste

activity measurement of the drums

buffer storage of the packed drums

in-drum drying of liquid waste

segmentation and packing of large components (metals)

storage of empty drums, metal boxes, and drum racks

transport routes of the waste packages (forklift, bridge crane, canister lift)

radiation shielding of the surrounding systems and facilities (especially for dried waste).

The encapsulation plant does not comprise facilities for long-term interim storage of the waste packages. The goal is to transfer the waste packages soon after packaging, via a small buffer storage, to the final repository. At the encapsulation plant, it is possible to store the waste amount generated during approximately one year of operation.

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Figure 1-1 presents a schematic layout of the facilities for the low and intermediate level waste management at the encapsulation plant.

Figure 1-1. Waste handling systems and facilities at the encapsulation plant.

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2 AMOUNTS OF WASTE

2.1 Initial data

The estimated amount of waste arising from the operation of the encapsulation plant is based on the amount of fuel assemblies and final disposal canisters to be handled there. According to the valid construction license granted to Posiva, the amount of spent fuel to be handled at the encapsulation plant and disposed of in the final disposal facility equals to 6500 tU at the maximum, which corresponds to about 3325 final disposal can-isters (Posiva 2015). The amount of fuel canisters allocated to power plant units is rep-resented in Table 2-1. The planned operational time of the encapsulation plant is around 100 years. In the previous report (Paunonen et al. 2012) the operation was assumed to last until 2130, resulting in an operation time of 110 years. Table 2-1. Number of disposal canisters and total amounts of fuel for each fuel assem-bly type.

Fuel assembly type

Number of dispos-al canisters

Total amount of fuel, tU

BWR (OL1-2) 1400 2950 VVER (LO1-2) 750 1050 PWR (OL3) 1175 2500 Total 3325 6500

Radioactive waste is generated mainly in the fuel handling cell, where small amounts of radioactive substances (crud and fission products) may be released from the fuel. These substances may further contaminate the equipment or structures within the fuel handling cell. During the decontamination, maintenance and cleaning operations, liquid and solid waste is generated, and this waste is handled and packed in the facilities for waste han-dling. In this report, the waste generated during the operation of the encapsulation plant is called encapsulation waste. 2.2 Liquid waste

As new information has been gathered after the previous estimate, a decision was made to update the estimated amount of liquid waste. This section presents the updated esti-mate (Table 2-2) of the liquid waste generated at the encapsulation plant, as well as the justifications for the update. Liquid waste includes rinsing and cleaning waters, as well as decontamination liquids and waters from drying of the fuel assemblies. The radiation shielding lid of the fuel transport cask is decontaminated with a wet wip-ing method instead of the water rinsing (Kukkola 2013). The wet wiping is estimated to consume not more than 1 l/m2 of water, resulting in water consumption of 1.3 l per lid (A=1.3 m2). Consequently, wet wiping of 3325 lids consumes 4323 l of water, i.e. ap-proximately 0.04 m3/a. Suikki et al. (2007) have estimated 7.5 kg of water being removed from the fuel assem-blies during drying of 12 BWR bundles or 4 EPR bundles. The spent fuel from Loviisa will be transported to the encapsulation plant in dry casks, and thus no separate drying

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at the encapsulation plant is assumed for it. Hence, the amount of OL1-3 fuel to be dried at the encapsulation plant corresponds to 2575 final disposal canisters, resulting in 19,313 kg (2575 · 7.5 kg) of liquid waste in total, i.e. 0.2 m3/a assuming a 100 y operat-ing time. More extensive maintenance works are planned to be carried out in the fuel handling cell about every five years, including the decontamination of the entire fuel handling cell as well as the decontamination centre. Instead of water rinsing of these facilities as suggested in the previous plan, wet wiping method is to be applied, resulting in a reduc-tion of liquid waste volume. If the floor and wall area of the fuel handling cell is 500 m2, the decontamination consumes 0.5 m3 (1 l/m2) of water every five years, i.e. 0.1 m3/a in average. The surface area of the decontamination centre is almost double compared to the fuel handling cell, thus the water consumption there due to the wet wiping is estimated to amount 1 m3 every five years, i.e. 0.2 m3/a in average. Also the fuel drying chambers are decontaminated whenever necessary, probably a few times during the operation of the plant. Water rinsing is an applicable method for that. Prior to water rinsing the loose crud particles shall be removed from the drying chamber and the fuel rack as efficiently as possible to be disposed of in a fuel canister. Waste handling is not dimensioned to receive all the crud into the drainage system and waste drying system. It is assumed that a drying chamber is rinsed three times during the op-eration of the encapsulation plant and the water consumption is 50 l each time. This results in an insignificant liquid waste amount compared to other waste streams, and is neglected in the waste volume estimate. Suikki (2013) describes the access system into the fuel handling cell. According to Suikki (2013), the protective suits are not rinsed when exiting the fuel handling cell, but the outermost protective suit is removed in the washing room, and put to the collection container to be either washed or disposed of. If the body contamination monitor alarm is triggered, contaminated personal protection equipment is removed and contaminated skin is washed in the washing room. Therefore the waste amount arising from the rins-ing of protective suits is removed from the waste amount estimate. As a difference from the original plan, encapsulation plant no longer includes its own laundry, but TVO's laundry is planned to be utilized. The estimated amount of laundry water (5 m3/a) was a significant fraction of the total amount of liquid waste (15 m3/a), and hence removing the laundry significantly reduces the amount of liquid waste pro-duced at the encapsulation plant. After the closure of the TVO's laundry, own laundry is quite easily arranged in the encapsulation plant, if needed. Contaminated overalls can also be disposed of. The transport trailer, transport cask and its shock absorbers are rinsed at the encapsula-tion plant, if contamination is detected on them. Contamination is unlikely, except in some accidental cases, and therefore the waste arising from the rinsing is not included in the waste amount estimate of the normal operation. It is estimated that around 300 l of decontamination solutions are generated in the de-contamination centre per month, resulting in 3.6 m3 of liquid waste annually (Kukkola

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and Eurajoki 2009). The total amount of liquid radioactive waste accumulated during a year including maintenance works in the fuel handling cell is about 4 m3. This accumu-lation rate results in 12 drums of dried waste during the operation time of 100 years. Table 2-2. Amounts of liquid wastes, previous (Paunonen et al. 2012) and updated es-timates.

Origin of waste

Paunonen et al. 2012

(m3/a)

Updated estimate

2015 (m3/a) Decontamination of the radiation protection lid of the transport cask 0.1 0.04 Drying of fuel assemblies 0.5 0.2 Rinsing of overalls 2 - Decontamination liquids 4 3.6 Decontamination of the fuel handling cell 1 0.1 Decontamination of the decontamination centre 1 0.2 Laundry water (if contaminated) 5 - Rinsing of the transport trailer, transport cask and shock absorbers (if contaminated) 1 - Total 14.6 4.14 pcs/110 a pcs/100 a Total amount of dried liquid waste drums pro-duced during the operation of the encapsulation plant 39 12 m3/110 a m3/100 a Total packed volume of dried liquid waste pro-duced during the operation of the encapsulation plant 9.6 3

2.3 Solid waste

The accumulation estimate of the solid encapsulation waste is revised and the assumed operation time of the encapsulation plant is updated to 100 years (previous assessment: 110 years), corresponding to the construction license application. In addition to that the ratios of the inner and outer volumes of the drums and metal boxes have been revised, resulting in a change of the amount of packed waste, although the amount of unpacked waste remains unchanged (Table 2-3). The most significant change in the processes of the encapsulation plant after the previ-ous assessment is the change of the welding method from electron beam welding to fric-tion stir welding. This affects the waste amount of copper chips resulting from the ma-chining of the weld. Machining after friction stir welding is estimated to generate around 200 kg of copper chips (Purhonen 2014a). This means that machining of all the final disposal canisters (3325 pcs) generates 665 t copper chips, which are not assumed to be contaminated, and which therefore can be cleared from regulatory control and re-cycled.

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If the weld of the final disposal canister does not meet the quality requirements, it is ma-chined open, and the fuel is transferred into a new canister. Such an exceptional procedure is estimated to take place once a decade, i.e. 10 times during the operation of the plant. Opening the weld generates roughly 30-38 kg (40 l, tested with a small amount of chips) copper chips (Purhonen 2014b), which are not expected to be contaminated. In total the waste due to opening of the welds is minor, corresponding to about two drums during the entire operation period of the plant. The inserts of the re-opened canisters will be disposed of in the LILW repository. The volume of one insert is 3,3 m3 on average. The amount and activity of crud are updated according to the new data (see section 9.2). The majority of crud is disposed of with the spent fuel in the final disposal canisters. Due to a typo, the estimated annual accumulation rate (0.11 kg/a) of crud by Paunonen et al. (2012) was too low by a factor of ten. It should have been 1.1 kg/a. The new, up-dated value is 0.33 kg/a. The exhaust air from the controlled area of the encapsulation plant is filtered via a HEPA filtration system, if activity is detected in the exhaust air (Posiva 2014a). The cooling and filtration system of the fuel handling cell also includes HEPA filtration, which is triggered whenever fuel is handled in the handling cell or if the limit value for the air activity concentration is exceeded (Posiva 2014b). In addition to the HEPA filter all the lines mentioned above contain also prefilters. It is assumed that the filters are changed every five years. The HEPA filters of the controlled area of the encapsulation plant are assumed to remain clean from contamination during normal operation and to be cleared from regulatory control as waste. However, the filters from the filtration sys-tem of the fuel handling cell are assumed to be contaminated. In total, 40 such HEPA filters and prefilters (20 + 20) are estimated to accumulate during the operation of the encapsulation plant. The HEPA filters likely do not fit into the drums but need to be disposed of in metal boxes. The assumed volume of a HEPA filter for the waste volume estimate is 0.16 m3, and that of a prefilter is 0.08 m3. Dismantling the filters from their frames is not practical, since it may disperse the contamination. The packing density of the used and that of contaminated components to be disposed of has been updated for the present assessment from 500 kg/m3 to 1000 kg/m3. This updat-ed value is based on the experience from the packing of the metal sent from the Loviisa NPP to a melting plant. This change halves the amount of the metal boxes from the pre-vious assessment. In average, 1 m3 of organic waste (mainly wiping rags, protective equipment etc.) is assessed to be generated annually per person. This value includes also the more exten-sive maintenance works every five years, which generate a larger amount of waste. As-suming 10 persons to work in the controlled area of the encapsulation plant, the amount of organic compressible waste is 10 m3/a. A compression ratio of 1:5 leads to a volume of 2 m3/a of compressed waste. The total amount of packed solid waste is reduced significantly from the previous esti-mate. The total estimated amount of solid and liquid waste, when packed, is 520 m3 during the operation of the encapsulation plant, whereas the previous estimate was 1270

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m3. The amounts of waste, unpacked and packed, as well as a comparison to the previ-ous estimate are presented in Table 2-3. Table 2-3. Amounts of solid and liquid operational waste, unpacked and packed, previ-ous estimate (Paunonen et al. 2012) and updated estimate. Paunonen el al. 2012 Updated estimate 2015 unpacked packed unpacked packed Waste type m3/a kg/a m3/110 a m3/a kg/a m3/100 a

Crud 1.1(*

120 kg, disposed in the canisters 0.33

33 kg, disposed in the canisters

Filters 1.5 220 0.05 6.5 Copper chips 168 22 - - - Organic waste (compressible) 15 330 10 224 Components 2500 660 2500 254 Rejected canister inserts 0.33 37 0.33 33 Total 17 2670 1269 10.4 2500 518 Liquid waste 14.6 9.6 4.14 3 Total, all waste 1279 521 *) There was a typo in the previous crud estimate (Paunonen et al. 2012); it was 0.11 kg/a, but the right value is 1.1 kg/a. 2.4 Decommissioning waste

The amount of decommissioning waste of the encapsulation plant have been evaluated in the report (Kaisanlahti 2012). A new evaluation have been made using the same prin-ciples as in operational waste assessment; packing density of the components, revised dimensions of drums and metal boxes, and specified dimensions of the HEPA filters and pre-filters. The waste amounts are presented in Table 2-4. The waste amounts are the same as in the previous assessment concerning components, organic waste and transport casks inserts. Only the liquid waste and filters have been re-evaluated. It is estimated that filters in the fuel handling cell air conditioning will be changed twice during the decommissioning, which means two HEPA and two pre-filters to be disposed of in the metal box. Because the total volume of the filters is small, only 0.5 m3, they will be disposed of among other waste, e.g. metal waste. Liquid waste is produced during the decontamination of the fuel handling cell and the decontamination centre, as well as during the decontamination of components taken out from the fuel handling cell. Active drainage water system will also be decontaminated. It has been estimated, that totally 10 m3 of liquid waste will be produced during the decommission-ing of the encapsulation plant.

