R Progress report on evaluation of long term safety of proposed SFL concepts. Lena Z Evins, Svensk Kärnbränslehantering AB. - PDF

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R Progress report on evaluation of long term safety of proposed SFL concepts Lena Z Evins, Svensk Kärnbränslehantering AB December 2013 Svensk Kärnbränslehantering AB Swedish Nuclear Fuel and Waste

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R Progress report on evaluation of long term safety of proposed SFL concepts Lena Z Evins, Svensk Kärnbränslehantering AB December 2013 Svensk Kärnbränslehantering AB Swedish Nuclear Fuel and Waste Management Co Box 250, SE Stockholm Phone ISSN Tänd ett lager: SKB R P, R eller TR. ID Progress report on evaluation of long term safety of proposed SFL concepts Lena Z Evins, Svensk Kärnbränslehantering AB December 2013 Keywords: SFL, Concept, Barrier materials, Safety functions. A pdf version of this document can be downloaded from Preface This report is a result of a collaborative effort in the SFL Concept study, performed at SKB between 2011 and The author has summarized the work performed by the project group: Mattias Elfwing, Per Mårtensson, Pär Grahm, Mikael Gontier, Sofie Tunbrant and Lena Z Evins. In addition, this report has benefited from the reasoning and contributions from the following experts: Patrik Sellin, Ignasi Puigdomenech, Björn Gylling, Ulrik Kautsky, and Jens-Ove Näslund. SKB R Abstract This report aims to summarize work performed in the project SFL Concept study, carried out at SKB between 2011 and Different aspects of long-term safety are discussed and a preliminary evaluation of the proposed concepts is presented. The concepts that are subjected to the evaluation are described in a general manner and they are in no way optimized with respect to performance. There are, at the time of writing, still some major uncertainties surrounding the long-lived waste, both in terms of inventory and waste form. In addition, no site is selected. Optimization of the repository design will therefore follow in later stages of the development of the design, and be linked both with improved waste characterization and the results of future safety assessments. The purpose of the current report is two-fold: to aid the choice of concepts to be further developed and analysed, and to provide a starting point for further research efforts required for a future safety assessment. Requirements from regulations have guided the formulation of evaluation factors, which are used as a framework around which the studies and arguments are connected. For long-term safety, two evaluation factors have been formulated: 1) Feasibility of making a post-closure safety assessment. 2) Robustness of the barrier safety functions. The proposed concepts are based around the idea of retarding the radionuclide release from the waste by surrounding the waste by different barrier materials in a geological repository placed 300 to 500 meters deep in crystalline bedrock. The different materials suggested are gravel, concrete, and bentonite. One concept combines all three of these materials, which suggests increased retardation potential, but also negative effects such as a possible detrimental interaction between the barrier materials. Also described are potential waste conditioning options, as well as a review of possible materials which may enhance sorption of some key radionuclides. The safety functions are based on the properties of the materials in terms of hydraulic conductivity, diffusivity, sorption and other chemical properties. Thus, processes that might influence these properties are identified and the potential effects, during the time frame required (at least 100,000 years) are evaluated. In light of the first evaluation factor, a brief appraisal of the current process understanding, as well as the potential to define, achieve and verify the initial state is provided. The second evaluation factor involves an estimate of performance in the long-term, focused around studies and arguments concerning the evolution of the safety functions. Processes in the natural barrier (the bedrock), as well as other features and processes in the natural system, are expected to affect the engineered barriers. The effect of the natural barrier therefore plays a part in the preliminary evaluation. Research efforts needed in this area for a future safety assessment are identified. Some work has already been initiated; however, this work is mainly connected to later stages in the development of the SFL repository. Nevertheless, a brief overview of important aspects to consider in this area for future research is provided here. Expressions for simple calculations for equivalent flow, i.e. the solute carrying capacity of water seeping in the rock surrounding a repository, is used in a comparison between the concepts. The results show that intact and pristine concrete is superior to clay when it comes to the retention of radionuclides; however, radionuclide transport will be dominated by diffusion in both, even