Symposium JJ – Scientific Basis for Nuclear Waste Management XXV
Research Article
In Situ Isotopic Analysis of Uraninite Microstructures from the Oklo-Okélobondo Natural Fission Reactors, Gabon
- Mostafa Fayek, Keld A. Jensen, Rodney C. Ewing, Lee R. Riciputi
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- 21 March 2011, JJ8.5
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Uranium deposits can provide important information on the long-term performance of radioactive waste forms because uraninite (UO2+X) is similar to the UO2 in spent nuclear fuel. The Oklo-Okélobondo U-deposits, Gabon, serve as natural laboratory where the long-term (hundreds to billions of years) migration of uranium and other radionuclides can be studied over large spatial scales (nm to km). The natural fission reactors associated with the Oklo- Okélobondo U-deposits occur over a range of depths (100 to 400 m) and provide a unique opportunity to study the behavior of uraninite in near surface oxidizing environments versus more reducing conditions at depth. Previously, it has been difficult to constrain the timing of interaction between U-rich minerals and post-depositional fluids. These problems are magnified because uraninite is susceptible to alteration, it continuously self-anneals radiation damage, and because these processes are manifested at the nm to μm scale. Uranium, lead and oxygen isotopes can be used to study fluid-uraninite interaction, provided that the analyses are obtained on the micro-scale. Secondary ionization mass spectrometry (SIMS) permits in situ measurement of isotopic ratios with a spatial resolution on the scale of a few μm. Preliminary U-Pb results show that uraninite from all reactor zones are highly discordant with ages aaproaching the timing of fission chain reactions (1945±50 Ma) and resetting events at 1180±47 Ma and 898±46 Ma. Oxygen isotopic analyses show that uraninite from reactors that occur in near surface environments (δ18O= −14.4‰ to −8.5‰) have reacted more extensively with groundwater of meteoric origin relative to reactors located at greater depths (μ18O= −10.2‰ to −7.3‰). This study emphasizes the importance of using in situ high spatial resolution analysis techniques for natural analogue studies.
Nature and Effect of the Alteration Layer During Nuclear Waste Glass Dissolution
- A. Gauthier, P. Le Coustumer, J-H. Thomassin
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- 21 March 2011, JJ13.4
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The goal of this study is to understand the role of the interface developed during R7T7 glass alteration. This glass has been leached in two different aqueous media (pure water, silica rich water and phosphorous rich water). The lixiviation tests have been optimized to assess the role of the alteration layer developed on the surface of the glass. The solution and the solid have been characterized by ICP-MS and TEM/X-EDS respectively. The results put in evidence that a complex alteration layer is formed. Its texture, structure and chemistry are discussed with respect to the evolution of the solution during the tests. The alteration layer is always present on the surface of the glass and is considered to control (at short time) diffusion of the different species through the layer. Further study must be undertaken to assess the evolution and the stability of the interface for longer time periods.
The SR 97 Safety Assessment of a KBS 3 Repository for Spent Nuclear Fuel – Overview, Review Comments and New Developments
- Allan Hedin, Ulrik Kautsky, Lena Morén, Patrik Sellin, Jan-Olof Selroos
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- 21 March 2011, JJ4.2
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In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Nuclear Fuel and Waste Management Company, SKB has carried out the longterm safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock.
The future evolution of the repository system is analysed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion.
The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. The results of the assessment also serve as a basis for formulating requirements and preferences regarding the bedrock in site investigations, for designing a programme for site investigations, for formulating functional requirements on the repository's barriers, and for prioritisation of research.
SR 97 has been reviewed both by an international group of OECD/NEA experts and by Swedish authorities. The NEA reviewers concluded that “SR 97 provides a sensible illustration of the potential safety of the KBS-3 concept”, and no issues were identified that need to be resolved prior to proceeding to the investigation of potential sites. The authorities' conclusions were in principal consistent with those of the NEA.
Uncertainties and lack of knowledge in different areas identified in SR 97 have strongly influenced the contents and structure of SKBs most recent research programme, RD&DProgramme 2001.