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Table 2-4. Amounts of solid and liquid decommissioning waste, unpacked and packed, previous estimate (Kaisanlahti 2012) and updated estimate. Estimate 2012 Updated estimate 2015 Waste type Unpacked Packed (m3) Unpacked Packed (m3) Components 75000 kg 157.3 75000 kg 78 Filters 3.5 m3 6.4 0.5 m3 0.5 Liquid waste 100 m3 1.1 10 m3 0.24 Organic waste 70 m3 17 70 m3 16 Transport casks inserts 25 m3 25 25 m3 25 Total 206.8 120 2.5 Materials ending up in the LILW repository

Table 2-5 presents an estimate of the amounts of organic waste, cement, steel and iron, ending up in the LILW repository during the operation and decommissioning of the encapsulation plant. During the operation, the cement amount includes the floor of the repository hall (thickness 0.4 m), shotcreting of the hall (thickness 80 mm) and the con-crete basin. The filling of the concrete basin will be done after the decommissioning. Concrete is assumed to contain 420 kg/m3 of cement (Nummi 2012). Cement amounts are based on the size of the LILW repository presented in (Paunonen et al. 2012). The steel amount includes the components to be disposed of and waste packages. Transport casks inserts are included in the decommissioning waste (steel). Iron amount includes the rejected fuel canister inserts and the reinforcement of concrete. The latter is estimat-ed not to exceed 7% of the cement amount (Karvonen 2011). The size of the repository hall was assumed to remain unchanged, even though the waste volumes have decreased (see section 5). Table 2-5. Estimated amounts of materials ending up in the LILW repository during the operation and decommissioning of the encapsulation plant. Material Operation (t) Decommissioning (t) Total (t) Organic material 74.7 5.3 80 Dried liquid waste 2.35 0.05 2.4 Cement 156 13 169 Steel 295 99 394 Iron 144 - 144

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3 WASTE HANDLING SYSTEMS FOR LIQUID WASTE

The handling systems for liquid waste are the collection and treatment system of drain-age water of the controlled area and drying system for radioactive water. The drying system for the liquid radioactive waste at the OL3 NPP unit has been presented as an alternative to drying liquid waste from the encapsulation plant. Furthermore, this section discusses the treatment of such liquid waste, for which drying is not an applicable treatment method. 3.1 Collection and handling system of drainage water of the controlled area

The estimated accumulation rate of radioactive liquid waste at the encapsulation plant is 4 m3/a. In addition about 60 m3/a of cleaning water, not expected to be contaminated, originates in the controlled area. A majority of the drainage water generated during the normal operation can most probably be released from the encapsulation plant due to its low activity level. The drying system of the radioactive water is dimensioned according to the waste generation rate of 15 m3/a, which is based on the original accumulation estimate of the liquid waste (Paunonen et al. 2012). Changing the dimensioning of the system according to the updated waste amount estimate is not deemed necessary. The collection and treatment system of drainage water of the controlled area (PK.341) contains two collection tanks. One of the tanks (Figure 3-1 red tank) receives the most radioactive drainage water from the controlled area, such as those from the fuel han-dling cell, decontamination centre, hot workshop, drying system of the fuel and the room for drying the radioactive liquid waste. The other tank (orange tank) receives the rest of the drainage water from the controlled area. These are mainly cleaning water expected to contain so little radioactivity that they can be led via the drainage system of uncontrolled area of the encapsulation plant (PK.765) to the disposal facility groundwa-ter release system (P.767). In addition to the collection tanks, the drainage water system contains two intermediate tanks, one for the water below a set level to be led out from the plant (green tank), and one for the radioactive water to be dried (purple tank). The volume of all of these tanks is 5 m3, and the tanks are located at level -2.90.

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Figure 3-1. The collection and handling system for drainage water of the encapsulation plant. The red tank is for waters that are likely to be radioactive. The orange tank is for cleaning waters that are assumed free of radioactivity. Water is released from the en-capsulation plant from the green tank and radioactive water is fed into to the drying system from the purple tank. The treatment of liquid in the collection tanks (red and orange tanks) is based upon the radioactivity determination. The activity is determined by circulating the water in the collection tank, and thereafter taking a sample to be analysed in a laboratory, e.g. at TVO. During the sampling, the drainage lines into the collection tank are closed and water is prevented from entering the drainage system by administrative means. If the content of the collection tank for water expected to be inactive (orange tank) does not exceed the level set for its release, the following procedure is followed: the water is pumped to the intermediate tank for clean water (green tank), and from there via contin-uous activity measurement (PK.554) to the floor drainage system of the uncontrolled area, PK.765. If the release level is exceeded (not likely during the normal operation of the plant), the water is pumped to the active water collection tank (red tank). From the active water collection tank the water is pumped first to an intermediate tank (purple tank), and then further to a dosing tank with a volume of 200 l at level +15.10. From there, the water is led gravitationally to the drying system of radioactive liquids (PK.343). If the content of the collection tank for active water (red tank) does not exceed the level set for its release, the water is pumped to the collection tank for clean water (orange tank), and from there via the intermediate tank (green tank) out of the plant to the dis-posal facility groundwater release system (P.767). There is no direct connection from the collection tank for active water (red tank) to the intermediate tank for clean water (green tank), thus preventing the active water to be brought to the intermediate tank for clean water (green tank), e.g. as a result of a human error. The connections between the tanks are presented in Figure 3-1. At the same level with the collection tanks, there is a pumping pit (volume 5 m3) for the drainage water, with a connection from the collection

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and intermediate tanks. If any of the tanks leaks, the content of the leaking tank ends up in the pumping pit, from where it can be pumped back to either collection tank. 3.2 Drying system of radioactive drainage water

The active waste water is dried in 200 liter steel drums. The waste is dried using heating elements surrounding the drum, and a vacuum pump to enhance the evaporation. The collection system of drainage water and transferring the water to the drying equipment are described in section 3.1. The drying system is situated in the waste handling facili-ties, at level +10.30. The drying process is automatized, and does not require continuous manual control. An example of a drying equipment is presented in Figure 3-2.

Figure 3-2. Drying system for active liquid waste; on the left two drying units and on the right the control unit. (Picture: Dennis Brunsell, Diversified Technologies Services Inc., email 16.4.2014). An empty drum is manually set into the position surrounded by the heating elements. On top of the drum, a lid element with inlet for the liquid waste and outlet for the steam is automatically attached and sealed. After attaching the lid element, the automatized drying process is started. The automatic system measures the liquid level and dose rate of the drum, and controls the flow to the drum, the heating and vacuum pump. The steam exiting the drum is condensed and collected to the collection tank to be further led to the collection tank for presumably inactive water via the collection system of the drainage water of the controlled area. The drying process continues with cycling filling and drying as long as the drum is almost full of dried waste. Thereafter the drum is fur-ther heated to ensure sufficiently low moisture. The drying capacity of the reference equipment presented in Figure 3-2 is 7-11 l/h. Due to low waste generation rate, filling of one drum takes almost nine years, based on the assumptions that the salt concentration of liquid waste is 10 g/l in average, and that

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the majority of the salt is sodium chloride (NaCl, density 2164 g/l). The filling time is very long, and therefore using smaller drums could be considered. At the moment, most commercially available in-drum dryers are designed to use 200 l drums, but a modifica-tion for a smaller drum size is believed to be possible. The surface dose rate of the drum is measured continuously during the drying process. If necessary, the waste inlet rate into the drum is limited to keep the dose rate below a target value. During the transport and storage, the drum shall be shielded in such a way that the dose rate on the surface of the radiation shield does not exceed 2 mSv/h. After drying the drum is lifted with a bridge crane into the radiation shield, and the drum lid is installed manually. Finally, the lid of the radiation shield is closed. Using the bridge crane, the drum is transferred within its radiation shield to the vicinity of the gamma scanner (see section 4.4.1) for activity measurement. The lid of the radia-tion shield is opened, and the drum is lifted via remote control from the radiation shield to the gamma scanner. After the measurement, the drum is returned into the radiation shield, and transported to the LILW repository or to the interim storage in connection to the waste handling facilities, from where it will later be transported to the LILW reposi-tory. Due to the possibly high dose rate of the dried waste, the drying system of the radioac-tive liquid waste shall be situated in a shielded room. The control unit of the drying sys-tem is to be situated in a separate room.

3.2.1 Provision for using drying systems in OL3 unit

TVO's OL3 unit is provided with a drying equipment of radioactive liquids that might be utilizable for the treatment of the radioactive water from the encapsulation plant. Especially during the early phases of the operation of the encapsulation plant, this alter-native is recommended, allowing operational experience from the accumulation of the radioactive water to be gained. If the accumulation rate of liquid waste or average salt concentration therein were lower, the filling time of the drum dried at the encapsulation plant would increase even further. With the operational experience, a better applicable treatment method could be selected. A provision for utilization of the drying system of OL3 is made at the encapsulation plant by installing a pipeline from the intermediate tank for active water to the transfer corridor for the fuel transport cask, situated at level +1.90. There the liquid waste can be pumped into a transfer container, lifted to the reception room of the fuel transport cask, and further transferred to a vehicle for transportation to OL3. At OL3, there are nozzles for leading the liquid waste to the drying system. The procedures (design and handling of the container, licensing etc.) related to the transfer and transport of liquid waste shall be designed separately. For example, at the Ringhals NPP three different transport containers made of steel are used to transport the liquid waste within the power plant area from the power plant units to a centralized waste treatment facility at Ringhals 1. One lead-shielded container is for sludge, and contains a sieve. Another lead-shielded container is for intermediate level ion exchange resins, whereas the third container having no shield is meant for low level

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resins. The weights of the containers are 1.8 - 5.5 t. According to the power plant the user experience of the containers is good. 3.3 Handling of other liquid waste

In addition to the radioactive drainage water, maintenance and decontamination opera-tions at the encapsulation plant could result in generation of small amounts of other ra-dioactive liquid waste, such as oil or solvents. At the Olkiluoto and Loviisa power plants, the corresponding waste is stabilized with clay-based agents. Stabilization is done in 200 l drums. For example Fluid Tech Inc's silicate-based pow-ders (Aquaset/Petroset) can be used as solidification agents. The dosage of the powders is determined in solidification tests. The solidification agent is mixed with the waste using a drum mixing device, one of which is available at the Olkiluoto power plant. The solidification product is left to dry with the lid open. When the solidification product is dry, the lid is installed, and the drum is transferred to the gamma scanner for activity measurement. This method can be applied also at the encapsulation plant, due to the small amount of waste, it might be reasonable to utilize the facilities and equipment of the Olkiluoto power plant as long as it is in operation, and thereafter acquire a drum mixer to the en-capsulation plant. It is assumed that the solvents and oils to be solidified are low level waste, so they are disposed of in a similar manner as the encapsulation waste drums. If necessary, the most active drums can be shielded with a radiation shield for the dried waste during interim storage and successively transported and disposed of in the concrete basin with the dried waste.

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4 HANDLING OF SOLID WASTE

4.1 Sorting of waste

The solid waste is collected and sorted into waste bags, at the place, where the waste is generated. The bags are transferred to the waste treatment facilities at level +10.30. In the waste packing room, the counting rate of the waste bags is measured with a handheld radioactivity meter in order to have them sorted into those that can be cleared from regulatory control and those that are to be disposed of as radioactive waste. At this point, the most active waste fractions can be removed from the bags that otherwise could be cleared from regulatory control. Compressible and non-compressible waste is sorted into separate 200 l steel drums. Compressible waste is compacted into a smaller volume with a waste compactor. Non-compressible waste is packed directly into 200 l drums, except the large metal pieces that are packed into metal boxes. For example, at the Loviisa power plant, the weight of the packed maintenance waste drums varies typi-cally between 50 kg and 200 kg, depending on the amount and density of the waste. The active waste is packed into the drums in such a manner that the activity is distribut-ed as evenly as possible in the drum. However, if a bag with a higher dose rate than the others should be packed into a drum, it should be put in the middle of the drum. When a drum is full, a lid is installed, and the drum is transferred to the gamma scanner for activity measurement. After the measurement, the drums are set to racks with four drums, and stored temporarily in a room connected to the waste handling facilities prior to transportation to the LILW repository. It is estimated that 944 drums to be disposed of as radioactive waste are generated during the operation of the encapsulation plant, i.e. about 10 drums per year. 4.2 Compaction of compressible waste

After the compressible waste is sorted and packed into bags, the bags are put into 200 l steel drums, in which the waste is compacted. The waste compactor is situated in the waste handling facilities at level +10.30. A suitable compactor is e.g. 27 tons (60,000 lbs) hydraulic compactor that is able to compact the waste by a factor of about five (Figure 4-1). When the drum is full of compacted bags, the lid is installed, and the drum is transported to the gamma scanner for activity measurement. A shielded area for storing the sorted bags prior to packing into the drums, is located in the vicinity of the waste compactor.

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Figure 4-1. An example of hydraulic waste compactor (55E, www.ramflat.com). 4.3 Handling of metal waste and large components

Metal parts and components are decontaminated in the decontamination center, if neces-sary, and large components are segmented, if it is possible and reasonable. If a compo-nent does not fit into a drum, it can alternatively be packed in a metal box of approxi-mately 3 m3 (1.3 × 1.3 × 1.9 m) in volume. The packing density of the metal boxes is assumed to be 1000 kg/m3 (based e.g. on the packing density of the metal sent from the Loviisa NPP to a melting plant). Hence, the gross weight of a packed metal box is 4000 kg at its maximum. It is estimated that the accumulation of metal and components to be disposed of in metal boxes corresponds to about 80 boxes during the operation of the encapsulation plant, or less than 1 box (< 3 m3) per year. There shall be appropriate facilities for segmenting and packing the metals. At the en-capsulation plant, these measures can be taken e.g. in the decontamination center or in the hot workshop, i.e. contaminated components are not needed to be transported long distances. The segmented pieces can be packed directly into drums, to be further trans-ferred via inactive workshop to the waste handling facilities. Larger pieces can be trans-ferred e.g. on a pallet, and packed into metal boxes in the waste handling facilities. When filled, drums and metal boxes are transported into the LILW repository by the canister lift. At the Loviisa power plant, metals are segmented with hydraulic cutters (pipes, small metal structures and pieces), band saw (thicker rectangular or cylindrical pieces) and plasma cutter (pieces, whose size or shape prevent the use of hydraulic cutters or band saw). At the Olkiluoto power plant, metal is treated with a Tyrannosaurus® shredder, cutters, oxy-fuel cutting, and a 200 t compactor, the latter of which is used e.g. for com-pacting pipes into a smaller volume.