in situations when the waste has a high hydraulic conductivity. Finally, using the evaluation factors as a guide, and the studies and arguments presented in the report, the preliminary evaluation suggests that the Concrete repository and the Clay repository are the most promising concepts. The Gravel repository is not suitable without further added safety functions. Using a combination of these barrier materials involves potential problems due to both concrete-bentonite interaction, as well as uncertainties concerning the initial state. SKB R Sammanfattning Denna rapport syftar till att sammanfatta arbete som utförts i projektet SFL Konceptstudie, vilket genomfördes vid SKB mellan 2011 och Olika aspekter av långsiktig säkerhet diskuteras och en preliminär bedömning av de föreslagna begreppen presenteras. De koncept som utvärderas här beskrivs på ett allmänt sätt och de är inte på något sätt optimerade. Det finns, i skrivande stund, fortfarande några stora osäkerheter kring det långlivade avfallet, både vad gäller radionuklid inventarium och avfallsformer. Dessutom återstår platsval för SFL. Optimering av förvarets utformning kommer därför att följa i senare stadier av utvecklingen av förvaret, och kopplas både med förbättrad avfallskarakterisering och med resultaten av kommande säkerhetsanalyser. Syftet med denna rapport är dubbelt: att underlätta valet av förvarskoncept som ska vidareutvecklas och analyseras, samt att ge en utgångspunkt för den ytterligare forskning som krävs för en framtida säkerhetsanalys. Krav från föreskrifter har styrt utformningen av de utvärderingsfaktorer, som används som ett stöd för de beskrivna undersökningar och argumenten. För den långsiktiga säkerheten har två utvärderingsfaktorer formulerats: 1) Möjligheten att genomföra en säkerhetsanalys. 2) Robusthet hos barriärernas säkerhetsfunktioner. De föreslagna koncepten baseras kring idén om att retardera radionuklidutsläpp från avfallet genom att omge avfallet med olika barriärer i ett geologiskt slutförvar, vilket placeras på meters djup i urberget. De olika föreslagna barriärmaterialen är grus, betong och bentonit. Ett koncept kombinerar alla tre av dessa material, vilket tyder på ökad retardationspotential, men också negativa effekter såsom en skadlig växelverkan mellan barriärmaterial. Möjliga konditioneringsmetoder beskrivs också, samt en översikt av material med potential att förbättra sorption av vissa nyckelnuklider. Säkerhetsfunktionerna är baserade på hydraulisk konduktivitet, diffusivitet, sorption och andra egenskaper hos dessa material. Processer som kan påverka dessa egenskaper identifieras och de potentiella effekterna av dessa under den föreskrivna tidsramen (minst år) utvärderas. I ljuset av den första utvärderingsfaktorn ges en kort bedömning av den aktuella processförståelsen, liksom möjligheten att definiera, uppnå och kontrollera initialtillståndet. Den andra utvärderingsfaktorn innebär en uppskattning av funktion på lång sikt, där fokus ligger på studier och argument avseende utvecklingen av säkerhetsfunktionerna. Processer i den naturliga barriären (berggrunden), samt andra funktioner och processer i det naturliga systemet, kommer att påverka de tekniska barriärerna. Och spelar därför en viss roll i den preliminära utvärderingen. Forskningsinsatser har identifierats, och visst arbete har redan inletts, men detta arbete främst kopplat till senare skeden i utvecklingen av SFL-förvaret. En kort översikt ges här över viktiga aspekter att beakta för framtida forskning inom dessa områden. Uttryck för enkla beräkningar för ekvivalenta flödet, dvs. vattnets transportkapacitet, används i en jämförelse mellan koncepten. Resultaten visar att intakt betong är bättre än bentonitlera när det gäller radionuklidretention, men i båda fallen kommer radionuklidtransporten att domineras av diffusion. Detta gäller även när avfallet har en hög hydraulisk konduktivitet. Slutligen, med ledning av utvärderingsfaktorerna, samt de undersökningar och argument som presenteras i rapporten, tyder den preliminära utvärderingen på att Betong-förvaret och Bentonit-förvaret är de mest lovande koncepten. Grus-förvaret är inte lämpligt utan ytterligare säkerhetsfunktioner. Ett förvar som kombinerar dessa barriärmaterial innebär potentiella problem, på grund av interaktionen mellan betong och bentonit interaktion samt osäkerheter rörande initialtillstånd. 