Since SR 97, the methodology for probabilistic consequence analyses have been further developed. Analytic approximations to the numerical transport models used in SR 97 have been developed. The new models have been used to extend the probabilistic calculations in SR 97.
Radioelement transport in the Bangombé nuclear reactor zone (Gabon) – Evidence from U and Sm isotopes
- Bros R., Kamei G., Ohnuki T.
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- 21 March 2011, JJ8.1
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As a natural analogue study of relevance to the safety margin of radioactive waste repository conditions, mineralogical and isotopic studies were carried out to assess the effects of the alteration of the Bangombé natural nuclear reaction zone. Mobilization and retention of reactor products were identified around the reactor. The high temperature alteration associated with the heat released during nuclear reactions lead to the partial dissolution of uraninite and the sorption of released U mostly by Al-chlorite. This was followed by the precipitation of secondary U-phases such as Zr-bearing uraninite and U-Ti oxides. The recent (<1 Ma) low temperature alteration lead to further dissolution of reactor core uraninite and mobilization of U and fissiogenic REE. Migrations took place laterally through hydraulically conductive fractures within sandstones and vertically through compacted shales, likely by diffusion-type process. Adsorption onto clays and Fe-oxides is the dominant mechanism of retention as inferred by the results of selective extraction experiments. The amounts released appear to be low compared with the quantities initially produced in the core and the transfers may have been limited within several meters around the reactor.
Aging of a Bitumen Waste form in Wet Repository Conditions
- M.I. Ojovan, N.V. Ojovan, Z.I. Golubeva, I.V. Startceva, A.S. Barinov
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- 21 March 2011, JJ11.75
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SIA ‘Radon’ has been performing field tests of bitumen waste forms holding low and intermediate level wastes (LILW) for about three decades. The waste forms were made at an industrial bituminization plant from the actual LILW wastes. This paper presents results from analyzing the samples of the bitumen waste material taken from the bulky bitumen block with waste salts loading 31 wt.% after storing the block in a shallow-ground repository for 12 years. Rich in natrium nitrate NPP-operational waste was incorporated in bitumen.
Salts were separated from bitumen in some waste form samples. Non-homogeneous distribution of both salts and radionuclides was detected in vertical direction of the bitumen block. Specific radioactivity of the extracted bitumen was of the same order as the specific radioactivity of salts, in some cases even greater.
Bitumen samples free of salts were further separated into three main bitumen fractions, asphaltenes, saturated and aromatic hydrocarbons, using methods of solvent extraction. Essentially all radioactivity of the bitumen was associated with the asphaltene fraction. Aging of the bitumen waste form led to increase in asphaltene fraction content (with minimum 4%) and hardening of the waste material. The study has revealed a significant transfer of the waste salts radioactivity to the asphaltene fraction of the bitumen matrix. Changes in the properties of the bitumen waste form will be taken into consideration in modeling the long-term behavior of the bitumen waste materials under repository conditions.
Building the Safety Case for a Disposal Facility for Spent Fuel, HLW and Long-lived ILW in Switzerland
- Paul A. Smith, Piet Zuidema, Lawrence H. Johnson, Jürg W. Schneider, Peter Gribi
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- 21 March 2011, JJ4.5
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This paper describes a generic methodology for building the safety case for a geological repository, which is currently being applied to a possible facility for spent fuel, vitrified highlevel waste and long-lived intermediate-level waste in the Opalinus Clay of Northern Switzerland. The methodology involves:
1. the identification of certain basic disposal principles,
2. the choice of a disposal system, via a flexible repository development strategy,
3. the derivation of the system concept, based on current understanding of the phenomena that characterise, and may influence, the disposal system and its evolution,
4. the derivation of a safety concept, based on reliable, well understood and effective pillars of safety,
5. the illustration of the radiological consequences of the disposal system through the definition and analysis of a wide range of assessment cases, and
6. the compilation of the arguments and analyses that constitute the safety case, as well as guidance for future stages of the repository programme. A range of measures, including audits, are used to promote completeness of the phenomena considered in the safety case, and to avoid inadvertent bias.