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The components with dimensions larger than the metal box and that cannot be segment-ed are disposed of as such and protected to avoid dispersion of loose contaminated par-ticles. At the moment, the intact final disposal is planned at least for the internal parts of the transport casks, as well as possibly for the inserts of the rejected final disposal canis-ters, the welds of which do not meet the requirements. Other large components, the ac-tual size or segmentation possibilities of which are not known yet, comprise of the fuel handling equipment in the fuel handling cell. Large components are transported to the repository along the access tunnel. A possibility, worthwhile considering, is sending the metal waste to an external melting plant. At the Olkiluoto nuclear power plant this option has been utilized. Also from the Loviisa power plant one batch of metal scrap was sent to melting. According to the Loviisa experience, the volume of the returned secondary waste was 8% of the original waste volume. 4.4 Activity measurement and records management of packaged waste

4.4.1 Activity measurement of drums

The packed waste drums are analyzed with a gamma scanner. The equipment comprises of roller conveyor, weight measuring table, collimator, gamma detector and dose rate sensor. A workstation and a printer are also needed in the same room. The software in the workstation controls the mechanical equipment of the gamma scanner and analyses the spectrum. The gamma scanner shall be situated in a room with low background ra-diation. Figure 4-2 shows the gamma scanner of the Loviisa power plant.

Figure 4-2. Activity measurement system for drums at Loviisa NPP (Picture: Vesa Laakkonen, Fortum).

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In the characterization of the waste, the uncertainties in the quantitative determination of alpha and beta emitters should be duly considered. The half-lives of the typical key nuclides, Co-60 and Cs-137, are short in comparison to the interim storage time of the spent fuel, and therefore the quantitative determination of the alpha and beta nuclides is not as straightforward as at the power plants. The waste drums are transported to the LILW repository in campaigns. For the transport, the drums need to be sealed and free from external contamination. The con-tamination can be determined e.g. by wipe sampling of each lot to be transported to the LILW repository.

4.4.2 Activity measurement of metal boxes and large components

The activity measurement of the metal boxes and large components is not possible with the same gamma scanner that is used for the measurement of drums. One way to deter-mine the activity is acid wipe sampling. The samples should be taken to be representa-tive for the average and highest surface activity of the objects. Several samples are tak-en from larger objects. Acid wipe samples are analyzed in a laboratory. Prior to the transport of the metal boxes to the final disposal facility, the cleanliness of their outer surfaces is ensured by wipe sampling. Equipment for the activity determination of large components or waste packages is also available on the market, but, as it is rarely used, it is not very cost-efficient to purchase it for the encapsulation plant. At the moment such equipment is neither available at the Olkiluoto or Loviisa power plants. The measurement of metal boxes is possible to be procured as a service, offered by external service providers. If the metals are submitted for melting, the melting facility provides also an activity inventory.

4.4.3 Records management of waste

After the analysis, records of the waste packages are made in the waste records man-agement. The nuclide wise activities are transferred automatically from the gamma scanner to the records management. The purpose of the waste records management is to have an up-to-date information on the amount, type, radioactivity and location of the waste stored or disposed of in the repository. Regarding the waste cleared from the reg-ulatory control, the amount, type and radioactivity is to be recorded. According to YVL Guide D.4 (STUK 2013a), at least the following shall be recorded with regard to waste packages transferred into a storage:

waste type and the amount of waste; treatment and conditioning method and year; waste package identifier and storage location; activities of dominant nuclides, potential surface contamination, and the date of

activity determination; classification as nuclear use item or other exceptional composition; origin and owner of the waste; and any other information required for wastes to be disposed of.

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After the analysis, a package sticker is attached to the lid of a drum or a metal box, con-taining at least the identification of the package, packing date, and waste type. Also add-ing the weight of the package is recommended. In the package stickers of the drums containing waste stabilized with clay-based agents, the type and amount of stabilization agent should be included. According to YVL Guide D.5 (STUK 2013b), the holder of an operating license of a disposal facility shall maintain records of the disposed waste, providing at least the following information to an accuracy of an individual waste pack-age:

the waste type, its processing and packaging method and structural and material characteristics significant to safety;

a waste package identifier and location in the emplacement room; the upper limits for the activities of the dominant nuclides, to an accuracy of an

individual disposal canister in case of spent fuel and to an accuracy of an indi-vidual emplacement room in case of other waste; and

It is recommended that the possibility to combine the waste records management with fuel assembly data is considered.

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5 FINAL REPOSITORY FOR ENCAPSULATION WASTE

The low and intermediate level waste (LILW) generated during the operation and de-commissioning of the encapsulation plant are disposed of in the final repository for the operational waste (LILW repository) (P.149) at the depth of about 180 meters. Paunonen et al. (2012) assessed the required volume of the LILW repository to be 3750 m3. The present estimate of the generated waste volume (640 m3, decommissioning waste included) is roughly a half of the estimate in the reference above (1500 m3), and hence it is estimated that also the required final disposal volume could be half of the estimate given by Paunonen et al. (2012). The same applies to the concrete basin de-signed for the final disposal of the dried waste. Based on the updated estimate of the waste volumes, the volume of the LILW repository hall would be approximately 1600 m3 (space requirement factor 2.5).

5.1 Waste handling systems in the repository

At the LILW repository for the operational waste, the waste packages are handled with a forklift and bridge crane. From the encapsulation plant the waste packages are mostly transported to the LILW repository level using the canister lift. The encapsulation waste drums are transported in racks of four drums or separately one by one, and the dried waste drums are transported inside a cylindrical radiation shield. The metal waste is transported in drums or metal boxes. At the LILW repository level the waste packages are transferred from the canister lift to the loading area of the bridge crane with a forklift. The lifting capacity of the forklift shall correspond to the one at the encapsulation plant (see section 8). The waste packag-es are moved to their final positions with the bridge crane. The drums containing dried waste are moved by remote control from their radiation shield into the separate concrete basin. The control unit of the crane shall be radiation shielded. Large components with dimensions larger than those of the lift and waste packages are transported via the access tunnel on a transport trailer or a corresponding vehicle to the LILW repository, and further with the bridge crane to their final positions. The lifting capacity of the bridge crane shall correspond to the heaviest component to be disposed of. This may be the canister insert (for EPR fuel 18 tons), if it cannot be decontaminated for clearance. 5.2 Challenges related to the long operational lifetime of the repository

The estimated operational time of the encapsulation plant is 100 years. During opera-tion, the waste is disposed of continuously, and therefore the LILW repository needs to remain open until the decommissioning of the encapsulation plant is completed. This results in requirements for the environmental conditions in the final disposal repository as well as for the waste packages, since the waste packages shall not essentially degrade before the closure of the repository. The condition of waste packages shall be monitored during the storage period (STUK 2013a). Based on the experience from the Loviisa LILW repository, there can be corrosion in the drums already after a storage period of two decades. The leakage of water into the

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repository and air humidity have caused also other problems with materials and equip-ment. Therefore the control of the leakage water plays a crucial role, and especially the leakages to the waste drums should be prevented. For example, in the maintenance waste tunnel 1 (HJT1) of the Loviisa LILW repository, the drums are protected with a tarpaulin, and in the maintenance waste tunnel 2 (HJT2) there is a ceiling to prevent water leakages to the drums. At the moment, a study of al-ternatives to re-handle the waste packages on the HJT1 and HJT2 is under progress. One alternative is to sort the waste packages according to when they can be cleared from regulatory control. At the same time as sorting, the waste in damaged packages could be re-packed in intact drums. Plans for the control of the leakage water in the maintenance waste tunnel 3 (HJT3) of the Loviisa LILW repository are still under progress, but one of the considered alterna-tives is a tunnel sealing (a ceiling of plastic fabric). This solution has been adopted e.g. in SKB's LILW repository SFR-1, where material and aging problems have also been encountered. SFR-1 was commissioned in 1998 and, according to the original plans, it should have been closed in 2009. However, according to the present plans, the closure takes place as late as in 2072, and therefore a renewal process of equipment and struc-tures has been initiated. The tunnel fabric roof was installed in SFR-1 in 2010. Even if favorable conditions for the waste packages in the LILW repository were achieved, it is possible that the drums and metal boxes do not withstand a storage time of over 100 years. A few alternatives to limit early failures are presented below.

1. A provision is made for unloading and re-packing the waste packages brought to the repository. There shall be sufficiently space for transferring, sorting, and un-loading. Special attention is to be given to the control of leakage of water into the repository and air humidity.

2. The waste drums are packed into concrete boxes, as is the current practice at TVO. Prior to that the drums can the compacted to a size about half of the origi-nal. Concrete boxes limit effectively the release of the radioactive substances from drums even if they were corroded. The net volume of the larger concrete box in use at the Olkiluoto power plant is 5.2 m3 (gross volume 6.8 m3), and it takes 32 compacted drums. The 1011 drums (including decommissioning waste) fit into 32 boxes, resulting in a total volume of 218 m3 (to compare with the volume of 1011 drums, equal to 240 m3). The amount of concrete in 32 boxes is 51 m3.

3. All the waste from the encapsulation plant is handled at the Olkiluoto power plant, and disposed of in the TVO's LILW repository for operational waste (VLJ-repository). In this case TVO's VLJ-repository needs to be kept open long-er than planned at the moment. Alternatively, a new repository can be construct-ed prior to the closure of the TVO's VLJ-repository.

4. An interim storage facility is constructed aboveground for the operational waste in connection or in the vicinity of the encapsulation plant. It is easier and less expensive to keep the conditions optimal aboveground than underground. The LILW repository could be constructed later. During the aboveground storage, experience would be gained on the waste accumulation, allowing it to be utilized for the optimization of the design and construction of the LILW repository.

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The alternatives shall be assessed from several points of view, such as legislation and requirements for the waste storage and final disposal, safety issues, security arrange-ments, costs, and efficiency of the operations.

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6 LIMITING THE AMOUNTS OF WASTE

According to the guiding principle in the Nuclear Energy Act, the amount of nuclear waste generated in the use of nuclear energy shall be as low as reasonably possible with practical measures, regarding both activity and amount (YEL 1987). Furthermore the requirement 402 of YVL D.4 (STUK 2013a) states that "the generation of waste that needs to be stored or disposed of shall, among other things, be limited by means of re-pair work and maintenance planning, and decontamination and volume reduction". Limiting the amount of waste is based among other issues on environmental aspects. With the design, it is possible to significantly influence the waste amounts. Since all the waste generated in the controlled area is to be handled as radioactive waste, it is crucial that no excess material is brought to the controlled area. All the excess pack-ing material (box, factory package, transport pallet, rain protection etc.) as well as bind-ers meant to be removed before the use of a product, and not necessary for the remain-ing transfers, shall be removed from all the deliveries to the controlled area. The generation of wood waste (such as transport pallets or boxes) shall be minimized, since wood can neither be decontaminated nor compressed with waste compactor. Wood also makes up excessive fire load. Also the use of plastic wraps shall be limited and considered case by case. Protecting floor surfaces (by plastic foil) is often to be avoided, since it is more reasonable to wipe the contaminated surfaces. The floor sur-face treatment should allow easy cleaning. Compressible waste is compacted with a waste compactor to a smaller volume, thus reducing the total volume of the waste to be disposed of, the number of waste packages and finally the required repository volume. With the decontamination of contaminated equipment and components, it is possible to decrease the amount of (metal) waste to be disposed of, and enable their reuse and recy-cling. However, the use of decontamination solvents shall be kept reasonable. The amount of liquid waste can be reduced by avoiding excessive use of water in the cleaning, maintenance and repair works. It is often sufficient e.g. to wipe the surface with fabrics. This is especially important in the decontamination measures of the fuel handling cell.

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7 CLEARANCE OF WASTE

7.1 Regulations regarding clearance

The Radiation and Nuclear Safety Authority STUK has given guidelines for clearance procedures in the YVL Guide D.4 (STUK 2013a). The clearance procedures can be general or case-specific, which have different clearance levels. In a general clearance procedure, the destination of the materials released from the facil-ity need not be designated, or is only designated in its outline, and the activity levels to be applied are fixed. The general clearance procedure is not applicable to waste that is volatile or flammable or is otherwise particularly prone to cause radiation exposure. In a case-specific clearance procedure, the recipient of the material and the maintenance process must be defined; the activity levels will be imposed based on case-by-case as-sessment. In clearance one shall ensure that the activity concentration of the waste is below the prescribed levels given in YVL Guide D.4 (STUK 2013a). In a case-specific clearance the activity concentration shall be below the limits given in the particular decision. The material to be cleared may not contain any nuclear material defined in Section 3(1)(2) of the Nuclear Energy Act. After clearance Nuclear Energy Act is no more applied to the waste and the material is no longer considered nuclear waste. 7.2 Clearance procedures at the encapsulation plant

7.2.1 Clearance of drums containing solid waste

As the encapsulation waste generated at the encapsulation plant does not contain short-lived nuclides, interim storage for a later clearance is not worthwhile. The waste to be cleared from regulatory control is separated during sorting and packed into drums. If considered reasonable, the waste can be compacted to a lower volume. Waste drums are measured with a gamma scanner. A drum can cleared from regulatory control, if the activity is below the limits given in the YVL Guide D.4 or in a decision for the case-specific clearance. After the measurement, the drums to be cleared from regulatory control can be transferred to a buffer storage to wait for the transport to the landfill, incineration or other treatment.