6 SKB R-13-41 Contents 1 Introduction Background the current and previous SFL concepts Availability of updated information A note on uncertainties and the iterative process Purpose and contents of this report 9 2 Method of evaluation and evaluation factors A background to safety assessments Safety functions Application of evaluation factors 12 3 Concept descriptions The method for concept identification and development Candidates for expanded analysis The Gravel repository The Concrete repository The Clay repository The Super silo Conditioning methods and other additional features Sealed container Melting of metallic waste Vitrification Chemically improved retention 18 4 Processes in the engineered barriers Gravel barrier processes Concrete barrier processes Bentonite barrier processes Chemically improved retention 21 5 Processes in the natural system Geosphere processes Climate and scenarios The surface system and the biosphere 26 6 Preliminary assessments of near-field transport capacity Description of the conceptual models The hydraulic cage Diffusion dominated transport through buffer Transport by flow through the waste Degradation of concrete and bentonite Comparisons based on the preliminary results 29 7 Evaluation of concepts according to evaluation factors Evaluation factor 1 Feasibility of making a post-closure safety assessment Evaluation factor 2 Robustness of the barriersafety functions Summary of evaluation according to evaluation factors 38 8 Outlook and concluding remarks 39 References 41 SKB R 1 Introduction 1.1 Background the current and previous SFL concepts The project SFL Concept study, carried out at SKB between 2011 and 2013, is a first step to constrain possible repository concepts for long-lived low and intermediate level waste. The goal is to choose maximum two different concepts for continued study and future evaluation using safety assessment methodology. The background to these studies is found in the 1999 safety assessment of SFL 3 5 (SKB 1999) and the reference inventory used for that assessment (SKBdoc ). It was noted in that previous assessment that mainly two factors were important for the calculated result. Both of these factors were site dependant (hydrogeological conditions and biosphere conditions). Thus, the suggestion from the previous preliminary assessment was to reduce the release to the far-field by improving the engineered barriers. The repository design should be modified so that the barriers are more effective for reducing the release of the dose-dominant radionuclides: for the SFL waste, these were, in the 1999 study, mainly Cl-36 and Mo-93 (section 10-3 in SKB 1999). 1.2 Availability of updated information There are two main types of waste from which the source term is derived: one type of waste already exists and is in storage on the Studsvik site (so-called legacy waste, which includes waste from the early Swedish nuclear research programmes), while most of the other main type (metallic parts from the nuclear reactors) is expected to accumulate in the future. Thus, the 1998 waste inventory was based both on knowledge of existing waste, and forecasts. The radionuclide inventory was calculated from this, using correlation factors (SKBdoc ). In parallel with the SFL Concept study, the SFL Reference inventory from 1998 is updated. The idea is to have an updated inventory for the next phase of deeper evaluation of the chosen concept(s). This new inventory is reported elsewhere (Herschend 2013). However, it should be noted that the legacy waste is, in its present state, complex, containing various types of materials. This waste may require further treatment and conditioning in order to agree with acceptance criteria. Thus, for the present report, the Reference inventory from 1998 is considered. 1.3 A note on uncertainties and the iterative process Clearly, there are still some major uncertainties surrounding the SFL waste. This has implications for choices of waste conditioning and repository design of the engineered barriers. Therefore, if and when there is new information, it should be possible to go back and adjust the suggested concept appropriately, if deemed necessary. The changed concept should then be assessed in a similar manner; this iterative process is inherent in the process of designing a repository with long-term safety in focus. 1.4 Purpose and contents of this report The purpose of this report is mainly to summarize the work performed regarding aspects of long-term safety, on which the evaluation of the proposed SFL repository concepts is based. In addition, this report is expected to function as a starting point for the more elaborate evaluations and assessments of the chosen concepts that are planned for the future. To reach these goals, this report describes the progress regarding the evaluation of long-term safety of different repository concepts suggested for the SFL Concept study. Information is provided on identified issues of concern for the long-term safety of a future SFL repository. In addition, the identification of such issues triggers further investigations and analysis, in order to aid the evaluation of the proposed concepts. Therefore, this report describes both the results of preliminary studies as well as suggestions for future research programs in different areas. SKB R 2 Method of evaluation and evaluation factors The evaluation of the proposed concepts with regards to post-closure safety is strongly connected to estimates of how well the chosen barriers will function together. In this chapter, certain basic concepts are presented and discussed, followed by a description of the approach used here to perform the evaluation. The present evaluation represents the first steps in a more long-term endeavour, based on research to enhance process understanding and modelling approaches. As understanding proceeds, this is expected to provide feed-back to the technical aspects of the proposed concepts and repository designs. 