The Influence of Near Field Redox Conditions on Spent Fuel Leaching
- Kastriot Spahiu, Ulla-Britt Eklund, Daqing Cui, Max Lundström
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- 21 March 2011, JJ14.5
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In a repository, the spent fuel could come in contact with groundwater if the canister or container has breached. The system may be quite complex with oxygen-free water, uranium dioxide, a corroding metal, such as iron, and a radiation field present at the same time. In an anaerobic environment iron and mild steel will corrode and hydrogen will be evolved. The equilibrium hydrogen pressure for this reaction is very high. At some time after water intrusion, there will be large amounts of dissolved hydrogen in the near field, corresponding to a partial pressure at least equivalent to the hydrostatic pressure at the repository depth. For this reason, we investigated the leaching behavior of 0.25-0.5 mm sized fragments of PWR spent fuel (43 MWd / Kg U) in simulated groundwater solution (10 mM NaCl and 2 mM HCO3-) under 5 MPa hydrogen and argon pressure. In a leaching experiment under 5 MPa hydrogen at 25 °C, the total U concentration was found to be <10−8 M. After refilling of the autoclave with new solution at 70°C, the total U concentration first increased to 10−6.3M, and then quickly decreased to 10−8 M. The leaching behavior of uranium and other fuel components indicates that under pressurized hydrogen, the spent fuel dissolution is substantially hindered. Leaching results obtained after the substitution of hydrogen by argon at the same pressure and temperature are also presented. Finally, some results on spent fuel leaching under pressurized argon are presented and comparatively discussed.
Leaching of Np and Tc From Doped Nuclear Waste Glasses in Clay Media: The Effects of Redox Conditions
- Véra Pirlet, Karel Lemmens, Pierre Van Iseghem
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- 21 March 2011, JJ13.6
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Boom Clay is the candidate geologic formation in Belgium for disposal of vitrified high level waste. In the waste glass, 237Np and 99Tc are some of the principal radionuclides. The leaching behaviour of these nuclides has been studied in dissolution tests involving glass doped with 237Np and 99Tc and was found to depend strongly on the redox conditions of the media contacting the glass. Static tests with the reference R7T7 glass and the reference PAMELA SM539 glass were performed in two clay media that may interact with the glass during the geological disposal. The first test medium consisted of FoCa-clay (Fourges-Cahaignes-clay), which is an oxidised clay. The second test medium was a mixture of backfill material, consisting mainly of FoCa-Clay, with metal corrosion products and Boom Clay. Lower Np and Tc concentrations were found in Boom Clay compared to FoCa-clay. In FoCaclay, Np and Tc were predominately in their oxidised form. Although Tc is mostly present in the soluble pertechnetate form, the Np concentrations are lower than expected for the chemistry of Np(V) in the medium. The concentrations of the radionuclides are steady-state concentrations rather than thermodynamic concentrations. Different mechanisms can be invoked to explain these Np concentrations. In Boom Clay, the measured Np concentrations are close to the solubility limit whereas the Tc concentrations are slightly higher than those calculated in Boom Clay at thermodynamic equilibrium. The initial specific activity of the radionuclide in the glass was found to influence the soluble concentration. Np and Tc were not found to be retained in the glass reaction layer for both media.
Effect of ZrO2 on the glass durability
- M. Lobanova, A. Ledieu, P. Barboux, F. Devreux, O. Spalla, J. Lambard
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- 21 March 2011, JJ15.1
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Borosilicate glasses were prepared with the molar composition 70 SiO2-15 Na2O-15B2O3-n ZrO2 with n ranging from 0 to 10. The glasses were studied by conventional static dissolution tests of powders at 90°C in pure water and in buffered solutions for long times (months) and short times (minutes). During the first minutes of alteration in a buffered solution, sodium is rapidly leached until its loss becomes controlled by the silicon hydrolysis. The experimental data show that the introduction of zirconium drastically reduces the initial dissolution rate (Vo) of the glass. Zirconium strengthens the silica network but also strongly modifies the porous layer morphology. In the case of glasses with small Zr contents (less than 2%), the silica dissolution rate decreases but the formation of a passivating alteration layer is also delayed. As a result, small amounts of zirconium paradoxically decrease the loss of silica but increase the final loss of sodium and boron in the static leaching tests. Larger zirconium contents (above 5%) increase the durability of the glass regarding the initial dissolution rate and the final concentration of all elements.