7.2.2 Clearance of metal waste

For the clearance of metals a control measurement is done with a handheld radiation meter which detects gamma and beta radiation. If no deviation from background is ob-served, the item is packed e.g. into a drum. Large items, not considered appropriate for segmenting, can be transferred on pallets. If a handheld radiation meter alarm is triggered in the control measurement, the contam-ination is localized and removed, if possible. If the contaminated item cannot be re-moved, it is cut into pieces and the contaminated parts are packed into waste containers for final disposal. After treatment of all the contamination-free metal, the waste lot is

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transported through a radiation detection gate for vehicles, if one is available. If the gate gives an alarm, the load is returned to the controlled area. The item causing the alarm is localized with a handheld radiation meter and removed from the load. If no radiation detection gate for vehicles is available, metal waste can be measured with the gamma scanner for drums as long as the waste items can be packed into drums. If this is not the case, acid wipe samples can be taken. The samples should be taken to be representative for the average and highest surface activity of the objects. Several samples are taken from larger objects. Based on the analysis of the acid wipe samples, waste lot is either cleared from regulatory control or disposed of as radioactive waste. The cleared waste is transferred to an interchangeable truck bed or a corresponding place to wait for the transport to recycling, and the empty waste packages are returned to the storage space reserved for them. According to the present information, TVO is considering a procurement of a radiation detection gate for vehicles that could be avail-able also for Posiva.

7.2.3 Criteria for water released from encapsulation plant

In this report, the limit value for the activity in the water released from the drainage system of the uncontrolled area of the encapsulation plant to the drainage system of the final disposal facility is set to 5 Bq/ml. This limit value does not deviate much from the limit values for water released to the sea from the Loviisa (4 Bq/ml) and Olkiluoto (11 Bq/ml) power plants, and it can be easily verified with measurements. Regarding other characteristics, the water has to meet the quality criteria for the floor drainage sys-tem of uncontrolled area (PK.765) and to the disposal facility groundwater release sys-tem (P.767). The quality criteria will be defined later. It is estimated that about 60 m3/a of cleaning and washing water are released from the encapsulation plant to the disposal facility groundwater release system (P.767). The calculated maximum release rate is hence 0.3 GBq/a (60 m3/a · 5 Bq/ml), but in reality the release is likely to remain significantly below this value. For example the limit for the releases via the cooling water channel from one power plant unit in Olkiluoto is 148 GBq/a (OL1-2 in total 296 GBq/a, incl. other beta active nuclides except tritium). As the releases from the encapsulation plant will possibly be added to the same site-specific quota with the releases from the OL1, OL2, and OL3, a common assessment should be done. 7.3 Other nuclear use items

A licensee shall keep records of nuclear commodities, and their location at the plant shall be known. According to the Nuclear Energy Decree, nuclear commodity means nuclear materials and materials, devices, equipment, nuclear information and agree-ments (including the components of devices and equipment) that prove pertinent to the proliferation of nuclear weapons. Nuclear material means special fissionable materials or source materials, such as ura-nium, thorium and plutonium (e.g. nuclear fuel, fission chambers) suited for obtaining nuclear energy. Other nuclear use items mean nuclear commodity which is not nuclear

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material (e.g. the casks for spent fuel). Nuclear commodity classification has not yet been done for the encapsulation plant, and it is beyond the scope of this report. Other nuclear use items can be disposed of with the rest of operational waste as long as it does not contain nuclear materials. The major part of the other nuclear use items is disposed of as part of the decommissioning of the encapsulation plant, but some com-ponents classified as other nuclear use items may be replaced and disposed of already during the operation of the encapsulation plant. In the disposal of other nuclear use items a component or device shall be made inopera-ble so that it can no longer be used for the original purpose, as required in the decree. The write-off of nuclear devices and equipment and other materials (except nuclear ma-terial) subject to nuclear safeguards shall be reported to STUK at least two weeks prior to the foreseen activity. After write-off of a product, it can be cleared from regulatory control or disposed of as nuclear waste, depending on its radioactivity content.

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8 LOGISTICS

The empty waste containers are brought to the encapsulation plant via entrance in the canister shaft end of the encapsulation plant. About ten drums and one metal box are estimated to be needed per year, which determines the required storage space. The aim is to transfer the drums in racks of four drums, so 2-3 racks are needed annually. From the encapsulation plant the waste packages are mostly transferred to the LILW repository level by the canister lift. The capacity of the lift is 30 tonnes. At least 100 mm gap shall remain between the waste package and the walls of the lift cage. These issues determine the maximum dimensions of the waste packages transferred by the lift. The waste packages are transferred to the lift with a forklift and encapsulation plant lift between levels +10.3+ and +1.90. The capacity shall correspond to the heaviest waste package. The metal boxes may weigh around 4000 kg, but the weight of the heaviest single components may exceed this. Therefore, a lifting capacity of 5000-6000 kg is recommended for the forklift. The forklift shall be able to enter the workshop at level +10.30, since large components should be brought to the waste handling facilities via the workshop. Large components could be transferred to the LILW also via access tun-nel. In that case these components will be transferred to the reception area in the encap-sulation plant by forklift. The forklift is also used in the transfers of waste drums in the waste handling facilities. The drum with dried waste can be transferred with a forklift, when shielded. The unshielded drum is transferred via remote control with the bridge crane, if the dose rate is too high. The waste packages are transported to the final repository in campaigns. Before a cam-paign the waste packages are stored in a buffer storage in the waste handling facilities. The buffer storage has a sufficient capacity for waste generated approximately during one year of operation. If the metal waste meant for clearance were to be measured with the radiation detection gate for vehicles, the waste packages are transferred to a vehicle via the reception area for transport cask. All the waste checked for compliance with the clearance levels is transferred e.g. to an interchangeable truck bed close to the encapsulation plant to be further transported to e.g. recycling. Empty waste packages are returned to the storage space reserved for them. If liquid radioactive waste is transported to OL3 for treatment, the pumping of water to the transport container takes place via a nozzle located in the transfer corridor for the spent fuel transport cask. The container is brought to the reception area of the transport cask, from where it can be transported to OL3. All the waste to be treated at the Olkiluoto power plant units is packed at the encapsulation plant in such a way that con-tamination cannot disperse during the transport. The treated waste is returned back to the encapsulation plant via the reception area of the transport cask, and it is handled in a similar way as the waste packed at the encapsulation plant (activity measurement, inter-im storing and transport to the LILW repository).

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9 ACTIVITY INVENTORY OF ENCAPSULATION WASTE

This section estimates the activities released from the spent fuel during the operation of the encapsulation plant. Thus, the estimate includes the total activity both in the opera-tional and decommissioning wastes. All the radioactivity generated at the encapsulation plant will be disposed of in the LILW repository and the activity inventories are deter-mined at the time of the closure of the LILW repository. The activity content of a drum containing dried waste during the handling and transportation is estimated separately. The former is used as initial data for the safety case of the final disposal and the latter for the safety and radiation shielding analyses of the operational phase. During the encapsulation process, radioactivity may accumulate into the fuel handling cell from the fission products released from the fuel or from the corrosion deposits (crud) spalling off of the surfaces of fuel. The release of fission products takes place as a result of lost integrity of fuel cladding that may originate in the handling or equipment failure in the encapsulation process. Hence the frequency of fuel damages affects the activity inventory of fission products. Crud is assumed to spall off of the fuel surfaces in the normal handling procedures, hence the accumulation of crud activity is more evenly distributed. The majority of the activity that is not disposed of with the spent fuel ends up in the operational waste, but part of it will remain at the plant ending up in the de-commissioning waste. For the activity inventory both conservative and realistic estimates are given. The con-servative estimate represents an upper limit for the activity inventory with a high degree of confidence, but still without the impact of accidents. The realistic estimate is based on best-estimate values. The ratio between the conservative and realistic estimates gives a measure for the uncertainty of the activity inventory. As operational experience is gained during the operation of the encapsulation plant, the uncertainty in the activity inventory tends to decrease. Eventually, as the encapsulation plant is shut down and decommissioned, the waste inventory will be known relatively accurately, and the re-maining source of uncertainty is that of the activity measurements and determination. 9.1 Release of fission products from the spent fuel

9.1.1 Fuel defects in the encapsulation process

Spent nuclear fuel is brought from the interim storage facilities either in by dry or wet transportation. According to the present plans, the transport from Loviisa is dry and the one from Olkiluoto wet. In the fuel handling cell the fuel assemblies are lifted one by one to the drying chamber, where vacuum drying is applied to remove the humidity from the surfaces of the fuel assembly. According to current plans, fuel rods damaged at the power plant units will be packed in a dedicated, water- and gas-tight packages either at the power plant units in Olkiluoto or at the spent fuel storage in Loviisa prior to the transport to the encapsula-tion plant (Sorjonen 2015). The fuel assemblies are lifted from the drying chamber to the final disposal canister docked into the fuel handling cell. For the spent fuel from the Loviisa power plant, drying is not necessary as the fuel is transferred directly from the

36

transport cask to the final disposal canister. The fuel handling cell is presented in Figure 9-1.

Figure 9-1. Fuel handling cell at the encapsulation plant (Kukkola 2012). As a consequence of fuel damage in the fuel handling cell, a part of the fission products may escape from the fuel to the fuel handling cell. The fraction of released activity de-pends on factors such as speciation of a nuclide, fuel type, burnup of the fuel, external conditions (dry/wet) and the extent of the fuel damage. In dry conditions, only fission products present in gaseous or solid particle form are released. Submerged in water, solid fission products, such as Cs-137, I-129 and Cl-36, are released to a significantly greater extent, since water dissolves them from the surface of the fuel matrix. At the encapsulation plant, fuel is handled submerged in water only in the transport cask and only when wet transport method is used. The fuel transported in dry conditions is not in contact with water at the encapsulation plant. The release of solid fission products requires that water enters through the cladding into the fuel rod. This is possible only in case of significant fuel damage. Such significant fuel damage may take place, when the fuel assembly is lifted from a water-filled transport cask or during a wet transport. The probability for a damage during transport is approximately 2·10-7 per single fuel rod (Björkman & Kuusela 2012), meaning somewhat less than one fuel rod to be expected in average to be damage in transports from the Olkiluoto power plant during the entire operation of the encapsulation plant.

Docking station for the transfer cask

Docking station for the disposal canister

Fuel handling machine

Drying chamber

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In case of significant fuel damage occurs when the fuel assembly is lifted from a water-filled transport cask or during a wet transport, the space inside the cladding is filled with water only partially. In case of smaller defects (such as hairline fractures) occur during the lifting or transport the gases within the cladding prevent water from entering the cladding. If the damage takes place during the transport, part of the fission products is released to the water in the transport cask, and returned to the interim storage of spent fuel. As the fuel assembly is lifted from the wet transport cask to the fuel handling cell, the loose water is let to drip back to the transport cask. If the fuel damage is extensive, some of the water penetrated inside the cladding may drip back to the transport cask to be returned to the interim storage of spent fuel. In this case, though, it is possible that a small amount of water percolates to the floor of the fuel handling cell while the fuel assembly is transferred to the drying chamber. In case of a minor defect, water does not enter inside the cladding or remains there during the entire handling process. The trans-fer of the fuel assembly from the transport cask to the drying chamber takes typically a few minutes, meaning a limited release of fission products. It is anticipated that the moisture within the cladding is not completely removed in the drying chamber (Suikki et al. 2007). Is it not expected that a significant amount of radi-oactivity is released from the fuel in the drying chamber or during the transfer of fuel from the drying chamber to the final disposal canister. The release of solid fission products with water to the fuel handling cell is limited, since

A major defect should develop in the cladding during lifting of the fuel from the transport cask or during the transport to enable water to enter into the gas gap. The amount of fuel defect during the transport in minor.

Only a small fraction of water in the gas gap may end up in the fuel handling cell.

During the lift the fuel assembly is in contact with water for a short time only, meaning the dissolved activity be relatively low.

Based on this argumentation it is reasonable to assume that the release of solid fission products in the encapsulation process is unlikely, and on the other hand, the fraction of released fission products is small. Therefore all the fuel defects are assumed to take place in dry conditions. A significant fuel damage event during the lifting of a fuel as-sembly from the transport cask is considered an accident, and it is not discussed as a part of normal operation.