2.1 A background to safety assessments The complexity of the task to evaluate the long-term safety of a proposed repository is significant. The safety assessment methodology followed by SKB for a spent nuclear fuel repository is documented in SKB (2011) and a similar methodology will be followed in coming assessments of the SFL repository. A safety assessment requires knowledge about the initial state of the waste and the repository, and the natural system surrounding it. For all parts of the system, it is also required to evaluate the effects of all expected processes that can alter the system in the time frame required for compliance with the regulations. The time frame for the safety assessment of a repository containing long-lived radionuclides needs to cover at least 100,000 years, or the time for a glacial cycle, and it should not cover more than 1 million years (General advice to SSMFS 2008:37). This means that the time frame is comparable to that of the spent nuclear fuel repository. Due to the radioactive decay of the waste, the activity will decrease at a certain rate, depending on the radionuclide inventory. It is important to note the differences in decay time between the actinides, and the shorter-lived fission and activation products. The SFL waste contains both shorter lived radionuclides, such as Ni-63 (with half-life of 96 years), but also actinides and some longer lived activation products, such as Cl-36 (with a half-life of ca 300,000 years). Thus, the future safety assessment of the SFL repository is required to provide quantitative measures of risk for the time period of at least 100,000 years. However, the time following 100,000 years up to 1 million years, also need to be addressed, at least in a qualitative way. This means, that the future safety assessment should take into account all processes likely to affect the repository system during the one million years following closure. What is required is thus an evaluation of system evolution in a one million year perspective encompassing changes in climate, biosphere, surface system, bedrock, geochemistry, hydrology, engineered barriers, and waste. A safety assessment combines the most recent knowledge and development in all these areas, and extracts quantitative data for use in modelling of the system. The result is quantitative measure of risk as given by the model. It is also noteworthy that in the future safety assessment of SFL, the risk analysis needs to be most detailed for the first thousand years after repository closure. Available site-specific data and details regarding the early development of the repository need to be used as input. This requirement of a more detailed analysis of the first thousand years may be significant for SFL, depending on the possibility of slow radionuclide release already from the time of closure. 2.2 Safety functions In SKB s safety assessment methodology (SKB 2011), the barriers of a repository system are assigned one or several safety functions. The two main safety functions of SKB s spent nuclear fuel repository are containment and retardation. If the safety of the repository is built primarily upon containment, this implies that the chosen barriers are expected to fully contain the radionuclides within the repository. Applying retardation as a safety function implies that the barriers (both engineered and natural) are expected to efficiently retard the radionuclide migration once they have escaped the container, so that the exposure to the surface system is delayed and dispersed in time. SKB R A multi-barrier repository system is composed of different barriers each with its safety functions. A barrier safety function is defined as a role through which a repository component (barrier) contributes to safety (SKB 2011). An example is the bentonite buffer in the KBS-3 repository system; there are several safety functions provided by this barrier. One is protecting the waste container (in the KBS-3 case, the copper canister) from advective flow of groundwater. Another is filtering of colloids; this contributes to retardation of radionuclides. Following the methodology in SKB (2011), each safety function is associated with one or several quantitative indicators that show how well the safety function in question is upheld. In some cases also quantitative criteria can be defined such that when an indicator fulfils a criterion, the safety function is upheld. Due to the nature of the indicators, the corresponding criteria can be either quantitative or qualitative. It follows from the above that in a safety assessment, it is necessary to demonstrate how safety is related to the safety functions of the barriers and how these safety functions are affected by different processes which occur ov
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