Demonstration of the Feasibility of Recovering Americium and Curium Isotopes from a Lanthanide Borosilicate Glass
- Tracy S. Rudisill, David K. Peeler, Thomas B. Edwards
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- 21 March 2011, JJ11.69
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A solution containing kilogram quantities of highly radioactive isotopes of americium and curium (Am/Cm) is currently stored in a process tank at the Department of Energy's Savannah River Site. This tank and its vital support systems are old, subject to deterioration, and prone to possible leakage. To address the stabilization of this material, vitrification of the isotopes has been considered. Potentially, the glass could be shipped to the isotope production and distribution programs at the Oak Ridge National Laboratory for californium-252 production and use by the transplutonium research community. However, before the Am/Cm could be used in these programs, it must be recovered from the glass.
To demonstrate the feasibility of recovering the Am/Cm isotopes from a glass, a series of small-scale experiments was performed as part of a compositional variability study. Glasses fabricated during the study utilized lanthanide elements as surrogates for Am/Cm due to the high specific activity of these materials. In the dissolution tests, glass formulations representative of potential uncertainties in the composition of the Am/Cm solution were fabricated, ground to a -35 to +60 mesh particle size, and dissolved in 8M nitric acid at 110°C. Under these conditions, at least 98% of the lanthanide oxides in the glass dissolved in less than 2 h meeting a recoverability criterion established for the vitrification process and imposing no limitations on the acceptable glass composition region.
Dissolution of the lanthanide borosilicate glasses was described by a spherical particle model based on the observation that the rate of change of the mass to surface area ratio remains constant. Calculation of dissolution rates using the model showed that the rate was proportional to the lanthanide oxide concentration in the glass. When silicon oxide (SiO2) was replaced with a lanthanide element at higher (simulated Am/Cm) loadings, the glass became more easily dissolved in nitric acid due to the solubility of the lanthanide oxides compared to SiO2.
Alteration of a Zirconolite Glass-Ceramic Matrix under Hydrothermal Conditions
- Christelle Martin, Isabelle Ribet, Thierry Advocat
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- 21 March 2011, JJ6.4
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Glass-ceramic matrices based on zirconolite (CaZrTi2O7) are being considered for specific conditioning of plutonium or the minor actinides. The actinides are distributed throughout the zirconolite crystals and the residual glass phase. Since zirconolite alteration is extremely limited, however, actinide release from the glass-ceramic material is mainly attributable to alteration of the residual glass. Zirconolite glass-ceramic specimens and specimens corresponding to the residual glass phase alone were therefore altered under hydrothermal conditions (150°C) and under initial rate conditions (100°C) to compare their kinetic behavior and estimate the effect of the crystals on material alteration. Under hydrothermal conditions, alteration occurred during the first few days: SEM observations showed greater alteration of the glass-ceramic material due to a phenomenon of preferential glass alteration around the zirconolite crystals; after three days the alteration rate had considerably diminished and both specimens exhibited similar behavior. Under initial rate conditions the initial rates differed due to a variation in the reactive surface area of the glass-ceramic.