9.1.2 Frequency of fuel defects

The probability or frequency of a fuel damage in the encapsulation process can be esti-mated based on the operational experience of the power plants. Based on observed fuel damages at the Loviisa and Olkiluoto NPPs, Kukkola and Eurajoki (2009) have prelim-inarily estimated that not more than 1/1000 of the fuel assembly transfers (lifting a fuel assembly from one position to another) at the encapsulation plant results in fuel dam-

38

age. The conservative assumption behind that assessment was that all the fuel damages at power plant are caused by mishandling. The probability of mishandling has also been assessed in the probabilistic risk analysis (PRA) of the encapsulation plant (Björkman & Kuusela 2012). According to the PRA, around 21,000 fuel transfers have taken place at OL1&2 plant units during the latest 15 years. Three times a fuel assembly was lowered into a wrong position, 5 times it has swayed in water, and twice collided with the structures, resulting in 10 mishandlings in total. None of these led to damage for the fuel cladding. With a beta distribution this gives an expectation value for the frequency of a mishandlings equal to 1/2000 trans-fers. In the risk analysis it is assumed that mishandlings results in damage in one fuel rod. Based on the estimates presented above the damage frequency selected in the risk analysis is 1/1000 transfers. In addition to that, the risk analysis assumes that the dam-age frequency in the drying chamber is of the same order of magnitude as the transfers. In reality only a small fraction of the fuel damages developed at the power plant units are results of mishandlings. According to IAEA (2010) this figure for PWRs in 1994-2006 is approximately 0.4%. On the other hand, the damage mechanism is not always known, and therefore the figure may be somewhat higher. At the BWRs no fuel damages caused by mishandling were observed during the same years. According to IAEA (2010), the average damage frequency during the operation of the fuel at the PWRs around the world in 1994-2006 was 13.8/1000 assemblies removed from the re-actor. Assuming only 0.4% to be consequences of mishandlings results in a damage frequency due to mishandlings equal to 55/106 per fuel rods removed from the reactor. Assuming further three fuel transfer at all the plants prior to the removal from the reac-tor (including the transfer to the reactor and two transfers in the reactor) results in dam-age frequency equal to 18/106 transfers. Damages at the removal from the reactor or thereafter are not taken into account in the damage frequency. Assuming that the fuel handling at the encapsulation plant to some extent corresponds to the handling of the fuel at the nuclear power plants, this estimate is applicable also for the encapsulation plant. In dry conditions, water does not damp the possible colli-sions, but on the other hand, the work rate is slower than during the annual outages of the power plants. In addition, the risk analysis comprises different accidents with larger damages than that of one fuel rod. These events are not separately considered in the activity inventory. The nature of each accident is distinctive and it is difficult and not reasonable to assess their effect on the activity inventory in advance. Based on the estimates above, the conservative value for the damage frequency is 1/1000 transferred fuel assemblies, and the realistic one 18/106 transferred fuel assem-blies. The fuel transported in dry conditions is transferred in the fuel handling cell only once, but the fuel transported wet is transferred twice, and may also be damaged in the drying chamber. Hence the damage frequency for fuel transported wet is three times that of fuel transported dry. The estimates of damaged fuel rods in the realistic and con-servative cases are presented in Table 9-1.

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Table 9-1. Estimated amount of fuel defects in conservative (cons.) and realistic (real.) calculation cases.

Fuel type

Total BWR (OL1-2)

VVER (LO1-2)

EPR (OL3)

Number of disposal canisters (Table 2-1)

1400 750 2350 4500

Number of fuel assemblies per canister (Posiva 2013)

12 12 4

Number of fuel assemblies handled

16800 9000 9400 35200

Cons.

Probability of fuel damage per trans-ferred assembly

3/1000 1/1000 3/1000

Total amount of fuel damages

50 9 28 87

Real.

Probability of fuel damages per handled

assembly 55/106 18/106 55/106

Total amount of fuel damages

1 0 1 2

9.1.3 Fuel defects at the reactor units

In addition to the fuel rods possibly damaged at the encapsulation plant, also fuel rods damaged already at the power plant units are handled at the encapsulation plant. The majority of the activity in the gas gap of these rods has escaped already before being brought to the encapsulation plant. Until year 2014 at OL1-2 power plant units 40 fuel assemblies (Posiva 2014c) and at LO1-2 power plant units 33 fuel assemblies have damaged. Of the latter 33 assemblies, 14 have been returned to the Soviet Union/Russia (Posiva 2014d). The number of leaked fuel rods is probably somewhat higher than the number of leaked fuel assemblies. In addition to that Nieminen (2007) has estimated that during the operation of the power plant units there will be the following number of fuel damages:

OL1-2 plant units: 71 fuel rods OL3 plant unit: 18 fuel rods LO1-2 plant units: 32 fuel rods

In the estimate for Loviisa fuel, the fuel assemblies returned to the Soviet Union/Russia have been deducted. In addition, the large fuel damage occurred at LO2 plant unit is removed from the damage frequency, but added to the final estimate. At the Olkiluoto power plant, the leaking fuel rods have been removed from the fuel assemblies, and packed water-tightly. According to the present plans (Sorjonen 2015) the leaking rods will be re-packed in gas-filled packages, of which a new fuel assembly is assembled. This assembly contains fewer rods than the normal assemblies, but the external dimensions correspond to those of a normal assembly. At the encapsulation plant the assembly consisting of leaking rods is handled in a normal way, i.e. the exter-

40

nal surfaces are dried in the drying chamber before the transfer into the final disposal canister. At the Loviisa power plant the leaking fuel rods are in their original positions in the fuel assemblies. According to the present plans (Sorjonen 2015) the assemblies containing leaking rods will be packed hermetically before the drying. The package will be de-signed to enable removal of the water therein as well as the drying of the fuel assembly therein in such a way that the packed assembly can be loaded into a final disposal canis-ter. In neither case activity is assumed to be released from the leaking fuel rods. The pack-age containing a leaking rod may damage at the lifting in the same way as a normal fuel assembly, resulting in lost integrity of the package. Since a major part of the activity in the gas gap has already escaped at the power plant or in the interim storage for spent fuel, the activity release at the encapsulation plant should be lower than that resulting from a damage of an intact fuel rod at the encapsulation plant.

9.1.4 Release fractions of fission products

The instant release fraction (IRF) describes the proportion of activity that is quickly released from the fuel in the final disposal conditions. The conservative release fractions for the fuel submerged in water are taken from Posiva (2013). In those cases, in which the realistic value is lower than the conservative value, the realistic release fractions have been estimated based on SKB (2010a). In dry conditions the release fractions of solid substances are significantly lower than under water, and a minor release of solid substances is anticipated. Rossi et al. (2014) have estimated that cesium is released in form of fuel particles in dry conditions, and the fraction of the activity in particulate form represents 0.001% - 0.003% of the total cesium activity within the fuel rod. However, if the gas gap fraction exceeds 1% of the activity in the rod, the fraction in particulate form is to be corrected upwards. Applying the conservative IRF value for cesium (5%) gives the fraction in particulate form and, hence, the release fraction of 0.015%. With the same approach the release fraction for strontium becomes 0.003%. The release of iodine depends on its speciation. In line with Rossi et al. (2014) iodine is assumed to be present as three species: aerosol 95 % (CsI), elementary form 4.85 % (I2), and organic form 0.15 %. Organic iodine is released as a gas, and does probably not end up in operational waste, but is released to the atmosphere. Other iodine species can be released in particulate form only and the release fraction of iodine in particulate form can be assumed equal to that of cesium. Realistic release fractions can be calculated applying the average fraction given by Ros-si et al. (2014) of 0.002% being in particulate form and realistic release fractions under water. For elements other than cesium, strontium and iodine, no data for the dry condi-tions is available. In this analysis it is assumed that the release fractions in dry condi-tions can be calculated similarly as for cesium.

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The radionuclides released in gaseous form (H-3 and Kr-85) are released through the ventilation to the environment, and do not end up in the operational waste. Regarding C-14, it is conservatively assumed that all the C-14 released from the fuel is released in a solid particulate form, in a similar as with cesium, and, hence ends up in the opera-tional waste, although it is possible that in reality a significant part may be released as gaseous form. The estimated release fractions are collected to Table 9-2. Table 9-2. Release fractions under water (Posiva 2013) and in dry conditions.

Release fractions under

water (%) Release fractions in dry condi-

tions (%) Nuclide Realistic Conservative Realistic Conservative C-14 2.2 2.2 0.004 0.007 Cl-36 7.4 7.4 0.015 0.022 Se-79 0.4 0.4 0.002 0.003 Sr-90 0.5 1.0 0.002 0.003 Mo-93 0.1 0.1 0.002 0.003 Tc-99 0.2 1.0 0.002 0.003 Pd-107 0.2 1.0 0.002 0.003 Sn-126 0.01 0.01 0.002 0.003 I-129 2.5 5.0 0.005 0.015 Cs-135 2.5 5.0 0.005 0.015 Cs-137 2.5 5.0 0.005 0.015

9.1.5 Activity inventory of spent fuel

The activity present in the fuel at the moment of encapsulation depends on the fuel type, burnup and cooling time of the fuel assembly. The data for the separate power plant units and fuel types are presented in Table 9-3. As the exact time schedule of the encap-sulation is not known, it is assumed for the inventory of the operational waste that the encapsulation of fuel from a particular plant unit takes approximately the same time as the operation of that plant unit, and that the encapsulation starts 40 years after the com-missioning of the power plant unit. The inventory of the operational waste at the mo-ment of the repository closure is not affected by the encapsulation rate.

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Table 9-3. Information related to reactor units and fuel types. Operational times for reactor units are mainly estimates and they are rounded to the nearest five years. BWR fuel type refers to ATRIUM 10x10-9Q which is the same as used in the activity invento-ry calculations.

Reactor unit OL1-2 LO1-2 OL3 Fuel type BWR VVER PWR Minimum cooling time requirement (a) (Raiko 2005)

20 20 20

Average realistic minimum cooling time (a) (Raiko 2005)

44 32 57

Average assembly burnup (MWd/kgU) (Posiva 2013)

38-39 39-40 46-47

Rods per fuel assembly (-) (Anttila 2005a)

91 126 265

Masses of fuel per assembly (kg) (Anttila 2005a)

180 120 530

Mass of uranium in a rod (kg) 1.98 0.95 2.00 (Estimated) start of operation 1980 1990* 2020 Estimated end of operation 2040 2030 2080 Estimated start of encapsulation 2020 2020 2060 Estimated end of encapsulation 2080 2060 2120 Closure of the LILW repository 2120 *The spent fuel from LO1-2 units was returned to the Soviet Union / Russia until 1996, which has been taken into account by delaying the start of operation. The fuel loaded into reactors after around 1990 will be disposed of by Posiva.

The activities of selected nuclides in the fuel after a cooling time of 30 years are pre-sented in Table 9-4. The average burnup has been assumed to be 50 MWd/kgU, being a slightly higher than the average burnup in the table above. The burnup of single fuel elements damaged during the encapsulation may even exceed the assumed average burnup value. On the other hand, the large amount of assumed damages in the conserva-tive estimate averages the differences between single rods, and in the realistic estimate the properties of average fuel assembly are used.

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Table 9-4. Activity inventories in spent fuel at 30 years of cooling with the burnup of 50 MWd/kgU (Anttila 2005b). The activity concentration of Mo-93 in the LO1-2 fuel is missing, but OL3 fuel activity concentration is used instead. Similarly, the activity con-centration of Cl-36 for OL3 fuel is missing, but OL1-2 fuel activity concentration is used instead.

OL1-2 LO1-2 OL3 Nuclide GBq/tU GBq/Rod GBq/tU GBq/ Rod GBq/tU GBq/ Rod C-14 3.76E+01 7.44E-02 1.55E+02 1.47E-01 2.49E+01 4.99E-02 Cl-36 1.40E+00 2.78E-03 1.39E+00 1.32E-03 1.40E+00 2.80E-03 Se-79 4.01E+00 7.94E-03 3.91E+00 3.71E-03 3.96E+00 7.92E-03 Sr-90 1.98E+06 3.92E+03 1.85E+06 1.75E+03 1.89E+06 3.79E+03 Mo-93 1.49E+01 2.96E-02 1.74E+01 1.65E-02 1.74E+01 3.47E-02 Tc-99 7.34E+02 1.45E+00 7.10E+02 6.74E-01 7.17E+02 1.43E+00 Pd-107 6.22E+00 1.23E-02 7.29E+00 6.92E-03 6.93E+00 1.39E-02 Sn-126 2.89E+01 5.73E-02 3.12E+01 2.96E-02 3.04E+01 6.09E-02 I-129 1.46E+00 2.89E-03 1.56E+00 1.48E-03 1.53E+00 3.07E-03 Cs-135 2.58E+01 5.11E-02 2.41E+01 2.29E-02 2.74E+01 5.48E-02 Cs-137 2.91E+06 5.77E+03 2.93E+06 2.79E+03 2.93E+06 5.85E+03 Total 4.89E+06 9.69E+03 4.78E+06 4.54E+03 4.82E+06 9.64E+03

9.1.6 Total activity of fission products at the LILW hall closure

To estimate the activity of the fission products at the time of closure of the LILW repos-itory, first the average annual activity released from the fuel is calculated. It is assumed that the encapsulation rate and the fuel defect rate remain constant during the operation-al time of the encapsulation plant. The released activity is estimated with the formula (9-1)

where is the release rate (Bq/a) of nuclide from the fuel of plant unit , is the activity inventory in one fuel rod (Bq), is the release fraction of the nuclide in dry conditions (-). The defect rate of the rods (1/a) in dry conditions is calculated with the formula

(9-2)

where is the estimated amount of defects in fuel type during the operational time of the encapsulation plant (-) and (a) the duration of the encapsulation of the given fuel type, here assumed equal to the operational time of the corresponding power plant unit. The activity accumulation is described by the differential equation

(9-3)

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where is the decay constant of the nuclide (1/a). The solution with the initial condi-tion is

(9-4)

The activity at the closure moment of the LILW repository is calculated taking the radi-oactive decay after the encapsulation of the given fuel type into account with the formu-la

(9-5)

where is the duration of the encapsulation of the given fuel type (a), the time between the end of encapsulation of the given fuel type and the closure of the final re-pository (a), and the cooling time (a) of the fuel prior to the encapsulation. To sim-plify the calculation it is assumed that that the cooling time for all the fuel is 30 years, based on average cooling times in Table 9-3. The cooling time does not affect the activi-ty inventory at the time of the closure. The closure is assumed to take place in 2120. The resulting activity inventories in conservative and realistic estimates are presented in Table 9-5 and Table 9-6, respectively. Table 9-5. Conservative estimate of the activity of fission products in the waste to be disposed in the LILW repository at the time of closure in 2120.