Natural Colloids in Groundwater from a Bentonite Mine- Correlation between Colloid Generation and Groundwater Chemistry -
- Yoshio Kuno, Gento Kamei, Hiroyuki Ohtani
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- 21 March 2011, JJ8.4
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Colloids, mainly montmorillonite, generated by erosion of compacted bentonite by groundwater flow might enhance the transport of radionuclides from a radioactive waste repository. The influence of aqueous chemistry (i.e. pH, cation concentration and valence) on the dispersion of montmorillonite colloid was studied. Colloids were flocculated under higher cation concentrations ([Na+] > 10−2 M, [Ca2+] > 10−3 M) and/or under acidic condition (pH = 4) by means of batch-type experiments. Derjaguin-Landau-Verwey-Overbeek (DLVO) theory was applied to estimate the stability of colloidal dispersion and then the limitation of theoretical calculations was pointed out. Groundwater samples were collected from two galleries at different depths of the Tsukinuno bentonite mine (northern Honshu, Japan) and investigated for the populations of colloids. The groundwater flows vertically through Tertiary sedimentary argillaceous rocks and fine tuff beds which are mined for bentonite. Low colloid concentrations were measured in these groundwater samples. This result suggests that the colloids cannot significantly disperse in the groundwaters under higher cation concentration ([Na+] > 10−2 M) or under acidic conditions. This result is consistent with those of the batch-type experiments.
Structural Alterations in Titanate Pyrochlores Induced by Ion Irradiation: Xray Photoelectron Spectrum Interpretation
- J. Chen, J. Lian, L. M. Wang, R. C. Ewing, J. Matt Farmer, L. A. Boatner
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- 21 March 2011, JJ11.34
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Titanate pyrochlores are candidate host materials for the disposition of plutonium from dismantled nuclear weapons. These pyrochlores also have potential applications as solid electrolytes and oxygen gas sensors. The radiation-induced microstructural evolution of titanate pyrochlores has been characterized using x-ray photoelectron spectroscopy (XPS). XPS analysis of the Ti 2p and O 1s binding energy shifts of REE2Ti2O7 (REE: rare earth) surface layers before and after irradiation shows that the primary manifestations of amorphization are distortions or changes in the coordination number of the titanium polyhedra. A model based on glass network structure was developed, and predictions of the relative susceptibilities for amorphization of the titanate pyrochlores are obtained that are consistent with the experimental results.
Creep Analyses of Titanium Drip Shield Subjected to Rockfall Static Loads in the Proposed Geologic Repository at Yucca Mountain
- Brett W. Neuberger, Charles A. Greene, G. Douglas Gute
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- 21 March 2011, JJ11.7
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The U.S. Department of Energy (DOE) has included a drip shield (DS) as a principle component of the engineered barrier system (EBS) for the proposed high-level nuclear waste geologic repository at Yucca Mountain, Nevada. The current DS design consists of titanium grade 7 (Ti Gr 7) plates and Ti Gr 24 support beams and bulkheads. The intended functions of the DS are to divert dripping water around and prevent rockfall damage to the waste package (WP). Sustained static loading of the DS may occur as a result of rockfall or drift collapse. These static loads may cause residual stress that approaches the yield stresses of the different DS materials. This level of residual stress would enable various creep mechanisms to transpire. A preliminary assessment of the potential for DS creep after a dynamic rock block impact is presented in this paper by expressing the DS residual Von Mises stress levels as fractions of the Ti alloy yield stress (YS). It was determined, using creep data from similar alloys that the residual stress levels within a DS after a 2-tonne rock block impact per DS segment length could cause creep in both the Ti Gr 7 plates and Ti Gr 24 bulkheads and support beams. The results of this study will assist the U.S. Nuclear Regulatory Commission (NRC) in evaluating the risk significance of the expected DS performance characteristics under actual repository conditions.
Heavy Ion Irradiation of Zirconate Pyrochlores
- J. Lian, L. M. Wang, J. Chen, R. C. Ewing, K. V. G. Kutty
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- 21 March 2011, JJ11.35
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Zirconate pyrochlores, A2Zr2O7, are important potential nuclear waste forms for Puimmobilization. The binary Gd2(Ti2-xZrx)O7 has been shown to have increasing resistance to ionirradiation damage with the increasing Zr content, and Gd2Zr2O7 is radiation resistant to a 1 MeV Kr+ ion irradiation at 25 K to a dose of 5 dpa. In this study, a 1.5 MeV Xe+ irradiation was completed for zirconate pyrochlores A2Zr2O7 (A=La, Nd, Sm, Gd). The radiation resistance decreases with an increase of the ionic radius of A-site cation. La2Zr2O7 is the first zirconate pyrochlore to be amorphized by ion beam irradiation, and the critical amorphization temperature, Tc, is ∼310 K. The susceptibility of La2Zr2O7 to ion beam damage is related to its structure, which shows the largest deviation from the ideal fluorite structure. These results are also consistent with calculations of the cation antisite formation energy in the pyrochlore structure. The ion irradiation-induced pyrochlore-to-fluorite transformation occurred in all of the irradiated zirconate pyrochlore phases. Based on the results for Gd2Ti2-xZrxO7 and A2Zr2O7, the defect fluorite structures are stable when the ionic radii ratio rA/rB≤1.54; beyond this limit, the defect fluorite structure becomes increasingly unstable relative to the amorphous state.
Development of the Post-Closure Safety Case for the Low Level Waste Disposal Site at Drigg, United Kingdom
- Len Watts
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- 21 March 2011, JJ7.2
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This paper provides an overview of the development of the post-closure safety case (PCSC) for the solid low level radioactive waste (LLW) disposal site at Drigg. The paper outlines the background to the site, the implementation of a systematic approach to understanding the site and undertaking a post-closure radiological safety assessment (PCRSA), communication activities, key factors in the PCSC and issues to be investigated further in the forward programme.
The Drigg site is the UK's principal near-surface facility for disposal of LLW. Disposals commenced in 1959 to trenches whereas current practice involves disposal to concrete vaults. Disposals are carried out under an authorisation from the UK Environment Agency. Periodically the Environment Agency reviews the authorisation to ensure consistency with current regulatory requirements. British Nuclear Fuels plc (BNFL) will produce an updated PCSC in 2002; in preparation for this, BNFL published in March 2000 a status report on the development of the PCSC.
BNFL and the Environment Agency conduct an information exchange process on the development of the PCSC. This process has led to notable improvements, as well as affording a mechanism for providing clarifications and building confidence in the PCSC. BNFL undertakes other communication activities including scientific publication, conference presentation, participation in international programmes (such as those of the IAEA and NEA) and liaison with local communities. Peer review of the PCSC is also being carried out.
Central to the safety case is a systematic PCRSA, which promotes transparency and traceability and demonstrates a rigorous treatment of relevant uncertainties. The approach aligns with international best practice for near-surface disposal facilities and includes the assessment of features, events and processes and the use of tools such as conceptual model uncertainty forms. This formalisation helps to identify and understand the key factors for the PCSC. Additional qualitative or simple quantitative information is used to augment the PCRSA so as to provide a wider context to inform decision-making.
An Overview of the H12 Performance Assessment in Perspective
- Kaname Miyahara, Hitoshi Makino, Tomoko Kato, Keiichiro Wakasugi, Atsushi Sawada, Yuji Ijiri, Aki Takasu, Morimasa Naito, Hiroyuki Umeki
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- 21 March 2011, JJ4.3
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The H12 performance assessment (PA) provided a test for the robustness of a HLW repository system concept based on structured siting and design, taking account of a wide range of potentially suitable Japanese geological environments. The generic nature of the host rock in the H12 assessment means, however, that emphasis is placed verymuch on strong EBS performance. The assessment included a comprehensive evaluation of uncertainty and potentially detrimental factors, including perturbations due to external events and processes. Despite the considerable uncertainty at the current stage of the Japanese program, a safety case that is adequate for the aims of the assessment can be made by a strategy of employing conservatism where there is uncertainty and stressing the reliability and effectiveness of the performance of the near-field. The aim of this paper is to present the H12 PA in a way which makes the PA process clearer and the implications of the results more meaningful, both to workers within the PA field and to a wider technical audience.
Incorporation of Cerium and Neodymium in a Uranyl Hydroxide Solid
- C.W. Kim, D.J. Wronkiewicz, R.J. Finch, E.C. Buck
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- 21 March 2011, JJ11.66
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The migration behavior of radionuclides in a nuclear-waste repository will be influenced, in part, by their equilibrium solubilities in the presence of radionuclide-bearing solids and/or adsorption affinities as trace components onto the surfaces of solid host phases. Uranium phases that precipitate on the surface of altered spent nuclear fuel may thus influence the mobility of radionuclides released in the near-field environment, since by proximity, these will be the first phases that the radionuclides encounter following their release from the spent fuel matrix.