Total activity (Bq) Nuclide OL1-2 LO1-2 OL3 Total C-14 2.58E+05 9.18E+04 4.86E+04 3.98E+05 Cl-36 3.05E+04 2.61E+03 8.64E+03 4.18E+04 Se-79 1.19E+04 1.00E+03 3.32E+03 1.62E+04 Sr-90 9.34E+08 7.18E+07 6.62E+08 1.67E+09 Mo-93 4.37E+04 4.39E+03 1.45E+04 6.26E+04 Tc-99 2.18E+06 1.82E+05 6.02E+05 2.96E+06 Pd-107 1.85E+04 1.87E+03 5.82E+03 2.62E+04 Sn-126 8.59E+04 7.99E+03 2.56E+04 1.19E+05 I-129 2.17E+04 2.00E+03 6.44E+03 3.01E+04 Cs-135 3.83E+05 3.09E+04 1.15E+05 5.29E+05 Cs-137 7.41E+09 6.17E+08 5.29E+09 1.33E+10 Total 8.35E+09 6.89E+08 5.95E+09 1.50E+10

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Table 9-6. Realistic estimate of the activity of fission products in the waste to be dis-posed in the LILW repository at the time of closure in 2120. In the realistic estimate, no fuel defects is assumed for LO1-2 fuel and thus no fission products are released from the fuel.

Total activity (Bq) Nuclide OL1-2 LO1-2 OL3 Total C-14 2.95E+03 0.00E+00 1.98E+03 4.93E+03 Cl-36 4.16E+02 0.00E+00 4.21E+02 8.37E+02 Se-79 1.59E+02 0.00E+00 1.58E+02 3.17E+02 Sr-90 1.25E+07 0.00E+00 3.15E+07 4.40E+07 Mo-93 5.83E+02 0.00E+00 6.90E+02 1.27E+03 Tc-99 2.90E+04 0.00E+00 2.87E+04 5.77E+04 Pd-107 2.46E+02 0.00E+00 2.77E+02 5.24E+02 Sn-126 1.15E+03 0.00E+00 1.22E+03 2.36E+03 I-129 1.45E+02 0.00E+00 1.53E+02 2.98E+02 Cs-135 2.56E+03 0.00E+00 2.74E+03 5.30E+03 Cs-137 4.94E+07 0.00E+00 1.26E+08 1.75E+08 Total 6.19E+07 0.00E+00 1.58E+08 2.19E+08

9.2 CRUD

9.2.1 Handling of crud at the encapsulation plant

In addition to fission products, there are activated corrosion product deposits, also known as crud, on the surfaces of the fuel assemblies, which may spall off during the encapsulation process. The fraction of crud ending up in the operational waste of the encapsulation plant is, however, believed to be minor due to the following reasons:

If there is crud that may spall off easily, it probably spalls off already during the loading and unloading of the transport cask. In these cases, the crud will remain in the cask and return to the spent fuel storage, to be further treated.

During the drying process, the conditions (temperature, pressure) are relatively stable and therefore no significant spalling of crud is expected. However, if some crud spalls off during drying, it will end up in the operational waste.

If crud spalls off after the drying process, when the fuel assembly is being load-ed into the canister, most of it will be swept or vacuum-cleaned and disposed of into the final disposal canisters (Kukkola 2012).

As there is no operational experience on the handling the spent fuel from the Finnish nuclear power plants in conditions similar to those in the encapsulation plant, the activi-ty of crud ending up in the operational waste is estimated with available data. In addi-tion to the activated corrosion products, the crud may include also some actinides, for which a coarse and conservative activity inventory has been estimated. There are several uncertainties related to the crud activity inventory ending up in the operational or decommissioning waste of the encapsulation plant, such as:

Crud amount at the surfaces of the fuel assemblies

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• Specific activity of different nuclides in the crud • Spalling fraction of the crud during the encapsulation process • The fraction disposed of in the empty positions of spent fuel canisters vs. that

disposed of in the LILW • Power plant unit and fuel type related differences in all the factors mentioned

above.

As the available data on the existing fuel is sparse, and as operation experience of all the fuel (e.g. EPR fuel) to be encapsulated does not even exist at the moment, the assess-ment task is approached from several points of view, and the inventory is produced as a synthesis of the available information. This approach is selected to create confidence in the order of magnitude of the resulting inventory.

9.2.2 Activity inventory in crud

Radioactivity is induced in isotopes of corrosion products, when they are deposited at the surfaces of fuel and core structures, and exposed to neutron flux. The induced activi-ty is proportional to the amount of the exposed corrosion products as well as the neutron flux, activation cross section and exposure time. The activity inventory of each activa-tion product in the corrosion deposit can be estimated with the following equation:

(9-6)

where

= total activity (Bq) = decay constant (1/s) = irradiation time (s) = decay time after the irradiation (s) = mass fraction of the target element (-) = abundance of the target isotope in natural element (-) = mass of the crud (g) = Avogadro constant (6.022·1023 1/mol)

= atom weight of the target isotope (g/mol) = activation cross section (cm2) = average neutron flux density (1/cm2s).

9.2.3 Amount and elemental composition of crud

BWR The estimated total amount of the crud from the BWR fuel assemblies is based on the assumed crud deposit of 20 g/m2 at the surfaces of the fuel assemblies, as well as an estimate on the spalling crud, i.e. 0.8 % of the total amount (Salo 1983). The TURVA-2012 safety case (2013) makes reference also to newer corrosion product measurements on the TVO fuel, and in light of these as well as some further information (Ranta-aho

47

2011), the estimated amount of corrosion products 20 g/m2 is still an applicable value. On the amount of spalling crud there is no newer data available. According to Salo (1983) the amount of crud spalling off of a BWR fuel assembly with 64 fuel rods amounts to 1.58 g. Considering the uranium mass in the element, 0.178 tU, gives the amount of spalling crud per uranium mass equal to 8.9 g crud/tU. Assuming the BWR fuel to amount to 2950 t, this results in a mass of crud equal to 26.3 kg. How-ever, in later fuel assemblies at Olkiluoto 1&2, the cladding diameter is somewhat smaller resulting in a higher area/mass ratio. Hence, the total mass of the spalling crud is assumed to be 30 kg. The specific activity of the nickel isotopes is estimated based on the available data of the elemental composition of the crud. Since Ni isotopes do not emit gamma rays, there are no measurement results available for a comparison of the calculated value. There-fore 60Co, for which there are measured values available, is assessed in the similar way as Ni isotopes. About the spent fuel assemblies from TVO, the following data is available (Salo 1983):

Ni content in the crud: 5 … 13 % Co content in the crud: 0.1 … 0.2 % Total amount of crud: 6.2 … 7.4 g/m2

According to more recent measurements by TVO, the nickel content has been 5-10 % (Ranta-aho 2011). Co-60 activity has been about 10 GBq/g. TURVA-2012 safety case (Posiva 2013) mentions 2-6% as a typical nickel content. The measured deposit data referred to above originate in fuel assemblies with burnup significantly lower than the current maximum. Whether the deposit grows linearly with burnup is unclear, but some kind of correlation is likely to exist. According to (Neeb 1997), the Ni content of BWR fuel rod deposits is 4.0 … 6.2 % and Co content less than 0.6 %. The area related mass of the deposits is approximately 0.2 mg/cm2 (2 g/m2), i.e. one tenth of the value assumed for TVO assemblies (20 g/m2) and less than one third of the measured value. Assuming the majority of the spalling crud to be deposited and disposed of in empty positions of the final disposal canister, the fraction remaining in the operational and decommissioning waste of the encapsulation plant can be assumed to be 1 - 10 %, i.e. 0.3 - 3 kg of the crud spalling off of the BWR fuel. Moreover, the nickel content in the crud is assumed to be 10%, i.e. 0.03 - 0.3 kg. Cobalt content is assumed to be 0.2 %. VVER and PWR In the visual examinations of the spent fuel from the Loviisa power plant, the cladding has been almost without deposits. The crud content of VVER and PWR fuels is so low that it does not have an impact on the source term, as shown in the TURVA-2012 safety assessment (Posiva 2013). On the other hand, SKB (2010b) have in their long-term safety case considered also the crud from PWR fuels.

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Regarding the spent fuel assemblies from Loviisa, the following data is available about the fuel assemblies with significant deposits observed in 1995:

• Nickel content in crud is 2.32 % - 4.08 % (Rosenberg et al. 1998) and 2.6 % - 5.0 % (Rosenberg et al. 1996).

• Cobalt activity in crud is 48 - 176 MBq/gcrud (Rosenberg et al. 1996). Cobalt content in crud is 0.02 % - 0.04 %.

It is to be noted that this data relates to fuel with a low burnup. According to (Neeb 1997), the NiO content in the typical compositions of PWR fuel rod deposits is 3 … 9 %, and the CoO content is less than 0.05 %. SKB's report (2010b) gives a PWR canister a Ni-59 inventory value of 3.18·108 Bq, whereas the correspond-ing figure for a BWR canister is 7.03·109 Bq, reflecting the lower crud amount in the PWR fuel. It is assumed that the fraction of loosened crud as well as that of ending up in the opera-tional waste for VVER and PWR fuels are equal to those for BWR fuel. If the amount of crud per uranium mass is one tenth of that for BWR fuel, the crud amount in the op-erational waste originating from the VVER and PWR fuel is 0.01 - 0.1 kg and 0.05 - 0.5 kg, respectively. The measurements on the Loviisa crud indicate somewhat lower nickel and cobalt contents than those selected above for BWR fuel, but same assump-tions are adopted for all the fuel types.

9.2.4 Activation rate

The integral in the Equation (9-6) above gives the activation rate per target nuclide. This can be approximated as a product of Westcott's neutron flux and the effective activation cross section. With typical reactor core data, the latter in turn can be approximated based on the thermal cross section and the resonance integral for each reaction.

(9-7)

(9-8) where

= Westscott's neutron flux (Loviisa core average 2·1013 1/cm2s) = effective neutron cross section for the activation reaction (cm2)

= thermal neutron cross section at 0.0253 eV for the activation reaction (cm2) = resonance integral (cm2)

The different reactor types are not assumed to differ from each other to such an extent that it should be taken into account in the reaction rate estimates. According to a rough estimate based on the available data of the core parameters the neutron fluxes are ex-pected to vary not more than 20 %, whereas the uncertainties e.g. in the crud amounts

49

and compositions are clearly greater. The data for the activation reactions of interest are presented in Table 9-7. Table 9-7. Activation rates and the parameters used for calculating them for the most important nuclides Ni-59, Ni-63 and Co-60. Ni-58 (n,g) Ni-59 Ni-62 (n,g) Ni-63 Co-59 (n,g) Co-60 Thermal cross sec-tion (b)

4.67 14.37 37.63

Resonance integral (b)

2.20 6.89 75.48

Effective cross sec-tion (b)

5.78 17.81 75.37

Reaction rate per target nuclide (1/s) (w = 2·1013 1/cm2s)

1.16E-10 3.56E-10 1.51E-09

9.2.5 Residence time in reactor

The fuel assemblies are typically irradiated in the reactor for 3 … 4 years, which is the maximum residence time for the crud to deposit at the surfaces of the fuel assemblies. However, there are indications of the residence times of crud deposits being significant-ly shorter. The residence times can be assessed based on the ratios of the activation products and the target nuclides of the corresponding activation reactions. Neeb (1997) mentions that in some cases residence times have been even shorter than 1 day. Barton et al. (2001) have observed apparent mean residence times for some gamma-emitting nuclides in the core deposits in Sizewell B PWR, and the range from less than 1 day up to 67 days, with the vast majority being less than one month. According to Riess et al. (2006) the residence times at least for cobalt, based on 60Co/Co ratio are longer for BWRs than for PWRs. The activity level mentioned for BWRs, 1 TBq/g, corresponds roughly to a residence time of half a year. The residence times can be calculated also from the measured values from Loviisa and Olkiluoto NPPs. For 60Co they are a few months in the Loviisa case (specific activity 3·1011 Bq/gCo, Rosenberg et al. 1996), and about a year for the Olkiluoto case (specific activity 2·1012 Bq/gCo, Salo 1983). As no measured data on the radioactive nickel isotopes in the crud on fuel assembly surfaces are available, and as the residence time obviously depends on the material, the data on other elements cannot be applied as such. However, even though the typical residence times may be significantly shorter than the total irradiation time, in this activity inventory assessment it is assumed that the resi-dence time equals to 3 years. Even though this value is considered clearly conservative, especially for PWRs, it is not completely beyond the reasonable range, when the related uncertainties are taken into account.

50

9.2.6 Activity concentration in crud immediately after activation

Based on the parameter selections described in the previous subsections, as well as the related natural constants, the activity concentrations of Ni and Co isotopes in the irradi-ated crud deposits immediately after the residence time in the core equal to:

Ni-59: 2.28·106 Bq/g Ni-63: 2.56·108 Bq/g Co-60: 1.00·1010 Bq/g

The calculated Co-60 activity is roughly equal to the highest measured Co-60 activity reported by Salo (1983) and Ranta-aho (2011) for Olkiluoto fuel, but clearly higher than the values reported by Rosenberg et al. (1996) for the Loviisa fuel. The differences can be explained by the lower cobalt content in the Loviisa crud as well as the fact that the measured value originates in an exceptional crud accumulation, in case of which the residence time in reactor should have been short.