We have evaluated the potential for incorporating radionuclides into crystalline compounds by precipitating uranyl phases from aqueous solutions containing dissolved rare earth elements (REE; 2.1 ppm Ce4+, 4.6 ppm Ce4+, or 286 ppm Nd3+). Rare earth elements serve both as monitors for evaluating the potential behavior of REE radionuclides and as surrogate elements for the actinides (e.g., Ce4+ and Nd3+ for Pu4+ and Am3+, respectively). Although the crystalline compound that formed in the present set of experiments has not been positively identified, x-ray diffraction profiles suggest the presence of a uranyl hydroxide (UO2(OH)2) as the principal reaction product. An analysis of the crystalline products indicate a progressive decrease in concentration of cerium; from 26, to 20, and finally 11 ppm for crystals produced in 7-, 35-, and 190-day tests, respectively (Kd = 14, 11, 3, respectively). Results with neodymium display a similar trend, with concentrations in the solid decreasing from 1240 to 922 ppm between 7 and 35 days of reaction (Kd = 14 and 11, respectively). The decreasing concentration of REEs in the uranyl crystals can be correlated with both a coarsening in crystal size and a decrease in the concentration of dissolved uranium over time. Thus, REE incorporation in the crystalline solids decreases in conjunction with a decrease in the ratio of surface area/volume of the crystals, a decrease in the rate of crystal growth as uranium concentrations are lowered, or both. These data also suggest that adsorption of REE (and by analogy, actinides) onto crystal surfaces and subsequent trapping by crystal overgrowth processes may play key roles in the limiting the mobility of radionuclides in a nuclear waste repository.
Synthesis and Evaluation of Uranium and Thorium Imprinted Resins
- K.L. Noyes, M. Draye, A. Favre-Réguillon, J. Foos, A. Guy, K.R. Czerwinski
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- 21 March 2011, JJ12.6
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Ion exchange resins were created to complex either UO22+ or Th4+ from aqueous solutions. The nitrate salt of the target metal ion was dissolved in CCl2H2, while the resin was created around the ions to provide a unique structure based upon each metal. These resins were synthesized by a radical polymerization method, producing a reusable organic solid. The resins were qualified by obtaining values for their proton exchange capacities and data to define their complexation kinetics. Proton exchange capacities were determined using an indirect titration and were found to be 6.40 meq/g for the uranium-based resin and 4.61 meq/g for the thorium-based resin. Data for the resins' kinetics were obtained at pH 1.0, 2.5, 4.0, and 5.5. Results show that the templated resin rapidly removed the target actinides from aqueous solution under experimental conditions. Once loaded with metal, the ions can easily be removed with 5 M HNO3 and reused.
Precipitation of Crystalline NpO2 During Oxidative Corrosion of Neptunium-Bearing Uranium Oxides
- Robert J. Finch
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- 21 March 2011, JJ11.60
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We report crystallization of NpO2 during aqueous corrosion of unirradiated Np-bearing U3O8 under nominally oxidizing conditions. Powders of Np0.33U2.67O8 (Np:U = 1:8) were reacted in humid air at 90° and 150°c for several weeks in sealed stainless-steel vessels. Reacted solids were examined by scanning and transmission electron microscopies (SEM and TEM) and X-ray powder diffraction (XRD). Dehydrated schoepite, (UO2)O0.25–z(OH)1.5+2z (0 ≤ z ≤ 0.15), is the predominant U(VI) compound formed in these experiments and is a minor sink for Np (containing 2 wt.% Np, maximum). The primary sink for Np during corrosion of Np0.33U2.67O8 at 150°C is crystalline NpO2, which crystallized within 2 weeks of reaction in humid air. Corrosion of Np0.33U2.67O8 at 90°C resulted in crystallization of approximately equal proportions of crystalline NpO2 and Np2O5 during 16 weeks of reaction in humid air.