9.2.7 Activity inventory of crud

Since the activity inventory of many crud nuclides that are significant for the long-term safety has not been measured, their inventories are produced here utilizing the activity inventory presented by SKB (SKB 2010b). Ni-59 is selected as a reference nuclide, upon which the activities of the other nuclides are calculated, assuming that the activity ratios similar as in the BWR and PWR inven-tories presented by SKB. The activity of Ni-59 is estimated above. The inventory of Ni-59 due to crud from the BWR fuel is (0.3 - 3) kg · 2.28·109 Bq/kg = 6.84·108 - 6.84·109 Bq. Similarly, for VVER fuel the inventory is 2.28·107 - 2.28·108 and for PWR fuel 1.14·108 - 1.14·109 Bq. This may be compared with the Ni-59 inven-tory of one BWR canister presented in TURVA-2012 safety case report (Posiva 2013), i.e. 7.03·109 Bq. As there are 1400 BWR canisters, and if the fraction of crud ending up in the operational waste corresponded to 0.008% - 0.08% of the crud entering the canis-ter, the activity in the operational waste would be 7.87·108 - 7.87·109 Bq, i.e. close to the value presented above. Table 9-8 presents the average accumulation rates of crud activity, assuming the encap-sulation time of each fuel type equal to the operational time of the corresponding plant unit (Table 9-3), and the cooling time equal to 30 years. Table 9-9 presents the corre-sponding activities in year 2120, calculated in a similar way with fission product activi-ty. The values in the tables are produced scaling the values presented by SKB (2010b) in such a way that the specific activity of Ni-59 in crud corresponds to the value presented above and the mass of crud is scaled to the masses ending up in the operational waste. The scaling is applied to those values presented by SKB (2010b), which assume a burnup of 47.8 MWd/KgHM and 44.8 MWd/KgHM for BWR and PWR fuels, respec-tively. These burnup values are used only for scaling, since the crud activities calculated above are based on residence times in reactor and not fixed to any particular burnup

51

values. There is one exception from this scaling procedure: Ag-108m, whose half-life has been a subject to change during past years. As it is possible that the activity in the SKB report (2010b) is based on a half-life value of 127 a, the activity is corrected by calculating first the initial activity assuming a half-life of 127 a, and thereafter applying a value of 438 a (Posiva 2013) for the half-life. The values for Co-60 are calculated based on the activity concentration presented above. Values in Table 9-8 and in Table 9-9 correspond to the conservative crud activity. The realistic inventory is assumed to be one tenth of that. Table 9-8. Conservatively estimated average crud activity inventories (Bq/a) at 30 years of cooling.

Nuclide OL1-2 LO1-2 OL3 Co-60 9.67E+09 5.03E+08 1.60E+09Ni-59 1.14E+08 5.94E+06 1.89E+07Ni-63 1.31E+10 6.05E+08 1.92E+09Zr-93 5.11E+04 2.45E+02 7.78E+02Nb-93m 1.62E+09 2.88E+06 9.12E+06Nb-94 4.61E+06 7.74E+03 2.46E+04Mo-93 1.27E+05 1.27E+03 4.02E+03Tc-99 1.91E+04 2.17E+02 6.88E+02Ag-108m 1.25E+07 1.42E+06 4.50E+06Sn-121m 2.97E+06 7.34E+03 2.34E+04U-234 1.46E+02 2.67E+02 8.48E+02U-235 1.43E+00 3.05E+00 9.68E+00U-236 3.54E+01 6.00E+01 1.90E+02U-238 3.21E+01 5.50E+01 1.74E+02Np-237 5.72E+01 9.05E+01 2.88E+02Pu-238 5.91E+05 7.91E+05 2.50E+06Pu-239 3.33E+04 6.49E+04 2.06E+05Pu-240 5.49E+04 1.02E+05 3.24E+05Pu-241 3.81E+06 6.89E+06 2.18E+07Pu-242 3.73E+02 5.48E+02 1.74E+03Am-241 5.21E+05 8.37E+05 2.66E+06Am-242m 2.68E+03 2.76E+03 8.76E+03Am-243 6.35E+03 8.10E+03 2.58E+04Cm-243 1.91E+03 2.57E+03 8.14E+03Cm-244 4.20E+05 4.88E+05 1.55E+06Cm-245 1.66E+02 1.89E+02 6.00E+02Cm-246 5.01E+01 4.92E+01 1.56E+02Total 2.45E+10 1.13E+09 3.58E+09

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Table 9-9. Conservative crud activities (Bq) at the repository closure in 2120. Nuclide OL1-2 LO1-2 OL3 Total Co-60 1.02E+08 1.42E+06 1.63E+09 1.73E+09 Ni-59 6.84E+09 2.38E+08 5.66E+08 7.64E+09 Ni-63 4.56E+11 1.39E+10 4.40E+10 5.14E+11 Zr-93 3.07E+06 9.80E+03 2.33E+04 3.10E+06 Nb-93m 4.03E+09 4.14E+06 6.37E+07 4.10E+09 Nb-94 2.76E+08 3.09E+05 7.36E+05 2.77E+08 Mo-93 7.49E+06 5.00E+04 1.20E+05 7.66E+06 Tc-99 1.15E+06 8.67E+03 2.07E+04 1.18E+06 Ag-108m 4.86E+08 3.67E+07 1.09E+08 6.32E+08 Sn-121m 6.66E+07 1.08E+05 4.32E+05 6.71E+07 U-234 1.58E+04 1.74E+04 4.14E+04 7.46E+04 U-235 8.57E+01 1.22E+02 2.90E+02 4.98E+02 U-236 2.12E+03 2.40E+03 5.72E+03 1.02E+04 U-238 1.93E+03 2.20E+03 5.24E+03 9.37E+03 Np-237 4.04E+03 4.18E+03 9.43E+03 1.77E+04 Pu-238 1.90E+07 1.69E+07 5.54E+07 9.13E+07 Pu-239 1.99E+06 2.59E+06 6.17E+06 1.08E+07 Pu-240 3.27E+06 4.06E+06 9.71E+06 1.70E+07 Pu-241 6.58E+06 6.64E+06 1.31E+08 1.44E+08 Pu-242 2.24E+04 2.19E+04 5.22E+04 9.65E+04 Am-241 3.07E+07 3.56E+07 8.95E+07 1.56E+08 Am-242m 1.09E+05 7.47E+04 2.17E+05 4.01E+05 Am-243 3.78E+05 3.22E+05 7.69E+05 1.47E+06 Cm-243 1.79E+04 1.53E+04 1.01E+05 1.34E+05 Cm-244 1.45E+06 1.00E+06 1.24E+07 1.49E+07 Cm-245 9.88E+03 7.51E+03 1.79E+04 3.53E+04 Cm-246 2.97E+03 1.95E+03 4.66E+03 9.58E+03 Total 4.68E+11 1.42E+10 4.67E+10 5.29E+11

9.3 Summary of activity inventory in encapsulation waste

A combination of the activity inventories of fission products and crud in sections 9.1 and 9.2 results in the total inventory of the operational waste, the conservative estimate of which is presented in Table 9-10 and the realistic estimate in Table 9-11. The short-lived nuclides are not presented in the tables. The total inventory presented here ends up almost exclusively into the waste to be em-placed in the concrete basin to be located in the final repository. In the preliminary safe-ty assessment of the LILW repository, the amount of activity deposited outside the con-crete basin was assumed to be 1 % of the total activity (Nummi et al. 2012). In their review of the safety assessment, Keith-Roach et al. (2015) considered the assumption

53

overly conservative. On the other hand, reducing the assumed amount of activity that could be left outside the concrete basin cannot be easily justified due to uncertainties related to the overall activity inventory and the waste generation. The proportion of the acceptable amount of activity outside the concrete basin should be determined as a part of the long-term safety assessment for the entire disposal facility (including the spent fuel repository). This proportion together with the activity distribution in the waste, de-termines the dimensioning of the basin.

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Table 9-10. Conservative estimated of total activity (Bq) of encapsulation waste to be disposed of in the LILW repository at the time of the closure in 2120.

Nuclide OL1-2 LO1-2 OL3 Total C-14 2.58E+05 9.18E+04 4.86E+04 3.98E+05 Cl-36 3.05E+04 2.61E+03 8.64E+03 4.18E+04 Co-60 1.02E+08 1.42E+06 1.63E+09 1.73E+09 Ni-59 6.84E+09 2.38E+08 5.66E+08 7.64E+09 Ni-63 4.56E+11 1.39E+10 4.40E+10 5.14E+11 Se-79 1.19E+04 1.00E+03 3.32E+03 1.62E+04 Sr-90 9.34E+08 7.18E+07 6.62E+08 1.67E+09 Zr-93 3.07E+06 9.80E+03 2.33E+04 3.10E+06 Nb-93m 4.03E+09 4.14E+06 6.37E+07 4.10E+09 Nb-94 2.76E+08 3.09E+05 7.36E+05 2.77E+08 Mo-93 7.53E+06 5.44E+04 1.34E+05 7.72E+06 Tc-99 3.33E+06 1.91E+05 6.23E+05 4.14E+06 Pd-107 1.85E+04 1.87E+03 5.82E+03 2.62E+04 Ag-108m 5.27E+08 4.28E+07 1.09E+08 6.78E+08 Sn-121m 6.66E+07 1.08E+05 4.32E+05 6.71E+07 Sn-126 8.59E+04 7.99E+03 2.56E+04 1.19E+05 I-129 2.17E+04 2.00E+03 6.44E+03 3.01E+04 Cs-135 3.83E+05 3.09E+04 1.15E+05 5.29E+05 Cs-137 7.41E+09 6.17E+08 5.29E+09 1.33E+10 U-234 1.58E+04 1.74E+04 4.14E+04 7.46E+04 U-235 8.57E+01 1.22E+02 2.90E+02 4.98E+02 U-236 2.12E+03 2.40E+03 5.72E+03 1.02E+04 U-238 1.93E+03 2.20E+03 5.24E+03 9.37E+03 Np-237 4.04E+03 4.18E+03 9.43E+03 1.77E+04 Pu-238 1.90E+07 1.69E+07 5.54E+07 9.13E+07 Pu-239 1.99E+06 2.59E+06 6.17E+06 1.08E+07 Pu-240 3.27E+06 4.06E+06 9.71E+06 1.70E+07 Pu-241 6.58E+06 6.64E+06 1.31E+08 1.44E+08 Pu-242 2.24E+04 2.19E+04 5.22E+04 9.65E+04 Am-241 3.07E+07 3.56E+07 8.95E+07 1.56E+08 Am-242m 1.09E+05 7.47E+04 2.17E+05 4.01E+05 Am-243 3.78E+05 3.22E+05 7.69E+05 1.47E+06 Cm-243 1.79E+04 1.53E+04 1.01E+05 1.34E+05 Cm-244 1.45E+06 1.00E+06 1.24E+07 1.49E+07 Cm-245 9.88E+03 7.51E+03 1.79E+04 3.53E+04 Cm-246 2.97E+03 1.95E+03 4.66E+03 9.58E+03 Total 4.76E+11 1.49E+10 5.26E+10 5.44E+11

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Table 9-11. Realistic estimate of total activity (Bq) of encapsulation waste to be dis-posed of in the LILW repository at the time of the closure in 2120.

Nuclide OL1-2 LO1-2 OL3 Total C-14 2.95E+03 0.00E+00 1.98E+03 4.93E+03 Cl-36 4.16E+02 0.00E+00 4.21E+02 8.37E+02 Co-60 1.02E+07 1.42E+05 1.63E+08 1.73E+08 Ni-59 6.84E+08 2.38E+07 5.66E+07 7.64E+08 Ni-63 4.56E+10 1.39E+09 4.40E+09 5.14E+10 Se-79 1.59E+02 0.00E+00 1.58E+02 3.17E+02 Sr-90 1.25E+07 0.00E+00 3.15E+07 4.40E+07 Zr-93 3.07E+05 9.80E+02 2.33E+03 3.10E+05 Nb-93m 4.03E+08 4.14E+05 6.37E+06 4.10E+08 Nb-94 2.76E+07 3.09E+04 7.36E+04 2.77E+07 Mo-93 7.50E+05 5.00E+03 1.27E+04 7.67E+05 Tc-99 1.44E+05 8.67E+02 3.08E+04 1.76E+05 Pd-107 2.46E+02 0.00E+00 2.77E+02 5.24E+02 Ag-108m 5.27E+07 4.28E+06 1.09E+07 6.78E+07 Sn-121m 6.66E+06 1.08E+04 4.32E+04 6.71E+06 Sn-126 1.15E+03 0.00E+00 1.22E+03 2.36E+03 I-129 1.45E+02 0.00E+00 1.53E+02 2.98E+02 Cs-135 2.56E+03 0.00E+00 2.74E+03 5.30E+03 Cs-137 4.94E+07 0.00E+00 1.26E+08 1.75E+08 U-234 1.58E+03 1.74E+03 4.14E+03 7.46E+03 U-235 8.57E+00 1.22E+01 2.90E+01 4.98E+01 U-236 2.12E+02 2.40E+02 5.72E+02 1.02E+03 U-238 1.93E+02 2.20E+02 5.24E+02 9.37E+02 Np-237 4.04E+02 4.18E+02 9.43E+02 1.77E+03 Pu-238 1.90E+06 1.69E+06 5.54E+06 9.13E+06 Pu-239 1.99E+05 2.59E+05 6.17E+05 1.08E+06 Pu-240 3.27E+05 4.06E+05 9.71E+05 1.70E+06 Pu-241 6.58E+05 6.64E+05 1.31E+07 1.44E+07 Pu-242 2.24E+03 2.19E+03 5.22E+03 9.65E+03 Am-241 3.07E+06 3.56E+06 8.95E+06 1.56E+07 Am-242m 1.09E+04 7.47E+03 2.17E+04 4.01E+04 Am-243 3.78E+04 3.22E+04 7.69E+04 1.47E+05 Cm-243 1.79E+03 1.53E+03 1.01E+04 1.34E+04 Cm-244 1.45E+05 1.00E+05 1.24E+06 1.49E+06 Cm-245 9.88E+02 7.51E+02 1.79E+03 3.53E+03 Cm-246 2.97E+02 1.95E+02 4.66E+02 9.58E+02 Total 4.69E+10 1.43E+09 4.83E+09 5.31E+10

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9.4 Activity inventory in the dried liquid waste drums

Among other factors the activity inventory of the dried waste drums depend on the cool-ing time of the spent nuclear fuel to be encapsulated. The activity of Co-60 is especially affected by the cooling time, since its half-life is 5.3 years, and it is the main contributor to the external dose rate of the drum. The realistic average cooling times of the encapsulated fuel are greater than 30 years (cf. Table 9-3). Conservatively 30 years is here selected as the cooling time value for all the fuel types. It is assumed that the release of radioactivity from the fuel assemblies, such as spalling of crud, is not influenced by the cooling time of the fuel assembly prior to encapsulation. It is estimated that filling one drum takes about 9 years (see section 3.2). Hence, one drum receives the crud activity accumulated during nine years. The radioactive decay during the filling of the drum is neglected in the calculation, meaning the drum is filled up soon after a fuel damage or a crud spalling event. This influences most on the Co-60 activity, but the influence is relatively low compared to other uncertainties. Based on the fuel damage frequency and the filling time of a drum it is possible to esti-mate the number of fuel rods whose fission products are released into one drum. They are presented in Table 9-12. The realistic case considers a drum which receives the ac-tivity released by one fuel rod. This does not represent the average fission product activ-ity, but a maximum, due to the low total number of fuel damages in the realistic calcula-tion case. Table 9-12. The average number of rods that release fission products into a single drum.

OL1-2 LO1-2 OL3 Conservative esti-mate

6 2 4

Realistic estimate 1 0 0 In addition it is in this report assumed that maximum of two fuel types are encapsulated at the encapsulation plant in turns such that the activity released from these fuel types may accumulate into a single drum. In the first phase, OL1-2 and LO1-2 fuel, and later OL1-2 and OL3 fuel are encapsulated in turns. The highest activity of one single drum is a combination of waste accumulated in the encapsulation of OL1-2 and OL3. This is presented in Table 9-13.

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Table 9-13. Activity of a single dried liquid waste drum at 30 years of cooling, when the drum contains activity from OL1-3 reactor units. The half-lives are primarily from (Posiva 2013) and secondarily from (SKB 2010b).

Nuclide Half-life (a) Conservative estimate

OL1-3 (Bq) Realistic estimate

OL1-3 (Bq) C-14 5700 3.82E+04 2.98E+03 Cl-36 301 000 4.90E+03 4.16E+02 Co-60 5.27 9.42E+10 9.42E+09 Ni-59 76 000 1.11E+09 1.11E+08 Ni-63 101 1.27E+11 1.27E+10 Se-79 327 000 1.90E+03 1.59E+02 Sr-90 28.8 9.33E+08 7.84E+07 Zr-93 1 610 000 4.63E+05 4.63E+04 Nb-93m 16.1 1.46E+10 1.46E+09 Nb-94 20 300 4.16E+07 4.16E+06 Mo-93 4000 1.17E+06 1.17E+05 Tc-99 211 000 5.23E+05 4.66E+04 Pd-107 6 500 000 3.05E+03 2.46E+02 Ag-108m 438 1.07E+08 1.07E+07 Sn-121m 55.0 2.68E+07 2.68E+06 Sn-126 230 000 1.40E+04 1.15E+03 I-129 15 700 000 3.53E+03 1.45E+02 Cs-135 2 300 000 6.24E+04 2.56E+03 Cs-137 30.1 6.95E+09 2.88E+08 U-234 246 000 5.13E+03 5.13E+02 U-235 704 000 000 5.64E+01 5.64E+00 U-236 23 400 000 1.18E+03 1.18E+02 U-238 4 470 000 000 1.07E+03 1.07E+02 Np-237 2 140 000 1.81E+03 1.81E+02 Pu-238 87.7 1.66E+07 1.66E+06 Pu-239 24 100 1.23E+06 1.23E+05 Pu-240 6560 1.95E+06 1.95E+05 Pu-241 14.3 1.32E+08 1.32E+07 Pu-242 375 000 1.12E+04 1.12E+03 Am-241 432 1.67E+07 1.67E+06 Am-242m 141 6.35E+04 6.35E+03 Am-243 7370 1.73E+05 1.73E+04 Cm-243 29.1 5.38E+04 5.38E+03 Cm-244 18.1 1.07E+07 1.07E+06 Cm-245 8420 4.19E+03 4.19E+02 Cm-246 4710 1.15E+03 1.15E+02 Total 2.45E+11 2.41E+10

The dose rate of a dried waste drum has been assessed in the operational safety analysis of the encapsulation plant (Nummi 2015). The surface dose rate of a drum is 97 mSv/h using the conservative activity content, and 9.7 mSv/h using the realistic activity con-tent. At the distance of 1 meter the corresponding values are 3.6 and 0.36 mSv/h, re-spectively. The radiation shield of the drum shall be dimensioned to lower the surface dose rate down to below 2 mSv/h. For example a 10 cm thick radiation shield made of

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steel results in surface dose rate of 1.4 mSv/h in the conservative estimate and 0.14 mSv/h for the realistic estimate. 9.5 Maximum activity concentration of drainage water

The starting point for the quantitative determination of the maximum activity concentra-tion in the drainage wasters of the encapsulation plant is crud from OL1-2 fuel (Table 9-8) and fallen to the bottom of the drying chamber. This crud is assumed to remain at the bottom or in the fuel rack of the drying chamber to be released to the drainage dur-ing decontamination. In this estimate, a few decontamination campaigns of the drying chamber are assumed during the entire operation of the encapsulation plant, with the intervals of 35 years. During this time, the radioactive decay of Co-60 significantly re-duces its activity. In the calculation of the maximum activity concentration all the crud spalling off of the fuel is assumed to fall into the drying chamber. On the other hand, 90% of that is assumed to be removed and emplaced into spent fuel canisters, as de-scribed above. The drying chamber is assumed to be decontaminated with 50 liters of water. The calculated maximum concentration of drainage water is presented in Table 9-14.

59

Table 9-14. Maximum activity concentration of drainage water.

Nuclide Conservative

estimate (Bq/l) Realistic esti-mate (Bq/l)

Co-60 1.46E+09 1.46E+08 Ni-59 7.98E+07 7.98E+06 Ni-63 8.15E+09 8.15E+08 Zr-93 3.58E+04 3.58E+03 Nb-93m 5.86E+08 5.86E+07 Nb-94 3.23E+06 3.23E+05 Mo-93 8.86E+04 8.86E+03 Tc-99 1.34E+04 1.34E+03 Ag-108m 6.79E+06 6.79E+05 Sn-121m 1.68E+06 1.68E+05 U-234 1.02E+02 1.02E+01 U-235 1.00E+00 1.00E-01 U-236 2.48E+01 2.48E+00 U-238 2.25E+01 2.25E+00 Np-237 4.00E+01 4.00E+00 Pu-238 3.61E+05 3.61E+04 Pu-239 2.33E+04 2.33E+03 Pu-240 3.84E+04 3.84E+03 Pu-241 1.28E+06 1.28E+05 Pu-242 2.61E+02 2.61E+01 Am-241 3.55E+05 3.55E+04 Am-242m 1.72E+03 1.72E+02 Am-243 4.44E+03 4.44E+02 Cm-243 9.07E+02 9.07E+01 Cm-244 1.62E+05 1.62E+04 Cm-245 1.16E+02 1.16E+01 Cm-246 3.50E+01 3.50E+00 Total 1.03E+10 1.03E+09

9.6 Uncertainties in the activity inventory

The activity inventory has significant uncertainties, resulting in a variation of at least an order of magnitude between the conservative and realistic estimate. In reality, the varia-tion may be even larger than this, and furthermore the activity release may vary signifi-cantly between consecutive fuel batches. Hence, the uncertainties exceed significantly those of e.g. in the activity inventory of spent fuel. The most significant uncertainties in the activity inventory of the waste to be disposed of relate to the amount of crud ending up in the operational waste as well as the damage frequency of the fuel rods in the encapsulation process. Further uncertainties comprise e.g. the release fractions of fission products in dry conditions and the residence time of crud in the reactor. In addition, low uncertainty results from the assumed operational times of the power plant units and the encapsulation plant. Another issue affecting the activity content of the drum with dried waste is the filling time. One also shall bear in

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mind that the activity in the drums with dried waste may vary significantly from one drum to another. The uncertainties of the activity inventory may result in over-dimensioning of the sys-tems, such as the thickness of the radiation shield for the drum. The uncertainty can be reduced with further studies, e.g. by measuring the amount and spalling fraction of crud on the surfaces of the fuel assemblies, or based on operational experience from other fuel handling facilities. On the other hand, if the waste management in the initial phase of the operation relies on the OL3 power plant unit, the dimensioning of the waste han-dling systems at the encapsulation plant can later be revised based on the observed ac-tivity and volume accumulation of the waste. The uncertainties in the activity inventory increase the uncertainty of the results of the safety case, but this uncertainty is also ex-pected to decrease as operational experience is gained.

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10 SUMMARY

Solid and liquid radioactive waste is generated at the encapsulation plant. The waste shall be treated and packed in a way appropriate for the final disposal and long-term safety. In this report, the waste handling is designed assuming the operation of the en-capsulation plant as an independent facility. The design utilized to a great extent the operational experience gained from the handling and final disposal of the operational waste at the Loviisa and Olkiluoto nuclear power plants. However, it is to be pointed out that the waste amount generated at the encapsulation plant is relatively small, and thus utilizing the waste handling systems at the Olkiluoto power plant units may be worthwhile as long as they are in operation. A further assessment of this possibility is beyond the scope of this report. According to the guiding principle in the Nuclear Energy Act, the amount of nuclear waste generated in the use of nuclear energy shall be kept as low as reasonably possible by limiting the generation of waste. Another means for that is the clearance of waste from regulatory control, when possible. Solid compressible waste is compacted result-ing in a smaller volume of waste to be disposed of. Metal waste is decontaminated, and recycled if possible. Active metal may also be sent for melting. Active drainage water is dried in drums. One 200 liter drum can store the activity of drainage water accumulated during about 9 years. Especially at the initial stage of the operation of the encapsulation plant it may be worthwhile to utilize the in-drum drying system of OL3 power plant unit in the treatment of the liquid waste. In this way, operational experience and data on the real generation rate of radioactive liquid waste could be gained, and the best methods for their treatment could be selected based on that experience. Encapsulation waste is packed into drums and metal boxes, and the waste packages are disposed of in the LILW-repository for the encapsulation waste. The waste packages are transferred to the repository mostly using the canister lift, but larger components may also be transported via the access tunnel. The generation rate of drums is about 10 per year and that of metal boxes less than one per year during the operation of the encapsu-lation plant. A total of 12 drums of dried waste are estimated to be produced during the entire operational time of the encapsulation plant, and these drums will be emplaced into a separate concrete basin in the final repository. The total amount of waste to be disposed of as radioactive waste is about 640 m3 (including decommissioning waste). In the final repository special attention is to be paid on the condition of the waste pack-ages and release barriers, if it is to be kept open for the entire operation time of the en-capsulation plant. If the conditions in the repository cannot be kept favorable, the waste packages may degrade to an extent requiring re-packing. Another option worth consid-ering is an aboveground interim storage and a later construction of the repository. If the waste is treated in the waste handling facilities of the Olkiluoto power plant, it could be worthwhile to study the possibility for the final disposal in the Olkiluoto LILW-repository (VLJ-repository). This report presents an updated activity inventory for the waste from the encapsulation plant. It is assumed that no significant amounts of water can enter the gap between the fuel pellet and cladding as a consequence of fuel damages, and hence the release of fis-

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sion products remains low. Also the amount of crud from the fuel assemblies and end-ing up into the operational waste is re-evaluated. Due to the limited initial data the activ-ity inventory for the crud is formed by combining different sources of data. The combi-nation of the activity inventories for crud and fission products results in an inventory of the waste to be disposed of as radioactive waste. The inventory at the end of the opera-tion of the encapsulation plant (in 2120) is conservatively estimated to be 5.3·1010 Bq. The corresponding figure in the assessment made in 2012 was 1.7·1012 Bq (Paunonen et al. 2012). The lower inventory originates in the first place from the reduced amount of fission products released from the spent fuel. That is a consequence of reduced number of damaged fuel rods and the assumption that releases occur in dry conditions. The ac-tivity estimate for the spalling crud is about half of the previous estimate. The drum with dried waste can have a high dose rate, possibly up to 97 mSv/h accord-ing to the conservative activity estimate. Therefore the drum is to be handled by remote control and shielded during transport and storage.

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